ML20211P133
| ML20211P133 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/02/1997 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211P117 | List: |
| References | |
| 2.206, NUDOCS 9710200058 | |
| Download: ML20211P133 (3) | |
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U.S. NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50 282,50 306, and 7210' NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR PLANT, UNITS 1 AND 2 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION RECEIPT OF PETITION FOR DIRECTOR'S DECISION UNDER 10 CFR 2.206 Notice is hereby given that by a Petition filed purt .)nt to 10 CFR 2.206 on August 26, 1997, Prairie Island Coalition (Petitioner) requested that the NRC (1) suspend Northern States Power Company's (the licensee) Materials License No. SNM 2506 for cause under Section 50.100 of Title 10 of the Code of Federal Reoulations (10 CFR 50.100) until all materialissues regarding the maintenence, un!cading, and decommissioning processes and procedures, as described in the Petition and a similar Petition filed on May 28,1997,by the Prairie Island Indian Community, have been adequately addressed and resolved, and until the maintenance and unloeding processes and procedures in quostion are safely demonstrated under the scrutiny of independent third party review of the TN-40 cask seal maintenance and unloading procedure; (2) determine that the licensee v;olated 10 CFR 72.122(f) by using a cask design that requires periodic seal maintenance and emergency seal replacemant that must be performed in the plant storage pool; (3) determine that the licensee violated 10 CFR 72.122(h) by using a cask that must be placed into the pool for necessary maintenance and/or unloading procedures; (4) determine that the licensee violated 10 CFR 72.122(l) by loading casks and storing them before the licensee had procedures adequate to safely unload and decommission the TN-40 casks; (5) determine that the licensee violated 10 CFR 72.130 by using the TN-40 cask and failing to make
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{ provisions capable of accomplishing the removal of radioactive waste and contaminated materials at the time the independent spent fuel storage installation (ISFSI) is permanently
--decommissioned; (6) determine that the licensee violated 10 CFR 72.11 by falling to provide and include complete and accurate matulalinformation regarding maintenance ar.d unloading of TN40 casks in the application for the Prairie Island ISFSI and in subsequent submissions regarding cask maintenance and unloading issues; (7) d3termine that the licensee violated 10 CFR 72.12 by deliberately and knowingly submitting incomplete and inaccurate materialinformation regarding maintenance and unloading of TN-40 casks in the spplication for the Prairie Island ISFSI and in subsequent submissions regarding cask maintenanc,e and unloading issues; (8) require that the licensee pw/ a substantia! penalty for each cask loaded in violation of NRC regulations; (9) administer such other manctions -
for the alleged violations of NRC regulations as the NRC deems necessary and appropriate; (10) provide Petitioner the opportunity to participate in a public review of maintsnance, 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 the licensee's Technical Specifications for the Prairie Island ISFSI to ensure a demonstrated
= ability to in f act safely maintain, unload, and decommission TN40 casks; (12) review the licensee's processes and procedures for maintenance, unloading, and decommissioning, and if the licensee does not possess ::apability to unload casks, order the licrasee to build a " Hot Shop" for air unloading of casks and transfer of the fuel; (13) lnithte 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) initiate a formal 4
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- rulemaking proceeding to de*ine the parameters of retrievability required under 10 CFR 72.122(l); and (15) initiate a formal rulemaking proceeding for amendment 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 ISFSI.
The Petition has been referred to the Director of the Office of Nuclear Reactor Regulation. As provided by 10 CFR 2.206, further action will be taken within a reasonable
. time. Regarding the requests for formal rulemaking proceedings as detailed in items 13, 14, ar.d 15 in the Petition, the NRC staff is reviewing these requests in accordance with 10 CFR 2.802, " Petition for Rulemaking."
A copy of the Petition is available for inspection at the Commission's Public Document Room at 2120 L Street, NW., Washington, DC, and at the local public document room located at the Minneapolis Public Library, Technology and Science Department,300 Nicollet Mall, Minneapolis, MN.
Dated at Rockville, Maryland, this 2nd day of October 1997-FOR THE NUCLEAR REGULATORY COMMISSION a ector j Office of Nuclear Reactor Regulation l
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- Lake Ehmo, MN 55042
- Phone: 612 770 3861
- FAX 770 3976 August 26,1997 L. Joseph Callan Executive Director of Operations US Nuclear Regulatory Commission
. Washington, D.C. 20555
Dear Mr. Callan:
Please find enclosed a petition pursuant to Section 2.2%, Title 10 of the Code of Federal Regulations (CFR). The Prairie Island Coalition (PIC) hereby petitions the Nuclear Regulatory Commission (NRC) to suspend for cause the Northern States Power Co. (NSP)
Materials License No. SNM-2506 oceded to operate an independent Spent Fuel Storage Installation (ISFSI) at the Prairie Island Nuclear Generating Plant (PI).
A thorough review of the procedure developed by NSP for unloading Transnuclear dry storage casks (TN-40) in use at PI is necessary at this time because it is apparent that conditions for safely unloading TN-40 casks after a storage period have not been established. By operating the ISFSI at PI 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 States Nuclear Regulatory Commission. Toward this end, Petitioner also requests formal rulemaking proMngs under 5 U.S.C. 553(c) to examine the issues : addressed herein. ,
d Thank you.
Sincerely, wp ^W 4
George Crocker, Steering Committee Prairie Island Coalition Y
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BEFORE THE UNITED STATES NUCLEAR REGULATORY CCHMISSION 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:
- 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 violated 10 CFR $72.122(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 theie casks cannot be placed back into the pool to perform thue functions due to unresolved problems with fuel degradation during storage, flash steam, thermal shock, and the resulting potential for radiation dispersion.
D ng 339. ...d 2 p;
< t o S. 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 ur. loading procedures, while such '
placement will premacurely 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 $72.122 (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 $72.130 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 ISTSI is permanently decommissioned. This failure may-prevent decommissioning. .
- 6. Determine that NSP violated 10 CFR $72.11 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 $72.12 by deliberately 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 decnnmissioning, 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.
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- 43. 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 rules for prospective licensing proceec'ings 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 (hereinaf ter "PIC")
incorporates herein by reference the facts, argument, and 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 Installatiun (hereinafter "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 relationship 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 hnd waste management problems, including cask unloading problems. PIC participated in the NRC public meeting in Red Wing, 1W regarding NSP and Transnuclear TN-40 cask fabrication quality control problems. PIC petitioned for intervenor status in NSP's licensing proceeding before
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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 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 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 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 the 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 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 develcped, ,and there has been no assessment or evaluation regarding the actual ability of such measures to adequately protect worker and pubic health, and the environ = cat. Safety margin references may also be assumed to refer to the question of whether the disintegrated fuel could be physically unloaded at all.
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. . 8. Also in this lotter, Ms. Marcus reaffirms that SARs 'over-
. simplify matters" when they state that unloading is basically the reverse of loading, because such stataments do not reflect that the unloading process introduces different conditions and complications compared to the unloading 0 process.
- 9. In a letter dated July 10, 1997, from Beth A. Wetzel of the NRC staff to NSP, Ms. Wetzel requests additional information legarding thelPI 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 ar.d 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 e 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 troubling.
- 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.
- 11. On April 16, 1997, Jack W. Roe of the NRC sent an internal memo to another Staff member defining NRC's dry cask scorage 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 nean 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 gu-s-..w &,. en,-em ,.e_ . . ,.44,[ .. , . - , , -
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 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 pool to unload immediately, i e., at Prairie Island, or .
vhere a spent fuel cask has weld flaws, i.e., Palisades, where welds have failed. Mr. Roe did not address the issue or assurance that the utility can in fact unload the casks.
- 13. There are other reasons te unload a cask that have not been '
addressed in Mr'. Roe's letter. The NRC has clearly stated th'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 delayed 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 welds.
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- 14. .Another reason casks must be placed into the pool and opened is obligntory cask maintenance which must be completed on Transnuclear's TN-40 cask. Exhibit G, NSP SAR for Prairie Island ISTSI, 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 unloaded.
- 15. '
Another eenson the TN-40 casks at Prairie Island would require unloading is that state law requires that they be moved off of Prairie Isl.ind. This state requirement anticipates that the casks must be moved after a term of tamporary storags, 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 e
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 Derision 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 enticipation of the necessity of unloadlag is inadequate.
- 16. Yet another scenario where unloading is required is for decommissioning. NRC authority rests on the requireme'.t 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 ISTSI...is permanently decommissioned." Because there are unaddressed unloading issues suen 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 4
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- 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 howledge 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. T, 4/15/97 NRC's Susan Trant Shankman Letter to Sierra Nuclear. . In that particular case, the letter writer is concerned with degradation due to escape of helium, and amphasizes thatt sinoe one of the design requirements of the omsk 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 anticipated event, henca 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 cause is not addressed, nor is the method by whi,ch 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 problams, 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 importknt 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 3 canisters ,had been stored for several years, and the thermal damage was so severe that the canister could not be
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In the October 19,1995 letter, the writer unloaded.
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[T]he canisters had " setup"-in some fashions thermally, twisting, bowing, corrosion or other..."
The canisters had apparently taken on a set most probably thermally induced although possibly including twisting or other. The other laminated factors suchofasthe makeup bowing, 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 damages (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, although not probable, that additional canisters may be unremovable, it is also possible that canister number 18 is 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 ISTSI 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 .
Beth A. Metzel'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.
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- 22. .NSP consistently claims that casks can be unloaded,.and that l " 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, and in its legal argument. Exhibit J, In the Matter of a Request by Northern States Power Company for Certifice.21on of Compliance, Cl-96-2189, C8-96-2190, Respondent's Response, p. 5-61 Aff. of Jon Kapitz, p.2. In Mr. Kapitz's Affidavit, he first states that i The unloading procedure and the relevant design features for tae TN-40 casks approved for use at the PI '
i
- Plant are based upon features and procedures common to
[ existing Transnuclear casks used worldwide, including l shipping casks and storage casks like the TN-24P cask.
Exhibit J, Aff. of Kapitz, p. 2 (emphasis added) . He goes I
- on to say that
While NSP has not needed to unload any of the five TN-
- 40 casks that have been loaded at ths PI plant to date,'
l a comparable Transnuclear storage cask (a TN-24P cask) j has been successfu.jly unloaded as part of a pro $ect t
jointly sponsored by the Electric Power Research j _ Institute and the United States Department of Energy.
i l Id. (amphasis added). Although it is accepted practice to
- attach to an Affidavit any source used as the basis for that
- Affidavit, Mr. Kapitz did not do so! Mr. Kapitz did not i
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 l . quick review of the study provides a reason it may not have l been included with Mr. Kapitz' Affidavit. Exhibit X, EPRI,
! "The TN-24P PWR Spent-Fbel Storage Cask: Testing and l 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 l 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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- NSP's claims that the TN-40 casks can be unloaded are
. baseless. .
- 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 N5P cite and produce that demonstrates that a TN-40 cask can be unloaded in a pool?
-Conclusions '
NSF has violated 10 CR 72.122 (f) because it cannot maintain casks. 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(h) because the fuel is subject to degradation /n 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 regarding this issue.
NSP NA5 vinisted 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 18FSI 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 15FSI 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 hnloading 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 regarding maintenance and unloading of TN-40 casks in their ISFSI application and in subsequent submissions regarding cask
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. . c ua maintenance and unloading issues through its continual insistence that.it can unload TH-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, rurther, 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 casks that no cask used for long term storage has ever been unloaded in a pools 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
. ISTSI on Prairie Island and at .
'any other sites where the issues raised above remain unre;..,ed. 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 ISTSI operations on Prairie Island must be suspended.
During the term of suspension, no further casks shall be filled at the Prairie Island site.
Dated: M'4'I !
4.7 , , . .
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umiso stAras Exhibit A
] '
NUCLEAR REGULATCRY COMMISSIDN k *(
wasennetow, s.o. mean.em February 25, 1997 George Crocker, Steering Comeittee -
Prairie Island Coalition P.O. Box 174 Lake Elmo, MN 88042 6
Dear Mr. Crocker:
As the lead manager for dry cask issues in the Office of Nuclear Reactor '
Regulation (NRR), Nuclear Regulatory Commission (NRC). I am responding to your letter dated January 14, 1997, to Charles Haughney.
The safety analysis report (5Ait) for the independent spent fuel storage -
installation 15F51) at the Prairie Island Nuclear Generating Plant provides various estima(tes of radiation exposure associated with t.he 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 O'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 Raoulations 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 !$F5I includes an analysis of a hypothetical loss of confinement barrier which assumes the total inventory of radioactive gassi, within a cask are released. This hypothetical scenario results in a maximum individual whole body dose of 0.15 ren for a member of '
the public. Any credible accident involving a dry storage cask at Prairie
' Island would result in less exposure to the general public than does this hypothetical scenario. The possible eneration of steam during the refilling ,
of a storage cask would not >e a sign.ficant 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 : art of its assessments of licensees' procedures for unloading dry storage cas ts, the NRC staff considers the dry-run exercises perfweed to verify key aspects of unloading procedures, as well as licensees' actual experience in the loading and unloading of transportation caskt, loading of storage casks, handling of spent fuel assemblies under various conditions, and >crforming various activities associated with reactor facilities. :n the a ssence of L:tual experience in unloading spent fuel from a cask fo' lowing a long period W= =- = - - --
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S. Crocker - t-of storace. a ceneral understand < na of technical caoabilitier and related experiences enablos the NRC staf d to assess the adiauncy of a 'icensee's nrocedures for un' oad<no dry storaes casks. For taose examp'es of cask unloadings mentioned < n 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 cc .
early in the unloading process. In the case of Prairie Island, the licens e unloading procedure (Enclosure 1) requires personnel to determine if '
additional steps or precautions are warranted based on the analysis of the gt.
sample from the cask cavity. Additional surveys and samples are taken throughout the unloading process to ensure that the radiation doses received 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 t. ave not
involved the temperature differences between fuel and coolant that may occur i if a cask was unloaded after a period of storage, engineering evaluations and ,
experiences with transportation casks have shown that "themal shocking" is
,unlikely to cause operational safety problems.
j
^
Cask unloading would be expected to involve reflooding and opening the cask and withdrawing the fuel assemblies in a manner similar to nomal fuel handling practices. In the unlikely event that fuel degradation has occurreo during storage, the unloadina may reouire additional fi terina aid even vacuumino debris from the bottom od the cask. Such steps would se 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 pe11nts, 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 #v storage cask, the accumulated occupational dose to perform this activity aay be increased from the previously mentioned estimates. Fuel reactivity for critical < tv ,
. considerations could increase only under very ideal < stic and hiah' y un' kel, <
41sintecration oatterns in the fue' .
Upon dotection that fuel disintecratit ind occurred. special measures wou'd be deve oned and imolemented to assure >
adeocate safety margin is maintained durino unload no.
Some SARs do state that unloading is basically the reverse of loading and thi; statement, in a general sense, is true. However, such statements may tend cc 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 storace and related statement.,
made bv the NRC staff, includina those by Mr. Kualer the staff was emchasizing that l'censees need to identify the conditions and comolicati n
==
, w .6 h e ens .MMg M e-f]>Mr 'Demus-N%4eE EW N0-
.4, -
.S. Crocker '
that are associated with the unloadina nrocess and ensure that unloadina nrocedures address those concerns. The unloading procedure for the dry storage casks at Prairie Island was inspected by tte NRC staff and, following minor revisions, was found to provide adequate guidance to control the process.
unloadi]002;72-10/95002A 50-306/9 copy ofasMRC is provided EnclosureInspection 2 Report 50-282/95002; 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, *
-Sail 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:
Asstated(2) ccw/ enc 1: The Honorable Loren G. Jennings Minnesota House of Representatives Box 17 Rush City, Mt 55069 cc w/o enc 1: see attached page
Exhibit a
+f.
l NUCLEAR REGULATORY CIMMitalCN me . ;, o_
,,,,,4 July 10,1997 Mr. Roger 0. Anderson, Director Ucenelng and Management issues .
Northom States Power Company 414 Nicollet Mall .
Minneapolls, Minnesota 55401
SUBJECT:
RE6UEST POR ADDIT 10NAL INFORMATION ON THE PRAIRIE.
ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, AMENDMENT OF SPENT PUEL POOL SPECIAL VENTILATION -
TICHNICAL SPECIFICATION (TAC N08. M94782 AND M98753) .
4
Dear 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
- Speelfications pertaining to the spent fuel pool spoolal ventilation system. In order to review the proposed changes the staff requires some additionalinformation. Our .
request for additional information (RAI) is enclosed, in order to continue our review of your submittsl on an expedited basis, please [
provide your response to the staff's RAI as soon as practical. If you have any questions regaiding the content of the RAl, please contact me at (301) 41F-1355. ,
sincerely, h.
Both A.Wetzel, Pro et Meneger '
Project Directorate ill-l
, Divialon of Reactor Projects lil/IV Office of Nuclear Reactor Regulation Docket Nos. 50 282. 50 306
Enclosure:
As stated . ,
oc w/ encl: See next page e
e 0
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MEQUEST FOR ADDITIONAL INFORMATION FOH REVIEW OF
, THE AMENDMENT OF THE SPENT FUEL POOL SPECIAL VENTILATION ZONE TECHNICAL SPECIFICATIONS
- 1. Step 8.27 of D95.2, 'TN40 Cask Unloading Procedure
- directs the cook to be filled with water. The caution prior to step 8.27 reeds, 'The wator/steem mixture from the vent pon hose may conteln some redloactive gas. The eres directly above where the hose le discharging shall be closely '
monitored to determine if there is a redlological hazard.' le the opent fuel pool special ventilstlon system operable during the performerce of this step of the unloading procedure? If the spent fuel pool special ventilation system la inoperable during this stop and other portions of the unloading procedure because the overhood crane le supporting the cask through the open spent fuel pool enclosure slot doors, discuss why an inoperable ventilation system does not pose a redlological horard and give any precautions and protections that ensure that tO CFM Part 20 and Part 100 requirements are not exceeded.
- 2. Secilon 5.5 of the Pralrle Island 18FSI (independent spent fuel storego Installation) safety analysle report (SAM) states in pad, "After moving the cask into the fuel pool ares, the cavity will be depressurized and the cask lowered into tlw spent fuel pool." However, Step 8.4 of procedure D95.2 directs the cask to be depressurized while It la still located In the rail bey ares. 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 depressurisation evolution?
- 3. When the spent fuel cask is filled with water prior to unloading the fuel (per Step 8.27 of 095.2, "TN40 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 het fuel elements, if any fuel cracking la predicted, list the expected redlonuclidee sad quantkles 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 venting, describe how you plan to detect and prevent these radioactive gases from being released into the environment.
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j' */ *. '._ \a ano man, LH4ffE*) STATES ( wuct. An R 00L.ATORY.COMMl SloN WASNINefeN. 5.0. EugeMIM .
\e.... April 16, 1997 NENDRANDUM T0: Cynthia D. Pederson, Director Division of Nuclear Materials safety Reglen 111 -
b.L i .s a FRON: Jack W. Roe Director ( - Division o eactorProjects!!!/IV - Office of leer Reactor Regulation SUWECT: TASK INTERFACf AGREEMENT 96-0440; DEFININGDRYCASKSTdPAtit TERM 5(TACN05.M9734sANDM97347)
~
In response to your request dated November 24 1994, NRA have discussed the questions raised and offer,the follow /DRPi and NM58/5FP0 ing clarifications regarding the .tems ready retrieval and structural defects. The two basic masons to return a cask to the spent fuel pool and unload the spent fuel assemblias are either to (1 retrieve the fuel assemblies for furt5er processing or dissonal, or haspotentiallydegradedthedesign(2))respundtoaneventorcondit requirements established for the cask. ' The staffAq possura hasbenot identidia.1 seen withn athe un'endtna specifted timeofinaortfor cask as a meute.4 orotective to lim' .
- t the offs' te rynsecuences a sT, orate cask.
or an accicant invaivina tne release of racioacMWERTPTI1 from - In regard to the reaute -.- +kat cank desions must _ allow retrieval of the stent fuel for further p ecess4to or disposal (29 crn n.4r4 sne nm has cins'stenriy T,anen a po ation 4)at licensees can satisfy this r}e}q,uirement without maintainine the capability to retrieve sne sp nw iusi i,v. . ...k w'lthtn a s assir' ea nortoo of ume, ano say, ir nosassary, neveiop sisemate - er uun. re- fuel retrieval rr a cask umoncing 6.nnus us i-um .i1 . ,,.. ied dkw so a snoruo. or snacT"IT*1 scant rusi nooi. inis is sensicerec asseptanle because licensues have a great deal of flexibility in their ability J. to schedule and plan for the transfer of spent fuel from a storage cask to another cask for storage or shipment. . E Several of the actions required by IsFS! technical specifications or cask certificates of compliance specif,y that, in the case of certain events or . conditions, a cask may need to be unloaded, or otherwise returned to a safe storaea condition. The NEC 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 bility to unload a cask within a specified time. This position is based on - he absence of an identified event or condition involving the stnrage cysks - hat would result in an immediate threat to pubite 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 accosanodate CONTACT: William Reckley, NRR + - (301) 415-1314 p APR 1L I
- Q Q y w i T ' 3177 - - --~ ~
~
1
/ * . *. .
')
C. poderson -I-unloading of.a cask, and (t) several licensees sharing a single cask transport vehicle between different reactor sites. In the specific case of the Prairie Island !$F51, the NRC staff, in its safety Evaluation Report dated July 19g3, stated that its review of the accident analyses determined that, " Dose equivalent consequences, fm a single cask, toanyindividual.fromdirect.orindirectradiationandgaseousactivity) are less than the 50 e5v (5 ren release efter postulated accident limitestabitsnedin10CFR72.10s(b)." Assessment, dated July 28,1992 the staff assessed t events,A:'dtttonallyhe in its Environmental e accide siteboundar{ThedosesarealsomuchlessthantheProtectiveActionGuidesas, found that. "... established by the Environmental Protection Agency (EPA) for individuals exposed to radiation as a nsult of accidents. Because it has been shown that the dose equivalent to any individual from postulated accidents inyc1ving a single cask is below levels nquired for taking protective actions to protect a TN-40 public cask ishealth,ly a high unlikely event.the NRC staff However, considers following certainthat a time-ergent events or unloading the licensee is requimd to take cortective actions to ensure safe conditions,ditions ttorage con and to perfom inspections to ensure a cask continues to meet applicable design requirements. This may include returning a cask to the Auxiliary But1 ding and/or the spent fuel pool. However, once tis cask is in the spent fuel pool it- - safe storage condit\ons.does p t have to be unloaded immediately to maintainThe licensee wou options, required precautions, and other special considerations that may be involved in the required unloeding 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 claddine integrity during storage leads to restrictions on the fuel assemblies that can be initially loaded into the casks. Acceptance criteria for fuel assembides to be stond pertain to heat.
generation rates, initta) 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 ISFS! 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
'T$ 3.1.1. 6 - Fuel assemblies known or suspected to have structural defects or gros (s ) cladding failums (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.
M,. ,
*+ .s * }e . ,, ' 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 form. In the specific case of Prairie Island, neither 10 CFR 7212(1) or specific
!$F5! technical specifications introduce additional requirements for the fuel handlin,1 equipment used to actually load or unload the fuel assemblies into the cask since such matte regulationsandlicenses.psareaddressedunderexisting10CFRPart60 The structural nquirweents defined by the ISFS! '
3AR and technical specification are satisfied even if it is necessary to use a special handling tool to overcome problems in lifting selected fuel assemblies provided that these assemblies do not have gross cladding failures and will olhemise maintain fuel assembly geometries assumed in the design-basis analyses performed for the cask. The adequacy of the licensu's actions should be judged in the context of the regulations in 10 CFR Part 50 and the associated nactor facility operating license. If the licensu's actions are reasonable for the handling of fuel within the spent fuel pool, those same actions the structural can beintegrity creditedrequirements in the determination of the 15F5! of whether the Itcensee technical satisfies s >ecification and fuel retrievability requirement of 30 CFR 72.122(1). If, on tw other hand, the licensee's cornctive actions are deemed inadequate or the special fuel handline procedure increases the probability of a fuel handling accident g within the reactor facility, actions or inquiries from the NRC staff should be presented in the context of regulations such as Appendix B to 10 CFR 50 or 10 CFR 60.59. The NRC Office of the Seneral Counsel has reviewed this response and nas no legal objections. . Please contact W111tna Reckley of eiy staff at (3013 415-1314 if you have ,any additional questions or concerns regarding this manter. C. Hehl, RI. cc(w/ incoming): , . B. Mallett R!! ' R. Scarano,, RIV Prairie Island !$F51 Technical specification 1.3.2,
- Fuel and Cask Handling Activities," states: -
Fuel and cask novament and 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 OPR-42 and OPR-60 and associated tuhntral specifications. . .
- NR'C Inspection Report - Sierra Nuclear Corp. .
l- . 6 Saladed Raports index l Nanandlnfonnation ] NRC Home Page l E-gail April 15,1997 Mr. Art J.McSherry President Sierra Nuclear Corporanon One Victor Square ' Scotts Valley, CA 95066
SUBJECT:
NRC INSPECTION REPORT NO. 72-1007/97 204 AND NOTICE OF NONCONFORMANCE Deer Mr. McSheavy: This letter refers to the inspection cor ducted March 17 21,1997, at your theility in Scotts Valley, California, and at two of your fhbrication contra tors' fhcilities: March Metalfab. Inc., in Hayward, California; and Nor Cal Metal Fabricators, in Oaklanit, California. The team examined information about seal weld Bilures on dry spent fbel storage casks at the l'alisades and Arkansas Nuclear One (ANO) nuclear power pla'nts. At'.litionauy, the team ==ad het adquacy of your cotractive actions taken for the nndings identi6ed in Inspection . Reports 72-1007/96 204 and 96 208, regarding the Model VSC 24 dry spent ibel storage system manufactured under Certi6cate of Cnmpliance No. 72 1007. The enclosed report (Enclosure 1) presents the results of otw inspection. The team held an exit meeting with you in the Sierra Nuclear Corporation of5ces on March 21,1997. During the laspection, the team found tiat you fhiled ta meet coitain Nuclear Regulvory Commission requirements. The team identified four t onconformances regarding failures to perform work in accordance with your Quality Assurance Program The nonconformances 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 Cati 6cate of Compliance to correct a nonconservative requirement for the drain <iown time limit for a loaded cask, (3) , submit a Safety Analysis Ret
- change to correct the 1986 American Society of Mechanical Engineers Code omi 'on of nondestructive examinatiau requirements for temportry attachment:, and (4) control measuring test equipment.
Two of the nonconformances raise safety concens. First, the shield-lid weld failures affect the integrity of a f cask conhnement boundary. The root-cause of the shield-lid failures and the potential for delayed crt.cking rn j loaded casks must be understood. Ahhough the hilure of both the cask's inner shield-lid seal weld and outw I structural-lid weld would not pose an off-site threat to public heehh and safety, such an occurrence wotdd cause the loss of the helium atmosphere insid.: the cask. This loss could result in cladding degradation and i fbture ibel handline and retrievability problems. Since one of the design requirements of the cask is the king term protection of the fbel 444!ag (10 CFR 122(h)), such degradation would be unacceptable. Second, ibe no'monsavative Technical Speci6c:4 ion for cask drain down time sfrects the margin to criticality, k w h ljfjg*--?-)T - --
hm2 . Sierra Nuclear Corporation's lack c f timely and comprehensive action, in dealing with these imponant safety , issusa, is a signi8 cant regulatory concern. As the certi6cate holder, Sierra Nuclear Corporation la responsible for the adequacy of the design ofits fbel 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 ihnher. This meeting is open for public observation. At the meeting you should be prepared to discuss your shon term and longer term corrective actions to address the issues and concerns raised by our inspection. Please provide us, within 30 days &om the date of this letter, a written statement in accordance with the instructions specified in the attached Notice of Nonconformance (Enclomre 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 'itulos of Practice,' a copy of this letter, its enclosures, and your response will be placed in the NRC Public nam ==* Room (PDR). . Sincerely,
/dgned/
Susan Frant Shankman, Chief Transponation Safety and Inspection Branch Spent Fuel Project OfBce, NMSS
Enclosures:
+ 1 1,: ,
- 1. Inspection Report 72-1007/97-204
- 2. Notice of Nonconformance Docket No. 721007 4
e e OO
-m. wa.m> s as. -
e, - - - - - - - .; -- -- - - - . Confirmatsry Action Letter- Arkansas Nuclear Selected Reportdada l News and Information l NRC Home Page l E-giail May 16,1997 CAL No. 97 7 002 Mr. C. Randy HuteManaa Vice President, Operations ANO . Entergy Operations,Inc. 1448 S. R. 333 RusseUville, AR72801
SUBJECT:
CONFIRMATORY ACTION LETTER Dear Mr.Huichinson; During the week of March 17,1997, U.S. Nuclear Regulatory Commission staffinspected Sierra Nur: lear Corporation (SNC) and two of ha 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 welds joining the cask shield lid to the multi-assembly stgrage basket (MSB). The Palisades welding dh4 occurred in March 1995 and the first !a=we at ANO in December 1996. AAer the recent SNC insW problems arose while welding another ANO cask on March 26,1997. NRC is concemed about the difficulties e,nspyttered with the weldsjoining the shield lid to the MSB since this weld is part of the confinement
' boundary o7% VSC 2Ourtherm6re~, thiwe}il.betiveen%e, MSB and the iructuiallid'may be suscepdble to the same failbre
~
m'ochiInisms'gE1h hi. tid.jd weld. It is possible that these Dart cular weld Drobleri i may not devefoo un'ill after cask welds have underaone non-deumetwe examiWatron.11though'such welc. :adures
- -ivould not pose an off site threat to pubbc health and safety, s_uch an occurrence would cause the loss of the 1
Man strnosphere inside,,the MSB.
- i
/% and retrievability pro 6lems_.Q!,s,cpaditimLgpul,d r.e.,suh,@,$tel claddina.dg 4
i . The March 1997 inspection revealed that neither SNC nor the user licensees had perfonned a =-y-Mvs root-cause analysis of the 6rst two weld problems. An understanding of the root cause is essential to preventing recunence when welding future caska, and to assessing the possibility of additional weld - problems, perhaps undetected or delayed, in loaded casks. On May 6,1997 NRC held a public maating with . SNC representatives to discuss SNC's implemented and planned actions in response to the weld problems and . inspection 6ndings. Repranamatives of your staff attended thL meeting. As stated at this meeting, the staff ramalna 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 Heghney, Deputy Director of the Spent Fuel Project OfBce, OfBoe 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:- 4
+~ + -
. . .. .. . - . . . ._.m. ...
Pese 2
- 1) Detsmine that your welding and 'nspecti:n practices prcvide reasonable assurance that cracking, ,'
kp ,(including possible undetected or delauxi cracking, will not structural tid to the Mss. If necessary, moddy your welding processes to inhibit rocurrence of these welding problems, (2) On completion of this action, and at least 14 days before loading anott er VSC 24 cask with spent Ibel, you will submit to the Director, OfHee of Nuclear Material Safety and Safeguards, a written description of any procedural or design modifications made with respect to item 1. The submittal should include the technical justification for each modification. A copy of the submittal should be sent to William F. Kane, Director, Spent Fuel Projecs OfBee, and to your Regional Administrator, (You may include in this response the information required by item 2 below, confhming all the actions required by item I above.) Pursuant to Section 182 of the Atomic Energy Act,42 U.S.C. 2232, you are required to: (1) Notify me immediately ifyour under-Mag differs hom that set forth above; (2) Notify me in writing when you have completed the actions addressed in this Confirmatory Action Letter, Issuance of this Confirmatory Action Letter does not preclude issuance of an order formalizing the above
- ommitments or requiring other licensee actions; nor does it preclude NRC hom taking enforcement action f;r violations 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 Document 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, ifyou find it necessary to include such information, you should clearly indicate the specific information that youdesire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information &om the public. Sincerely, Malcolm R. Knapp, Deputy Director OfBee of Nuclear Material Safety - and Safeguards Dockets 721007,7213,50 313,50 368 4 da.., _ - - - . ,w- . . , _ -
ISPSI SAR TABLE 5.1 2 ANTICIPATED TIME AND PERSONNEL REQUIRD(ENTS FOR CASK HANDLING OPERATIONS Onoration No. of Time Avg. Distance Zgtsonnel fain), fft) from cask Receiving
- 1. Unioading (A1) * * *
- 2. Inspection (A2 through A7) * * * ,
- 3. Transfer to cask loading pool (AS) * *
- Cask Imac}ine Pool 4 Iower eask into pool (31) * * *
- 5. Load fuel (82 through 84) 5 * . *
- 6. Place lid on eask (35) 5 * *
- 7. Lift cask to pool surface (56) 5 30 5
- 8. Install lid bolts (56) 5 120 3
- 9. Drain cavity (87 through 311) 5 90 6
- 10. Transfers to decone==4 nation area (512) 3 60 10 Decentamination Aret ..
- 11. Decontaminate cask (C1, C2) 3 120 3
- 12. Remove vent plugs 2 30 5
- 13. Drying, evacating, backfilling (C3 through C13) 2 480 5
- 14. Install top neutron shield C14) 2 -
15 3
- 15. Install pressure .
transducers (C15 through C17),2 30 5
- 16. Pressurize interspace (C18) * * *
- 17. Check leakage (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. Land on transport vehicle (C23) 3 60 5
- 22. Transfer to storage area C24) 3- 60 10 TABLE 5.1 2 REV. 2 9/91
--......7_.- - _ _ _ _ . . . 4
ISFSI SAR TABLE 5.1 2 (Continued) ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS Deeration No. of Tina Avg. Distance Personnel IE1!d. (ft) from Cask Storag @
- 23. Unicad from vehicle ,
position in location (D1 D2, D3) 5 60 5 24 Check surface dose rate (D6) 5 30 3
- 25. Connect pressure instrumentathu (D4. DS) 5 30 5 ZgIlgdie 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
_ m ,1 w _- - - . __ 3 1950** 8 P
- No measurable dose associated with this activity. Therefore, the number of personnel, tima and distance are not significant.
# Farenthetical infor sation corresponds to Table 5.11 activity numbers.
** Total time to transfer cask c 6 replace lid seals, and return cask to ISTSI pad.
s 4 TABLE 5.1 2 REV. 2 9/91 l 1
. 18FSI AAR TABL2 5.1 2 (continued)
ANTICIPATED TIME AND FIASONNEL REQUIEDENTS FOR CASK HANDLING OPERATIONS pnaration No. of Time Avg. Distance Faraonnel lain). ffe) from cask Eterana Area
- 23. Uniosd from 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) 3 30 5 Periodic Maintenance
- 1. Visual surveillance (NA) 2 15 5
- 2. Repair surface defects (NA) 2 60 3
- 3. Instrument testing and calibration (KA) 2 180 5 4 Instrument repair (NA) 2 60 3
/ Maior Maintenanea (once in 20 years)
- 1. Replace cask lid seals 3 1950** 8 V
- No measurable dose associated with this activity. Ttsrefore, the assaber of personnel, time and c'.istance are not significant.
e, Parenthetic M information corresponds to Table 5.1 1 activity numbers, d
** Total time to trensfer cask to spent fuel pool, replace lid seals, and return cask to IEFSI pad.
e TABLE 5.1 2 Erv. 2 9/91 e
- A+- _
--W N MD *O
.* * ** *'S,% UNITso STATES o. C
\ NUCLEAR CECULATORY COMM'0540N neesoN O *a L f not wAnnewes noAo -
uets,u.unois ensan-4ast (,o.gg ... - June 30,1995 . Mr. E. Watzl, Vice President Nuclear Generation Northern States power Company , 414 Nicollet Mall - Minneapolis, MN 55401
Dear Mr. Wat:
12 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 residtat inspectors, selected R!ll based
- inspectors, and technical staff from the Office of Nuclear Reactor Regulation and the Office of Nuclear Materials Safety 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 nrooperational testino activities. '
We discussed the results of this insnetton with you and other meneers of your staff at a public exit meeting on April 28, 1995. At that meeting we identified five items that required further resolution. You prov<ded us with additional infomation for each of these items and we completed our review of
' the subject items during the next two weeks. On May ll, the NRC issued a a allowing you schedular to subatt the exemption results of your frompreoperational the requiraments of 10 test 1ne CFR than 30 Part a 72.82(d)ys 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 re> ort ids titles areas examined during the inspection. Within these areas, tie 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 safely load spent fuel into the TN-40 dry storage cask and transport this cask to the onsite ISFSI. '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 bf 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 this violation had been corrected and those actions were reviewed during this inspectinn. Therefore, no response with respect to this violation is rsquired. However, we are disappointed that NRC inspectors, rather than your own staff, identified these procedural deficiencies. - c.g. A.w ..m . . . . - - . - . m om e1NL _ ___ __ _ . _ .
L Watz) . We also identified several weaknasses with your overall performance relative to dry cask storage activities. These weaknesses included: 1) poor oversight of vendor activities until late in the dry cask storage project; 2) lack of effective engineering involvement in vendor fabrication activities; 3) the ineffectiveness of your quality assurance organization in assessing vendor performance during the cask fabrication process; 4) the absence of a comprehensive plan fot inspecting, audit?ng, and monitoring dry cask storage activities onsite, particularly those activities 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 you plan to implement to address the above weaknesses in dry cask .. activities. Please respond to thir request within 30 days of the date of this inspection report. We will continue to evaluate the effectiveness of your corrective actions to improve your performance in dry cask activities during future NRC inspections. s : } In accordance with 10 CFR 2.790 of the MRC's ' Rules of Practice," a copy of this letter, the enclosure, and your response to this letter will be placed in the NRC Public Documer.t Cson. The response requested by th4 Wr is r:ot subject to the clearance . procedures of the Office of Puaqmnt and Budget as required by the Paperwork Reduction Act of 1980 PL 96-M1 We will gladly discuss any questions you have concerning this inspection. Sincerely,
&A. A=4 Eoward G. Greenaan -
Senior Oversight Manager Region III Dry Cask Activities Docket No. 50-282' ' Oocket No. 50-306 Docket No. 72-10
Enclosures:
- 1. Notice of Violation
. 2. Inspection Report No. 50-282/95002; 50-306/95002; 72-10/95002(DRP)
~
See Attached Distribution ' d
' \
.I l
NOTICE'0F VIOLATION Northern' States Power Company Dockets No. 50-282; 50-306; 72-10 l Prairie Island Nuclear Plant Licenses No. DPR-42; DPR-60; SNM-0506 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 lis,ted below: 10 CFR Part 72.142 a quality assurance (b) re r storage Installation program ISFSI (Q(A) with regard quires a licenses to an Independent Spent to establish, Fuel maintain, Subpart 6, " Quality Assuranc)e."that satisfies each of the applicable criteria of . In meeting the Part 72.142 b requirement 10 CFR Part 72.142(d accepts a Commission-approved quality (a)ssurance progr,am
- I which satisfies the a)pplicable criteria of Appendix 8 to 10AsCFR Part 50.
such, the ISFSI safety Analysis Re > ort ' states that the n_reviously annreved Northern States Power QA program w)ich satisfies applicable criteria of 10 CFR Part 50, Appendix B,'will ne applied to activities, structures, systems, and comgper,ts of the ISFsI commensurate wnn snetr taportance to safety, r - 4
. Critet toa ii . ' Appendix 8 to 10 CFR Part 50 requires that activities affecting quality b rescribed by documented instructions, procedures, or drawings, of a type apywpriate to the circumstances and that these activities be
- accomplished in accordance with the associated instructions, procedures, or ,
drawings. quality. Cask handling, loading, and unloading are activities affecting G.Datrary to the above giot ernmnap nv anor,oveoprocedures af tyna = cask handlino.toleadino. iooropriate the and unlanrHan activities we
~
W cumstances_ as evidencec ay therroisowing 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 Procedure, SP 1075, 'TN-40 Fuel Selection and Identification," did not incor> orate the requirement of Technical Specification (TS) 4.1.2, whici states that "before inserting a spent fuel assembly into a cask.... the identity of each fuel assembly shall be independently verified and cocumented."
- 3. Procedure D95.1, 'TN-40 Cask Loading Procedure," s prerequisites section that SP 1077 be performe@pecified-in ays prior tothe loading a cask. However, the TS 4.19-requirement to perTorm a visual N~+ ion of the lifting device (lift beam and extension) ant vertfy operability])
of the device 7 days prior to use, was not identifieo in un.a. inwr= also of was the no procedure lifting device. identifying actions required to verify operability 6 h useum -e - aswam.
.m ..wa e=* *
's . .'.
. Notice of Violation ..
- 4. Procedure D95.1, "TN-40 Cask Loadina Procedure," did not include a step to perform radiation enevave of the cask surrace nerore movino a cask to Ene 13thi, as reautred by TS 4.6.1. --
- 5. Procedure D95.2, "TN-40 Cask Unloading Procedure," did not adeountely address the TS-requirement to t=mnie tne soent ruei cooi "or boron Tancentration witnin four hours of flooding the cask cavi<;y for unicading tne ruei assemoites. .-
- 6. Procedure D95.2, "TN-40 Cask Unloadina Procedur+" did not contain a hold point to ensure worx wouio not continue untti tl1e resuits of the inner ,
cask volume sample had been reviewed. Inis procedural hold point-is importans 69 ensure snat an unplanned and unmonitored release path is - not created while the cask is in the Auxiliary Building. ,
- 7. The ifcensee did not have a procedure for conducting 10 CFR Part 72.48 safety evaluations. * ' -
This is a Severity Level IV Violation (Supplament I) (50-282/95002-61; 50-306/95002-01; 72-10/95002-01(DRP)). With respect to this violation, the inspection showed that steps had been taken to correct the identified violation and to prevent recurrence. Consequently, no reply to the violation is required and we have no further questions regarding this matter. , Dated at Lisle, Illinois - this 30th day of June 1995 O e mo +e-,m<nm-v ..,mm--
.. t .. ~
o' While the inspectors recognized 'that finalizing the loading and unloading procedures was contingent upon completion of the dry run and I the subsequent incorporation of any lessons learned, there were many as sects of the procedures which should have been in ciace oerore Ine dry L tor u ampie, lechnicai apecirication requirements were noi, e factively incorporated into the loading and unloading procedures (paragraph 3.2). In addition, the licennee die not complete rev'ew and I _ approval of the unloading proceaure unti' the cav "o lowing subm'ssion I g or the creeperan onai test raDort. submission of th' s rennrt imoiied that the licensee was renov 16 load a cask with spent fuel and subsequently unload the cask, if necessary. lT e The licensee did not take a disciplined approach to inspecting the fuel designated for cask storage as evidenced by weaknesses identified by the , ins actors during observation of fuel inspection activities (paragr,aph * ' 7.3 . -
- Some weaknesses were noted with the licensee's Jocumented basis for safety evaluation conclusions (paragraph 8.2).
e 3
operational checks of vehicle brakes, lifting equipment, turntables, Jacks, and cask , links.-
- ' 3.1.5 Surveillance Procedure. SP 1075. "TN-40 Fuel Selection and Edentification" The inspectors reviewed SP 1075 and the cask loading procedure, 095.1, 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.
TS 3.1.1(6 required that, " fuel assemblies known or suspected to have , structural) sufficiently severe defects to adversely or gross affect fuel cladding handling and failures (other than pinhole leak transfer capab shall not be loaded into the cask for storage." The licensee originally * - - intended to 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 inspection techniques, the licensee elected to use video recording equipment to perfom the fuel inspection. The inspectors considered this a preferable method for identifying fuel anomalies and ensuring compliance.with TS 3.1.1. The inspectors observed portions of the actual fuel inspection and 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 not 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 assembly shall be independently verified and documented." The inspectors 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 requirements for fuel identification. The failure to incorporate the 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 inspectors reviewed the loading (D95.1) and unloading (D95.2) procedures for technical adequacy and to detemine 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" The original 095.1 procedure specified in the prerequisites section that SP 1077 be performed 30 days prior to loading a cask. However, the Technical g -Specification (TS) 4.19 requirement to perform a visual inspection of the 10 g#~-.-e --..m%,- -*_ w w - em , w s w ,- -e e w - . ~ . . = = - = + - *
; 1 .
lifting 7 daysdevice prior to (liftuse, bean was andnotextension)in identified D95.1.andThere verifywas operability of theindevice no procedure existence identifying actions required to verify operability of the lifting
. device. This issue v.as identified by the inspectors. The inspectors verified that 095.1 had been updated to include the preoperational testing requirements of TS 4.19.
In addition, the original procedure did not include ~a step to perform radiation surveys of the cask surface before moving a cask to the ISFSI, as required by TS 4.6.1 to ensure compliance with TS 3.6.1. The inspectors discussed this issue with the licensee and verified that D95.1 was revised to include specific steps for performing TS-required gamma and neutron dose rate i surveys. - , The failure to incorporate the requirements of TS 4.19 into D95.1, to develop a procedure identifying actions required to verify operability of the lifting - device, and to include a step for performing radiation surveys of the cask ' surface before moving a cask to the ISFSI as required by TS 4.6.1, are ' considered examples 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)).
= 3.2.2 D95.2. "TN-40 Cask Unloadina Procedure' The inspectors identified that the final revised and approved D95.2 unloading procedure did not adequately addre<< the TS-reauirement to sample the scent fuel cool for boren concentration. Specifically, T5 4.2.1.2 required ' verification within four hours of flooding the cask cavity for unloading the -
- fuel assemblies, thai,ethe dirsolved boron concentration in the spent fuel pool-water introduced into the cask cavity was greater than or equal to 1800 ppe.
However, D95.2 required sampling four hours prior to lowerin the cask in the
- pool. The inspectors noted that there may be some time dela between *
. . partially lowering the cask into the spent fuel pool and fil ing the cask. ..
/ /The subject TS requiremer1 .. is important in that it increased the i."
' defense in-depth for ensuring that there was not the potential for an
'. . i inadvertent criticality. The inspectors verified that D95.2 was revised to incorporate the TS requirement.
P The inspectors also verified that the D95.2 procedure contained specific steps for sampling the inner cask atmosphere to verify the integrity of the stored fuel. wan1ri work The insnectors nnt enetinue noted untilthat the095.2 samole did not contain results a hold nao peen ooirt +a ravsawet. Theansure# ) Tnspectors considered this procedural hold W int important to ensure that an %. N unplanned and unmonitored release path would not be created while tM cask was ~ in the Auxiliary Building. The inspectors verified that 095.2 was revised to , incorporate the subject hold point. The inspectors concluded that the final D95.2 procedure contained adequate guidi.t: to ensure that the sampling and other unloading evolutions were perDread in a manner that would maintain exposures to workers as-low-as-reasonably-achievable. The failure of D95.2 to adequately address the TS-requirement for samplina +he spent ruei pool ano to inciuae a noid coint to ensure tne results or the inner cask volume sample naa oeen reviewed before allowing work to proceed. are 11 e
*-we am ese see, e w - = _ , , , , , , ,
***""*O^ * ' ' - '
*4p_mem w+- e
considered examplen'of'a violation of Criterion V of hendix 8 to 10 CFR Part no (50-Z52/95002-0;,; 50-306/95002-01; 72 0/95002-01( m)). . 3.3 Emercenev/Off-Nomal Procedures The Part 72 license remirad tu lhy to develop an abnormal operating ry The inspectors asked the licensee if (e procedure any other emergency (AOP) foq-/ofi-normi rburied cask p vWo.. event.;s were required in addition to alara response procedures and the buried cask A0P. The inspectors reviewed the cask handling procedures to determine if contingency actions for abnormal events had been addressed..o , . . . . , , m ,, , .t.
- The licensee does not have any procedures, in addition to the buried cask, ,
which address off-normal events. The inspectors noted that step 5.0, of - procedure D95.1, "TN-40 Cask Loading Procedure,' stated that, 'Should anything not look right during the performance of this procedure it is imperative that - the issue be resolved prior to proceeding. All those involved in the performance of this procedure SHALL have their questions satirfactorily - - answered prior to having to perform their task." In addition, to this eneral precaution, the inspectors noted that D95.1 contained spacific " hold po nts" at various steps in the procedure which required that the loading evolution be stopped and any abnormal condition evaluated before proceeding. The , inspectors did not have any further concerns with this issue. - 3.4 conclusions The licensee did not complete developrent of the loading and unloading procedures until Ine day following suDmission or tne preoperationai net _ Tenort. suomission or Inis report impitec Inat the licensee was =cady to load a cask with soent ruel. While the inspectors recognized that finalizing these { lirocedures wac contingent upon completion of the preoperational testing evolution or " dry run" and the subsequent incorporation of any lessons learned, there were many aspects of the procedures which should have been in , slace before the dry run. For example, Technical Specification requirements - were not effectively incorporated into the loading and unloading procedures. - Assuming procedural adherence, the final procedures in place for cask handling and loading were adequate to ensure that these evolutions would be conducted safely. -
- t. 0 Audit Reoorts. Source Insnections. and Vander Records 4.1. Audit and Source Insnection Renorts -
The inspectors. reviewed.a sample of the licensee's audit and source inspection reports to determine if there were any issues that could affect the quality of the cask. This review included documentation pertaining to associated audit findings. The inspectors also reviewed several fabrication records to verify compliance with the design basis documents, including applicable industry standards. The following documentation was reviewed: O 12
; .f .
. 6. 6. Instrument Calibrations The inspectors reviewed the licensee's procedures for calibrating cask survey instruments and determined that the procedures were adequate to ensure proper calibrations. The inspectors will observe instrument calibrations and survey techniques during the actual cask loading evolution.
6.7 ISFSI Monitorina The inspectors walked-down the ISFSI facility and ensured that TS-required thermdluminescent dosimeters were in place. . 6.8 Radiation Protection (RP) Practices Durina Preonerational Testina The inspectors observed RP practices during the loading dry run and noted that workers were kept infomed of the radiological conditions and that RP * - ' personnel were prompt and thorough in performing dose rate surveys to monitor changing radiological conditions. The inspectors also considered the ' decontamination techniques used by the RP staff during the dry run adequate to ensure Technical Specification limits for surface contamination of the cask would not be exceeded. 6.g Neutron Shield Perforwance e The NRC issued a violation in NRC Inspection Report 72-0010/g4-212(NMSS) for in' adequate control of special processes pertaining to the neutron shield resin pour during cask fabrication. Specifically, the data record sheet associated with the resin pour procedure indicated that the temperature of the resin mix - before adding the catalyst was 63 degrees Fahrenheit rather than between 68 and 70 degrees as required by the procedure. In response to this violation, the licensee committed to perform a thorough survey of the cask following full load to verify that the integrity of the neutron shield was not affected by the procedure deviation. The inspectors reviewed the licensee's plans for surveying the neutron shield and determined that the survey techniques were adequate to confirm that the neutron shiel'd was performing its design function. -
, 6do Conclusions With the exception of the procedural content problems discussed in paragraph 3.2, the licensee developed and implemented an effective radiological controls program for monitoring cask loading and unloading activities and storage in the ISFSI. Cask handling procedures and associated RWPs ap addressed items such as dosisatry requirements for workers,propriately survey techniques and the use of calibrated instruments, required air sampling, protective clothing requirements, radiation and contamination area postings, and procedural hold points and work stoppage criteria.
7.0 Pre-coerational Testino (Leadina and Unloadine Dev-Run)
'The NRC license for the ISFSI required the licensee to conduct pre-operational testing to demonstrate cask handling capabilities before loading the first 22 O
"** *e- ..
- 4 v- e ---3-- . - - . e u .4 3e% g.w + e
e
. /. '
cask with spent fuel. The inspectors observed-and/or reviewed several - - pre-operational testing activities. These included: cask arrival and receipt inspection; transport vehicle pre-operational testing; cask transport to/from the ISFSI storage pad / Auxiliary Building; cask pressurc monitoring system pre-operational testing; cask vacuum drying, helium backfill, and seal performance testing; fuel inspection; placement of'the cask in the spent fuel . pool and simulated fuel leading; and cask removal from the spent fuel pool subsequent decontamination. LThe removal of the cask lid under water and th' , filling of the cask with water were two avuutions that were not demonstrat by sne iicenses aun na cry run activities.' These exemptions were approve the NRC to prevent any unnecessary damage to the lid seating surface duri g. the dry run and did not. affect the licensee's ability to demonstrate unload no a was perfomed to demonstrate that the transporter and cask would not tip ' over during cask transport should a seismic event occur. However, the subject SE did not address the consequences of a tip-over accident in the Auxiliary , Building rail bay The inspectors discussed this issue with the licensee and with representatives from NMSS. Based on these discussions and-the results of a previous analysis involving the loss of all cask confinement barriers during a spent fuel shipping cask handling accident, the inspectors concluded that if a release of radioactivity occurred due to a tip-over event in the Auxiliary Building, the release would be substantially less than 10 CFR Pt.rt 100 guidelines. Thus,- the inspectors agreed with the licensee's "no' response to'the subject question. However, the documented basis was incomplete in that it did not address the consequences of a cask tip-over event within the Auxiliary
%ilding.
While the inspectors noted some weaknesses with the quality of SE No. 344, the inspectors determined that the licensee's conclusion that operation of an ISFSI would not create an unreviewed safety due to an adverse impact on reactor plant operations, was valid. 7.1 Seal Perfomance Test The inspectors reviewed the licensee's methodology for perfomance testing of the cask seals. The lid sealing system was designed with three sets of double - - 0-rings: one set on the circumference of the main lid and one set on the flange covers for each of the vent and drain ports. The spaces between the 0-rings' for the lid and each flange were interconnected via drilled channels to the overpressure (OP) port. The 0P port was connected to the 0P tank which was designed to apply helium pressure to the volume of space between all of the 0-rings. Should inner-seal._ leakage occur, helium would leak from the 0,P - tank into the cask (because cask aressure was lower than OP tank nrecrure . ~ Should outer-seal leakaarocciTr. selium would leaLfrop_the OP tank _ta..t e Tenvironment. OP tank pressure would be monitored on the storage pad and an alam generated if tank pressure was. low. The pressure monitoring equipment was prepared and tested prior to cask transport. After cask placement, the pressure monitoring system would be installed on the cask and tested via a surveillance procedure. Completion of these activities was documented in D95.1. 23 6
- - - , * * * * - - + - - - - - = * - - - - - - - ~ - * ' -
, 4 L 8.4 OA Overview of Dry Cask Storaae Activities After receipt of the first TN-40 cask at the site, the inspectors determined . that the licensee did not have a comprehensive plan to inspect, audit, or monitor dry cask activities onsite, in particular, those activities that interface with the Part 50 license. The inspectors identified several issues that should have been identified by the licensee. After discussion with the inspectors, the licensee developed an ' Integrated Dry Cask QA Assessment Plan," which provided direction for the Nuclear Quality Department in the inspection, audit, and surveillance of dry cask storage activities. Once established, the licensee'.s quality verification efforts were effective in identifying issues with the dry cask storage pro, ject which required resolution by the line organization.
8.5 Retrievabi],1,ty, On May 3, 1995, the licensee submitted on the docket, correspondence that addressed the ability to unload the first TN-40 cask following completion of the May 1995, Unit 2 refueling outage and prior to receipt of the second cask onsite. The NRC's Office of Nuclear Material Safety and Safeguards responded on May 5,1995 to the licensee and stated that the plans described in the May 3 letter to address unanticipated unloading of a cask before another cask had been loaded, would allow ready retrieval of the spent fuel for further processing or disposal as required by 10 CFR Part 72.122(1). l 8.6 Exit Interview , The inspectors met with the licensee representatives denoted in paragraph 8.7 during the inspection period and at the conclusion of the inspection on April 28, 1995. The inspectors sumarized the scope and results of the inspection, and discusud the likely content of this inspection report. The licensee acknowledged the information and indicated that some of the information disclosed during the inspection could be considered proprietary in nature. 8.7 Egnpns contacted Northern States Power comoany
#E.* Watzl, Vice President Nuclear Generation fM. Wadley, Plant Manager fK. Albrecht, General Superintendent, Engineering G. Lenertz, General Superintendent, Maintenance fD. Schuelke, General Superintendent, Radiation Protection .
and Chemistry J. Sorensen, General Superintendent, Plant operations J. Goldsmith, General Superintendant, Nuclear Generation Services Engineering iT. Amundson, Director, Ganeration Quality Services iP. Kaman, Generation Quality Services
#J. Hill, Manager, Generation Quality Services fJ. Bystrzycki, General Superintendent, Pro,1ect Management 29 .
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' .N2C relects Prairie Island Indians' request to revoke plant's storage :icense try STEPHf.N INCLISH be as simple as reverving the leading L;. _ c ;.INrws process. Kugler poisted out several factors that could make the unloading aback, Adswi.-Ne s is Amasien presen unique, however.
lives eleser in a suelow power plant them " Depending on the nrusties" Kugler the people of the 3sarse taled ladian mid, "the fbst may have been la the ensk Commaity. When the procedme for for decades, and they need 4 evalunm in unicading spent nuclear bel storsse eetidition before miending it." Kagier essks at a Neriharn Sinass power (NSp) mid that pressurization could oesw when plant laceaud less thsa 2,000 feet See the tbs ask is floodsd, produsias steam when commaily was trought into gossties, te water is posed evw het fast He also .. I ^ commaity pendoned tbs Nualeur .. i
. brought up the couern est people ,
Regulatory casumission (NRC) to commest meerty could be rapeesd to adinden e procellural review at the plant. , during aloeding. , Honest Wantra, envirnessanal Jat strasna, spokseman for the NRCl ,f soordiasex for the sonenamity, said abs had received a respoons lemer toen es said that NSP) eurrent procedme for ..' unloading the caska was appewwd is a * ' NRC. "Basecally it said, ' Based se a . . ~! '
. prior review. "We have looked a emir .-
- prelinnaary review, we flad that se procedure far unloading these enska, and .
'amediate sedan is assessary.'" She said we found it to be asseptable," he mid.
- tbs casamunity launds to la the pedalen Kapits admewledged that the S '
fauew is eeune, costant with the leemrt unloading procons won't be as simple as s preasse that "fwther action wiD be alum orirmally asmaned wbsa the plant was Q' withis a reasonable tissa." The commuiry llossaed by tbs NRC in 1993 to more synet - had petitioned to hsw as planet sosisar anslaar fhst "From the oumids looking 38.- 'i wasas seersse lisasse maked pending . unloading is tbs reverse ofleading, his < h ihnbar misw. The plaat sortently asseus
' there att addlacesJ maps we would have '.9.I waste is saves show ground storego .. en take," he said, '*v .
caska, bat could legaDy Silig to 17. . "The issue is whether the equipussi .:,(q- ' At tbs sore of the emasamityk
' we how is adeiguses," Kapier said. "We'te .&
oneenns is the test that the mornse iussks senadant that we han the equipent sad '.'< J.. at the plant han sever besa unleeded. the precedures to properly and safely . .. .'. However, escarding in NSp project unked a ensk. Tk inbe isn't senuy . manager Jan Kapia, "Our ask design and raising any new issues hers; we've been . malanding procedures seine tem throu5h stus before with the NRC, and - . Transouclear (a nuclear waste asorage they told us our procedures were accept. cask manufssturer). While you aus any i able." ', est one of ours has sever basa =la= dad The coeuamity has thee os la side, t .. .' .'..
. for som Kapia said est the earliest abs thousands of Tranaauclear ensks have s ;. . ;
3-been unloaded worldirids. ' * ** ' " plant anticipates unloading any of the * Tbs mioading procedure was casks is five yours is the futurs. Strasma . .- brou5h t inte quazios last year by Andrew said the casks are permined to hold spes . .
*Kul",l Projaet mana6w of dry cask sucisar fust for 20 yeen. "The life of ths ** .
snorsge a the NRC. Until then, the general casks should be significantly loogw tbse 7/l-assumption at the plant and by the NRC 20 years," he mid. "(But) this will had been that unloeding the casks weedd eventually have to be denN with." -
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f ; . Exhibh J ( STATE OF MINNESOTA IN COURT OF APPEALS . In the Matter of a Request By Northem COURT FILE NO. C8-96-2190 States Power Company for Certifloation of Compliance in the Matter of the Northem States Power COURT FILE NO. C196-2189 Company Application for Certificata of Site
- Compatibility for the Goodhue County -
Independent Spent Nuclear Fuel Storage Faciitty AFFIDAVIT OF JON KAPITZ STATE OF MINNESOTA )
) SS.
COUNTY OF HENNEPIN ) JON KAPITZ, being first duly swom on oath, deposes and states as follows:
- 1. I am the Project Manager for Dry Cask Storage for Northem States Power Company ("NSP"). I am making this affidavit based upon matters within my personal
- knowledge. .
- 2. The Prairie island Nuclear Generating Plant (the *PJ, Plant") has a standing procedure for unloading spent fuel storage casks.
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- 7. NSP, together with General Electric participated in a project which involved the successful loading and unloading of. such casks in the mid-1980s in connection with the shipment of 1,054 spent fuel assemblies from the Monticello N' uclear Generating Plant to a General Electnc facility.
- 8. While NSP has not needed to unload any of the five TN-40 casks that have been loaded at the Pl Plant to date.--..--e*'{.iJiWeseF(d IN44Acask has_beerryWig!g!an.i.,JarkeigEihioiothespeapp1071
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- 9. A sixth storage cask is pret,antly sitting in the auxiliary building at the
/p/
Pl Plant waiting to be loaded. If that cask cannot be loaded in the next eight to ten
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1 weaks, it will have to be moved from the auxiliary building to the starage pcd n Pl Plant to permit the space in the auxiliary building to be used for other nece activities. Once loading is permitted, the cask will have to be retumed building for loading. 10. The following costs will be incurred eneh time a cask is moved altht the auxiliary building to the storage pad at from the storage pad to the , 55,200 Heavy Haultruck rental J MO NSP manpower (5 man days @ $250/ day)
- IA.150 TOTA 1. -
11. A seventh storage cask is scheduled to anive at the Pi Plant in late If the cask cannot be placed in the auxiliary building from the railcar de it too will have to be moved to the storage pad and later retumed to the building for loading once loading is permitted.'# gain, the costs ass the cask to atfrom the storage pad _will be as stated in the proceding par 12. While the short-term cask related cost to NSP of a stay of the Minnes Environmental Quality Board decision permithr.g the use of addition the costs of moving casks six and sevan to
- Pi Plant will be appi6T.;Ji 825,800 0.L
' and from tho storag6 pad), NSP's costs will increase dramatically if ca ,
delayed for an extended amount of time. If NSP is unable to load the five that have already been loaded, it ivill be forced to shut down on reactors at the Pl Plant in September,1998, (Unit 2) and would be forced the other reactor (Unit 1) in April,1999. FURTHER YOUR AFFIANT SAITH NOT. 3
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SUDSCRIBED AND SWORN to before me this # # day of 4/J/. . 1996.
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Topics: ' EP Electne Pow;f I Spent fuel storage Therm:My:Iraute monols Exhibh K EPRI NP 5128 Project 2406 4 , PNL 6054 HCt transfer UC45 Researen institute Radiate shiciding interim unn. Report ' n'""4 93 The TN-24P PWR Spent-Fuel , Storage Cask: Testing and g Analyses 4 Prepared by . Pacific Northwest laboratory Virginia Power Company and
- EG&G, Idaho National Engineering Laboratory -
, Mloorn p u e-% - z ,
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REPORT - SU MMARY SUBJECTS Analysis and t steg / Waste management / Fuel cnd core management
. . and development 1DPICS Spent fuel storage Heat transfer Thermal-hydraulic models Radiation shieldmg AUDIENCE Fuels engineers / R&D scientists !
The TN 24P PWR Spent Fuel Storage Cask: Thsting and Analyses [ This test program represents a major milestone in qualifying large metal casks as an alternative for on site storage of spent nuclear fuel, Successful testing cf the TN 24P spent-fuel storage cask has confirmed that it offers a technically sound and practical method for meeting the utility industry's increasing parage needs,
~
BACKGROUND Under a cooperative program sponsored ty DOE, $rginia Power Company, and EPRI, engineers at the Idaho National Engineering Laboratory (INEL) are testing three types of metal dry spent fuel storage casks. The cask described in this report is the TN-24P-a forged steel body cask. EPRI re-port NP das7 presents findings for one of the other casks-the CASTOR. V/21; test results for the third cask, the MC 10, will be published later in 1987, in addition to the INEL tests and as part of the cooperative program, 4 Virginia Power applied for and received a license to store fuel in metal - casks at its Surry nuclear power maten OBJECTIVES Te determme the theimal, ahioiding, and operational perintmance of the TN 24P storage cask and to demonstrate the abihty of computer codes to
.- model the cask system and predict its thermal performance, APPROACH The test program included cask testing and pre and posnest analyses. Ac.
~ tual testing, a: INEL's Test Area Nonh, required shipping of 24 PWR spent- 1 fuel assemblies from Virginia Pomer's Suny nuclear power station,instru-menting the cask and loading it with fuel assemblies, and testing it in honzontal and vertical positions with three different intamal storage ofwi-ronments (nitrogen, helium, and vacuum). The project team, usmg the COBRASFS computer code being developed by Pacific Northnost Labora-tory, predicted the cask's thermal performance and compared their predic-tions with actual test data. They then predicted the cask's posttest perfor-mance, taking into account differences between actual test conditions and protest prediction assumptions.
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RESULTS The tests demonstrated tnat the TN 24P cask is well suited to store spent fuel. The cask showed exceptionally good heat transfer performance, with peak cladding temperatures remaining under the allowable 380*C for he-' lium and nitrogen erwironments with a cask heat load of 20.6 kW. Shielding sPm NP stass 4 c.e , e e ed . e- .e. .me s _, ww.op. g- e g ...g.4 m -- - mesame.n--- d *s'N'**-**""'-"N"'**"
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perbrmartte met design expectatens, with the exceptons of the cask
.bonom and minor peaks at the sdewalls near the top arid bottom of the cask.
At conditions near the cask's thermal design lim'ts, the COBRA SFS code predicted well both the shapes of the temperature profiles and the actual temperatures. Protest predictions agreed within 20T of actual test data; differences were reduced to about 16T in a poonest analysis that corrected for host conduction in the fuel basket arW for assump. tions of heat transfer between the cask and the railcar test fixture. EPRI PERSPECTIVE This testing program quantified the thermej and shielding performances of the TN 24P casa, ft also demonstrated that handling and loading these 10CH containers are relatively straightforward processes, intro-ducing no unusual comands on personnel or facilities. The program achieved remarkably smooth testing despite the number of contractors ' and sites, the complicated Instrumentation, the shipping of major quan. tities of spent nuclear fuel, and the use of complex new computer codes. PROJECT RP24C54 EPRI Protect Manager: Ray W. Lambert Nuclear Power DMalon Contractors: Pacific Noahwest i.aboratory; Virginia Power Company; EGAG, loaho National Engineering Laboratory For further information on EPRI research programs, call EPRI Technical Informaton Specialists (415) 855 2411. 4 9 e.. .. .$ w. 4 :b:, - #AA+ -M&4=*- - - " ' " * * * ~
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' The TN 24P PWR Spent Fuel Storage Cask: Dsting *
, and Analyses -
i I NP 5128 i
- Research Project 24C6 4
- FNL 6054 '
UC-85 , i truenm Report. Aprd 1987
- Prepared ty -
PACIFIC NORTHWEST LABORATORY Banelle Boulevard i Rchland, Washington 99352 ! Pnnopal Irweigmors i J. M. Creer T. E. Mchener M. A. McKinnon J. E. Tanner E. R. Gilbert ! R. L Goodman 2 VIRGINIA POWER COMPANY EG&G. IDAHO NATIONAL ENGINEERING LABORATORY
' Pos Offee Box 26666 550 See,ws Street j Rchmond, Virgine 23261 loaho Falls. :$aho 83415
- Pnnopal Irwenigators Pr. nopal Irweengators
} D A. Dzadosz D. H. Schoonen
. E. V. Moore M. F. Jensen H. S. McKay C. K. Mul6en j . D R Batalo l Prepared for Virpne Power Company -
U, S. Department of Energy . and Elecinc Power Research inantute i 3412 Hailwow Awenue Palo Alto. Califome 94304 i EPRI Protect Manager R. W. Lambert LWR Fuel and Spent Fuel Storage Program Nuclear Power Dwson -f 4 l
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! ABSTRACT ( A performance test of a Transnuclear. Inc. Tbt4P storage cask configured for pres-l surized water reactor (PWR) spent fuel was performed. The test was the second of a j series of cask performance tests p1&nned under a cooperative agreement between ' Virginia Power and the U.S. Department of Energy. The performance test consisted of l he TM 24P cask with 24 PWR spent fuel assemblies from Virginia Power's j Surry reactor. Cask surface and fuel assembly guide tube temperatures were seas-l ured, as were cask surface gaans and neutron dose rates. Testing was performed with ! vacuum, nitrogen, and helium backfill environment in both vertical and horizontal - ! cask orientations. Limited spent fuel integrity data were also obtained. i i . Results of the performance test indicate that the Tb24P cask exhibited exception- } ally good heat transfer performance when dissipating about 21 kW. Maximum measured l assembly guide tube temperatures in vacuum, nitrogen, and helium backfills in a ' vertical / horizontal cask orientation were 278/268'C. 232/247'C. and 214/208'C. respectively. These are significantly less than the 380'C allowable for a total j heat load of 24 kW. Significant convection heat transfer was present in vertical i nitrogen and helium test runs, as indicated by peak temperatures occurring in the l upper regions of the fuel assemblies. Protest temperatbre predictions of the j COBRA-$FS heat transfer computer code were in good agreement (within 25'C) with test j data, and post-test predictions agreed exceptionally well (within 20*C) with data. t . j steasured cask surface gesuna and neutron dose rates were generally less than the , design goal of 60 meem/h. Localized peaks as high as 90 mres/h were measured on the ' { side of the cask. The auximum measured dose rate coincided with an empty bolt hole under the test lid (316 mrom/h) and would be less if the standard lid and protective , cover were in place. The anximum dose rate on the bottom of the cask was 135 arem/h ! . ganna plus 64 mrom/h neutron (199 mrom/h). This indicates the need for additional shielding on the bottom if the cask is stored horizontally. With minor refinements 3 to the shielding design, dose rates can be limited to less than 60 unres/h. i i 111 i j k 3 i 4 ,-, , . _ _ - - -,
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- From both heat traMfer and shielding perspectives, the TN.24P cask cith minor refinements can be ef fectively implemented at reactor sites and central storage facilities for safe storage of spent fuel. '
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CONTENTS Section Pace 1 INTRODUCTION . 1-1 , 2 CONCLUSIONS AND RECOMMENDATIONS 21 Conclusions 2-1 Roccamendations 2-4 3 CASK PERFORMANCE TESTING 3-1 TM-24P Cask and Associated Instrumentation 3-1 Surry FdR Spent Fuel and Associated Instrumentation 3-19 Data Acquisition System 3-44 Data Uncertainty Estimates 3-46 INEL Cask Testing Facility 3-47 Test Plan 3-61 INEL Cask Handling and Operatis.g Experience 3-63 4 ' CASK HEAT TRANSFER AND SHIELDING PERFORMANCE 4-1 Heat Transfer 4-1 Shielding Performance 4-36 5 COBRA-SFS ANALYS15 - 5-1 COBRA-SFS Computer Program 51 COBRA-SFS Models and input 5-6 COBRA-SFS Simulations Co.apared to Test Data
' 5-18 Maxima Cask Heat Load Predictions 5 49 6 REFERENCES 6-1 APPENDIX A FUEL ASSEMBLY DATA '
A-1 APPEND 11 B TEMPERATURE AND PRESSURE tEASUREMENT UNCERTAINTIES 8-1 APPENDIX C HEAT TRANSFER DATA C-1 APPENDIX D DOSE RATE DATA D-1
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- u.w-Figure 3-41. Long-Ters Surveillance Pad and Data Acquisition System Building Location TEST PLA't fne TW2aP cask performance test consisted of the six runs indicated in Table 3-9.
The test runs involved a fully loaded cask (24 spent fuel assemblies), three back- - fill redia (vacuum, nitrogen, ard helium), and two cask orientations (vertical and horizontal). A test. plan specified the order of the runs, the fuel assembly load pattern (see Figure 3-17 in Fuel Assembly Section), instrumentationsmeasurement g locations, calibration requirements, and gas and crud sampling intervals. The test - plan also sortressed cask-handling and fuel assembly characterization activities that were required before, during, and af ter performance testing. 3-61 csh p M 4-W % . gew-MMre-- - - * -
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N. g Figure 3-42. TN-24P Cat,k on Long-Term Surveillance Pad with Adjacent Data Acquisition System Butiding Table 3-9 CASK PERFORMANCE TEST MATRIX I Run ' Cask Backfill Number t Orientation Medium 1 VerticalHeliumD 2 VerticalNitrogen b 3 Vertica1 Vacuum 4 HorizontalHelium D 5 HorizontalNitrogen D , 6 Horizonta1 Vacuum . aAll runs were performed with a fully loaded cask (21 assemblies). The total predicted cask heat load was 28.4 kW at the beginning of the month-long test and 27.5 kW at the end of the test. D Gas samples were taken at the beginning and end of each of these test runs. 3-62
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- Refere Icading the cask with fuel *at INEL, VP inspected each fuel assembly using u
*ltrasonic scanning and video (see Fuel Integrity Section). The assemblies were then Shjpped from Surry to TAN, three at a time, in Transnuclear TN-BL shipping sks.
Upon receipt at TAN, 7*"f '- $Wa v"T6M
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l tthout first being tempcrarily stored in lag storage (see section on INEL cask testing facil-i ty ) . After the eighteenth assembly was loaded in the TN-24P cask, a TC lance was installed in a center assembly and the temperatures were intermittently monitored to ensure safe fuel rod operating temperatures. This procedure was followed until the cask was fully loaden. Once the TN-24P storage cask was fully loaded, the remaining d TC lances were inserted through the primary test lis into pide tubes of selected = fuel assemblies and basket locations. The cask was then moved from the hot shop to the warm shop on a double-wide rail Car. Instrumentation leads were Connected to appropriata sensors, and the tert matrix shown in Table 3-9 was completed. The fuel assembly load pattern used daring testing was previously shown in Fig-ure 3-17 The load pattern maintained 1/8 syerr.:try in the cask, to evaluate temperature and dose rate symmctry and to simplify the analyticsl modeling effort.
, [
The test plan required that Oas sancies be taken shortly after the cask was filled
' with a different gas, and immediately prior to evr.uating a gas from the cask. The gas samples obtained during the test are indicated in Table 3-9. Each time the cask backfill pedium was changed, the cask was pumped down, ' backfilled with the desired g
noium, pumped down again, and finally backfilled. This ensured purity of all back-
- fill redia to >991. Nitrogen was used issnediately prior to vPcuum test runs to obtain a low-pressure (1 to 3 mbar), low-conductivity, vacuum / nitrogen environment.
The test plan formed the basis for developing a set of detailed operating procedures by INEL that outlined the steps required to perform the cask performance test. The procedures are discussed in the next section. [ INEL CASK MAtIDLING AfD OPERATING EXPERIENCE Tnis section describes the cas'k handling and operating experience gained during cask performance testing. The tasks required to conduct cask performance testing included performing storage and shipping cask handling studies, assessing the use of existing facilities and equipment, installing cask ancillary and research equipm6:." at the IHEL TAN cask testing facility, operation preparations, storage cask receipt and f preparations, operational dry runs, a facility readiness review, fuel transfers and 3-63 l
* ' * " " * * * =.
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, 'l' oading, cask performance testing, fuel assembly inspections, and longoterm surveil.
lance. INEL personnel performed a dry run to train personnel and check out equip. { eent. After the dry run was evaluated. necessary equipment and procedure changes
- e were made. The cask performance test was then conducted.
When they arrived frohl the Surry reactor at the INEL TAN.607 Hot Shop in TN 6L ship. j ping casks. The fuel asserablies wer o the TN 24P cask. Opera. tions personnel monitored the cask during this time. Preliminary testing began when
- the cask was partially loaded with 18 fuel assembites. When the cask was fully i loaded with fuel, the cask was transferred to the warm shop test bay where formal ,
testing began. ' ! As the TN 24P cask was being loaded with fuel, certain selected fuel assembites wer-. visually characterized by video and still photograph. Crud and smear samples were i also' collected and analyzed. At the conclusion of the formal testing in the warm shop, the TN 24P cask was moved i to the hot shop were the cask was prepared for temporary storage and monitoring. The cask was then placed on the long-ters surveillance pad and connected to the data
- collection system.
1 Each of the tasks requirid to conduct tne TN 24P cask performance testing is described in the following subsections. 4 Storace and Shipping Cask-Handling Studies Two detailed cask. handling studies were performed. One study developed the handling logic for the TN 24P ttorage cask, and the other defined handling of the Transnuclear l 0 TN.8L shipping cask and fuel. The handling studies identified how the shipping and storage casks would be received at INEL and how fuel could be handled. They com. pared test requirements of the approved test plan with INEL site capabilities. The specific equipment and systems required to accompitsh the cask performance test were defined. The studies identified modifications or upgrades to base facility equip. , ment needed to accomplish the performance test. The two handling studies became the bases for identifying the equipment and operational preparation rubtasks required to accomplish the cask performance test. Facilities and Eouioment Existing INEL equipment and facilities were used as much as possible. Those at the Central Facility Area (CFA) and Test Area North.(TAN) were the most extensively 3 64 amm , w,ymm - e,, m ,,me,
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, c, sed. Facility equipment capability, modification requirements, and maintenance --
requirements were evaluated. Equipment at CFA -the rail system, gentry trane, and heavy haul trailers, ' tads, and bridges--were identified. Equipment located at the
, TAN-607 hot shop was also evaluated. Equipment systens evaluated included the hot l shop crane; manipulators; direct viewing shielding windows; floor loading; hot cell transporter; in shop utility pedestals; operating galleries and shielding; in-shop rail track; shielo doors; hot shop lighting; cask gas venting systeei; and facility safety support syst.as. Equipment and systems evaluated outside the hot shop were the local in plant ati track; turntable; locomotive; and the warm :. hop capability for cask testing. Areas along and adjacent to the hot shop wste evaluated as a site .
.for the long-term surseillance pad. These base facility systems and equipment were ' 3 previously discussed in the section on the !NEl. cisk testing facility.
Two types of project-specific equipment were identified: cask test support and . cast.. handling / operation equipment. Cask test support equipment was that required to gather test data. It consisted of the gas / vacuum / vent system and the data acquisition system. Cask handling / operation equipment was that required to handle the cask, such as lift yokes, cask lid lifting fixtures, cask surface seal pro-tectors, thermocouple lance template, and the cask gas / vacuum / vent valve tree. . The base facility equipment modifications and project-specific equipment systems were .1esigned, procured, and installed concurrent with performance of the tasks to prepare the facility for operation. After a review of the TM-24P cask manual and dr,4 wings, specialized handling fixtures and tools were developed for hands-on and remote cask operation. Thermocouple lance insertion was closely reviewed. Special semiremott insertion tools were developed to reduce personnel radiation exposure and
- contamination spread while installing and removing the lances.
Roth base and project-specific equipment were operationally tested before they were actually used for remote operation. The equipment was tested either during a dry , run training task or by an independent, formal system. operation test. When problems were encounter.'d, they were resolved, and the equipment or system was retested. Operational Preparations Numerous tasks were required to prepare the facilities for the cask testing proj- , ect. Operating documentation was develorad, personnel were trained, the Til-24P ttnrage cask was received and moved to the TAft test facility, and a dry run was per-formed to check out the egyipment an:' operating procedures. A facility readiness 3-65 m._.__..- :.
review, held prior to the first storage cask test, approw d tn3 facility for cask
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te4 ting. tio fcrmal readiness review was held for the TN-24P cask test. Documentation Development. Reviews compared the base facility safety and standard l operating documentation with the TN 24P cask testing work scope. A facility fuel criticality analysis was performed on the TN-24P cask, which indicated that, during cir loading of the cask in the TAN hot shop, if water fran f acility sources were to enter the cask, a aultipitcation factor above the limiting value (0.95) would occur. Poison rods were placed into 16 of the 20 fuel cssembly guide tubes at the reactor loading facility to provide the required INEL site criticality safety margin. Operational sequences were verified, and revised addendums to the facility Safety ' Analysis Report (SAR) and Operation Safety Requirements Document (OSRD) were pre. . pared and approved. Facility Standard Operating Practices were also reviewea anci revised as required to meet the cask-testing requirements. Site Work Releases (SWRs) or Hot Cell Work packages controlled $11 operating tasks performed at the facilities. The SWR work, general work using craf t labor, does not ' ( require rigorous control and review. It usually involves equipment installation or naintenance. Hot Cell Work packages identify the tasks or subtasks required to accomplish a specific scope of work, and delineate a specific sequence for facility operating tasks. These work packages usually contain one or more detailed operating procedures (DOP), which are step-by-step' instructions for performing a specific task. The SWRs, work packages, and DOPs are controlled documents. As such, they g must be revised and approved, should a work step need changing. Work stops wnile e the package is being revised, and does not resume until the revision is approved. The overall project statement of work and the Til-24P test plan were ussa to develop inis operating documentation. Information for preparing the procedures came fran the cask vendor, safety analysis, equipment drawings, and operating and naintenance manaals. Safety, quality, project, independent safety, and operations personnel l
*1gornusly reviewed these work packages and DOPs.
inree Hot Cell Work packages were prepared to provide herating instructions for the operating persnnnel. The first consisted of training procedures to familiarize personnel with the cask, check out and perform the cask operating equipment fit-up, and checa the operating procedure. The second work package provided instructions ; f ar a t sal fJel receipt and cask loading. The third package instructec personnel in ' cas< tes'.ing and long-tern surveillance at the pad. 3 66 -
- - . ~ - _ -. .
The 0*talled Operatitg procedures used f*r handling, operattIg, and testing the THet# dry fcel storage cask are IIsted la Table 310. ' A cocvMnt control office managed the release a' nd change control of the cask operating ano safety procedures and documents. The document control office l maintained the facility operating project, ressarch data, research photographs, project equipment, and operating cost and schedule flies. O r etional Training. personnel training requirements were seveloped for the TN.24P cask and the Transnuclear TN.lL shipping cask. Training consisted of both classroom and hands.on training, Geration technicians and supervisory personnel received . pt sonal training packages before beginning the classroma instruction. StoraQt _f.ath Receipt and preparation. The TN.24P storage cask was received at the CFA in late September 1985. It had traveled from the west coast to the INEl, on a special heavy depressed. center roll car. The cask, attached to a horizontal ship. ping and storco^ cradle, was off loaded fran the rail car (Figure 3 43). The 200 ton gentr. e rat.e was used for this operation. The cask remained at CFA while ( the TAti are6 was being prepared.
- A heavy. haul transport plan was prepared to coordinate moving the cask from the CFA to TAN. An aver.the. road 150. ton heavy. haul trailer was leased to tranrert the cask ' Figure 3-44). The cask was transported to TAN in mid October and moved to the TAtl.607 hot shop preparation area.
Operations technicians uncrated the cask from its special harde.ood container and prepared to c'f. load it at TAN (Figure 3 45). While the cask was being prepared, the hot shop was readied for lif ting the cask. a contamination. free zone was esteh. ligned insine the hot shop. The heavy heul ' railer containir.g the cask was inoved into the hot shop and located parallel to the bridge crane (Figure 3 35). The cask-l lihiv; yoke was attached to the crane hook artl connected to the top trunnions of ta* cant. . Ap* rations technicians used the crane bridge travel and hoist motion to rotatet the case to a vertical orientation (Figure 3-46). The rigging and crane hook were kept veettcal tnrour,hout cask rotation. Then, the cask was suspenced from the crane off tn tne Side of the heavy. haul trailer, and the trailer was removed from the hot shop. Tne hortrontal cast storage cradle wes removed and placed into tympurary storage. .c 3 67 . 1
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Table 3 10 OtfAllto 0 PIRATING PROCEDURCS FOR TN 24P CA$K mat 0 LING Ato fililtiG 00P Nmhee Title t 1.13.24 TAN $FSC Program Grid Marking for Instrumentation of TN 2aP Storage Cask 1.13.25 TAN $F$C Progrin Thermocouple 81ock Mounting for TN 24P $torage Cask 1.13.26 TM SF5C Program Transfer TN 24P 5 tor 49e Cask from Transporter to P103 DP11y and k%,e to Wars Shop I
, B 1.13.27 TAN SFSC Program TM 24P 5torage Cask Training Equipment Fit Up in Warm Shop 1.13.2d TAN $FSC Prngram Transfer TN 24P $torage Cask irem P103 Dolly to Hot
$ hop Storage Cask Work Platforn. Hove CASTOR V/21 Storage Cask to Warn Shop
*1.13.29 TAN 5FSC Program TM-24P 5torage task Training Fuel Assembly Fit-Up, and Fuel Transfer Dry Run in Hot Shop
( 1.13.30 TAN $FSC Program Fuel Receipt and Transfer from TN WL Shipping Casks to TN 26 Storage Casks. 1.13.31 TAN SFSC Program Checkout of Lance and Basket Thermocouples for the TN 24P Storage Cask 1.13.32 TAN $FSC Program Installation and Operational Check of Data Acquisition
$ystem (DAS) and Transducers IN 24P Storage Cask 1.13.33
- TAN $FSC Program Dosimetry Installation and Removal for TN 24P Storage Cask 1.13.34 TAtt $FSC Program Install .TC Lances in TN-24P Test Ltd and Leak Check 1.13.36 TAN 5FSC Program Move TH 24P to W/5 Move CASTOR-V/21 to H$
1.13.36 TAN SFSC Program TN 24P Cask Vertical Test Mattle Runs 1, 2, and 3
- 1.13.37 TAN $FSC Program Rotate TN 24P Cask to Horizontai !
1.13.3d TAN $FSC Program TN-2aP Cask Horizontal Test Matris Runs 4, 5, and 6 1.13.39 TAN SFSC Progran Move TN 24P to its Move CASTOR-V/21 to WS 1.13.40 TAN SFSC Progran Remove TC Lances from TN-24P Test Lid and Leak Cneck Penetr.stions 1.13.41 Tall 5FSC Program TM 24P Cask Interim Monitoring at Test Pad 1.13.44 TAN SFSC Program TN 24P Cast interir Gas $anpling at Test Pao 3 68 a
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Figure 3 45. Preparing TN 24P Cask for C*f Loading at TAN-607 Not Shop SE..?" ,.
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e ine cast was then placed into the hot shop work platfom. $hics were used to level the cask so the lid surfaces would be as level as possible. Level tid surfaces facilitate remote Ild removal and fuel assembly loading. Because l 4 contamination barrier was required to reduce the
- g l possiotitty of contaminating the outer cask surfaces. Teflon was used for this because of its high melting point. Teflon sheets were attached to the cast with Teflon tape. Cask familiarization, training, and ancillary equipment / tool checkout j
was performed. Using the personnel work platform for access and the hot shop crane l for hoisting, operations technicians wnted the cask cavity and removed the cask j rain cover, pressure-monitoring accessories, and standard shield lids. Seals and seal surf aces were inspected, and the lid surface seal protectors were installed. , l l The lid seals were inspected, and a shipping basket restraint was removed. Clear-i antes were checked to ensure they would allow a TC lance position template to be j installed on top of the fuel basket. The template was used to verify the test lid j lance penetretton locations and assist in lance insertion training. The test lid ! was then reinstalled. Thermocouple lance blind flanges covering the test lid pene- [ trations were removed one at a time with a semiremote reach tool. A mockup TC lance ' i was inserted through each lid penetration and template location hole. A rope block j and tackle and remote reach tools for positioning the tip of the lance were used for j installing the lances. When the TC lance installation training was conpleted, the blinr1 flanges were replaced and the test lid and the TC lance template were removed. 4 A remote shield Ild lifting fixture was attached to the shield lid, which provided a F a cylinder-into cylinder sliding handling effect rather than a disk into-cylinder ) effeet. This helped prevent the lid frcn bind (ng when it was placed into the cask j lid cavity. *
, Operational Ory Run. An operational dry run was performed in the hot shop. Because the hot shop personnel were trained previously on the TH-BL shipping cask, the dry run concentrated on the TN 24P storage cask handling and spent fuel loading into the
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- storage cast. The dry run trainad personnel and checkett out the operating facility, project-specific equipment. and procedures.
I The dry run began with the TH 24P in the hot shop work platform (Figure 3 47). The
- cask was vented through a controlled pas / vacuum / vent system. Technicians then removed the lid bolts and installed the lid-lif ting fixture. Numerous equipment Checkout procettures were performed to verify that all Cask-hand 1Mg equipment and f acility safety systems were ready for remote operation.
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I figure 3-47. TN-24P Cask in Hot Shop Work Platform Although the actual procedure would ensure that all personnel were ev cuated from ' the hot Shop, certain personnel remained in the hot shop for the dry run to observe the operation. However, they did not assist any of the remote operations. During I the dry run, different operating technicians repeated the following handling s*quences several times: .
- removing the TN-24P storage ' cask lid and placing it on lid support st6nds on the hot shop floor
. e installing the lid seal surf ace protectes on the storage casks
.e I connecting the crane to the fuel grappN sad attaching the grapple to the nockup fuel assembly
- using the power rotate on the grapple t$ align the fuel assembly vertically and rotationally ,
- towering the fuel assembly into the storage basket (Figure 3-43).
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I 4 h i i . 1 ! Figure 3-48. Mockup Fuel Assembly Installed into TM-24P Cask [ l During Dry Run 4 - { During these operations, the fuel grapp1e lead cell was monitored closely to prevent the possibility of hanging up the fuel assembly, which could cause the grapple sys-tem to fail. When the mockup fuel essembly had been inserted into each tube of the p (
- storege cask basket, the assembly and grapple were returned to their storage racks.
The operations technicians then remotely removed the cask seal surface protector j f rom the storage cask. Using a tool held in the overhead manipulator, the seal surf aces were wiped down to remove any foreign material from the mating surface. The test shield lid was then reinstalled on the storage cask. E At this point in the handling sequence, hands-on work was performed in the hot shop. Operations technicians installed and torqued the cask lid bolts. The cask was evacuated and backfilled with nitrogen. . Fuel Transfers and loadino The TN.24P storage cask was loaded with fuel during November and December 1985. Loading fuel into the TN.24p cask was routine, and no problems were encountered. ! The fuel transfers and loading followed the procedures verified during the dry 1 4 3-73 gg. .p
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run. The Iceding, vacuum pumpdowfi,'and decontamination operattoxs and esperience are af scussed in this section. Personnel radiation esposure levels estimated to have occurred shsring fuel transfers, loading, and testing are also presented. - Cask Leadino. The eight TN R. shipping casks arrived at the TAN facility at any [ hour of the day or night. Because each cask shipment was tracked along the shipping route, facility operations personnel received the estimated time of arrival. They then could Schedule the unloading with other facility project work. Generally,if the cask arrived on the weekend or late at night, it remained at the TAN guard gate. - Sometimes the area health physicist would escort the shipment to the TAN hot shop annen butiding, where it was parked inside. The cask woulo be unloaded during the ' nest regular work shift. The cask unloading turnaround time usually required two and one half shifts or about 20 h. Plannad overtime and weekend wort supported a shorter task shipping schedule. Eight shipments, with three fuel assemblies each, were received to fully load the TN 24P storage cask. No handling problems were encountered, and the cask handling ( equipment and operating ancillary equipment performed well. Gas samples were taken from each shipping cask and analyzed before the cask was vented, leo fission product gases were detected. The shipping cask radiological surveys were belon legal shipptog limits. 'The INEL shipping requirements made minor decontamination necessary before casks could be shipped from INEL. l The first and seventh shipments arrived in the same cask. The cask internals were I radiologically emeared after the first, and seventh shipment to determine if contami- ! nation had increased. Standard smears with 4.25 cm-diameter (1.67-in.) tabs over an ! area of 100 cm2 (15.5 in.2) with an applied force of 35 kPa (5 psi) wre taken from the fuel basket bottom and near the bottom on the side wall. Crud collection sam-ples were also taken from the bottom surface of'the cask fuel basket. The samples [ were sent to LLNL from the bottom surface of the cask fuel basket. The samples were sent to LLNL to compare the samples from the seventh shipment with those from the first (see Table 3 8). TN-24P cask surf ace temperatures reached 90'C. Therefore, Teflon sheeting with a high melting point replaced the standard polyethylene sheeting. The sheeting served l i 3-74
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* .as a barrier to prevent storage cask contamination. Teflon sheeting ta the tempera.
ture range required was available only in rolls 0.9 m (34 in.) wide. The sheeting was att, ached to the cask with Teflon tape. The seams were loosely overlapped and taped. l To prevent other top surfaces of the work platform near the cask from being contant-nated, the following arrangement was used. Teflon sheeting was attached to areas contacting the cask. Polyethylene sheeting was taped to the Teflon at a safe dis-tance frcun the cask. The contamination barrier also acted as a thermal barrier, preventing some heat transfer from the cask. However, the contamination barrier was used only while the cask was being loaded. It was removed during formal thermal ' cask testing. The TN-24P storage cask operations were very satisfactory. Cask handling and load. ing operations were easy. Vendor service and information response were good. Vacuum Pumodown. A valve tree connected to a cask nonttoring port allowed cask vacuum pumpdown and gas backfilling. The valve tree was connected by quick-( disconnects and vacuum hose to the ges/ vacuum / vent system. A pressure transducer, teed into the valve tree, monitored cask cavity pressure.
- The cask vacuum pumpdown system required approximately 3/4 to I h to pump the cask from 850 mbar atmosphere pressure (12.25 psi) at 1463 m (4800 ft) elevation d'own to less than 1 mbar (0.01 psi). Backfilling the cask with cover gas required about 15 min.
Decontamination. steMwfese phrsenoeve h e n N L eiiigw-GMN@ OF *ma aALhirL#@Wme'n.nz- Howeyer, sinor con-
.k tamination did occur. The contamination was localized to the personnel work plat-form between the casks. About 4 h of decontamination were required before personnel
( could continue hands-on work for cask shipping and interim storage activities. The fuel assembly grapple lower assembly and tool fingers becane highly contaminated after each fuel unloading. The lower assembly was decontaminated and begged between use. I Estimated Personnel Radiation Exposures. During the 3 months personnel were loading and testing the TN-24P cask, operational radiation and temperature monitoring were 3-75 l
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, performed. The monitorir.g provided current actual data for personnel safety. . Tem- [*
peratures and radiation increased with each fuel loading, and more personnel safety equipment' was used. A combination of materials was used to reduce personnel expo. '
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sures on top of the cask for lid bolt removal, lance installation, and gas samples. ' Thermal blankets were placed on the top and down the sides of the cask to reduce - the effect of the high temperature hazards. Blankets for thermal insulation were 0.61 m x 1.2 m x 2.5 cm (2 ft x 4 ft x 1 in.) thick. norated poly sheets, 2.5 cm thick (1 in.) were used for neutron shielding, and 1.27-cm tnick (1.5 in.) lead wool blankets provided gassa shleiding. This loose laid protective material was repo. - sitioned as required for access to different cask penetrations and for lid bolt installation. Wn it was not practical to use shielding, every effort was taken to reduce personnel exposures by reducing time in the radiation fleId. Personnel radiation exposures during the handling, loading, and testing of the
' TN 24P cask weret
( e fuel receipt and loading. 0.3 man-rom 1
- thermocouple lance installation and ' emoval r - 0.3 man rom [
o cask handling - 0.3 man-rom o testing (instrumentation) - 0.3 man-rom. 1 i .
. Loading and testing the cask required extensive hands-on operation. For example, thermal testing, radiation dose rate monitoring, multiple pas backfilling, and sang. I ling were hands-on operations. These operations were performed to support the cask performance test, but they would not be required for consercial power plant under-water fuel loading. Hence, radiation exposures under actual storage scenarios would be much lower than those encountered during this cask performance testing effort.
Cask Performance Testino Preliminary thennal testing was started when the TN-24P storage cask was loaded with 18 fuel assembites. This was done to ensure that fuel would not exceed maximum allowable temperatures charing the cask loading and to obta'in early cask heat trans-fer data. The preliminary data ensured that fully loaded cask surface temperatures and fuel temperatures would not exceed allowable values. l Fuel temperature data were collected using TC lances inserted through a penetration ! in the test lid and into the guide tube of a selected fuel assembly (Figure 3-49). The TC lances were connected to the data acquisition system. The OAS collected ( temperature and pressure data during cask pumpdown and interim storage between fuel 3-76 l e e . .e -. . . . my .g . , _ m_ m -N -M'm~ ""' " - " " - - " ' " ' '
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( Figure 3-49. Installing Thermocouple Lances into the Fuel Assembly Guide Tubes Through the TN 24P Test Lid Ioadings. Cylindrical spacers 8.0 cm (3.5 in.) long were placed between TC lance flanges anti the cask primary test lid to permit proper installation. { 4 TC lance was installed after each subsequent fuel loading untti the cask was fully loaded. Data were collected weekly and transmitted to PNL. '
' unen the TN-24P storage cask was fully loaded, it was tested according to the approved test plan. The test lid was bolted on. Operations personnel installed all nine TC lances through the test Ild into seva fuel assembly pide tubes and two I basket locations. The cask was double-pumped and osctf t11ed with helium cover gas to ensure pas purity. Then the cask was moved into the warm shop test bay, where it was externally instrumented with TCs at predeterntt.ed locations (Figures 3-50 and 3-33). All TCs and the pressure monitor were connected to the DAS.
The cask was already near the equilibrium temperature because of heat butidup during the cask loading period. Only a short period was required to reach the peak equi-l librium temperature. Monitoring continued for at least 24 h after steady-state 3-77 - h
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- i once the vertical heltwi test run data had been verified, the cask cover gas was -
changed to nitrogen. Gas samples were taken at the beginning and end of the ver. l tical nitrogen test run. Temperature and pressure monitoring continued throughout ( the vertical nitrogen test run. A vertical vacuum test run was conducted in a
. similar manner.
Two to three days were required for the cask to reach steady-state temperature af ter a position or cavity backfill gas change. When the cask was rotated from toe vertical to the horizontal orientation & ring the performance test, the cask TCs nad to be disconnected from the DAS. The cask was moved to the hot shop where the cask Wes rotated to the horizontal position with the hot shop crane. When the cask was back in the warm shop test bay, the TCs were ; , reconnected and the test run monitoring began. . k 3 78 - 4
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, measurements were taken on all cask surf aces using both portable survey instruments and dosimeters. [G&G istaho, Inc. health physt(s techniClans conducted the portable surveys, and Pf1L technicians conducted the dostmetry.
l At the completion of initial cask testing, the TN.24P cask was prepared for interim stor49e at the test pad. The TC lances were removed, and the cask cover gas pres-sure was adjusted to 1.5 snbar. The cask was placed vertical on the P103 dolly and moved to the long tem surveillance test paa area on March 13, 1986. Two locally , avallable cranes were used to off load the cask from the dolly onto the test pad (Figure 3 51). M+
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Fuel Assembly inspections . . During TN 24P cask loading, selected fuel assembites were inspected. A color TV camera system performed a four surface video characterization enemination while the fuel assemblies were being~ removed from the TN dL shipping cast. Two of the eight l assemblies were color photographed using a periscope for high resolution to confirn the condition of the fuel rods and crud buildup (Figures 3 24 and 3 25). Crud and smear samples were collected from these two fuel assemb11es. A 2.54 co-diameter (1.in.) rubber bottle stopper with a 2.54 cm. diameter (1.in.) unear filter attached to one side was used to take the smear samples. The rubber stopper was
' l held in the overhead manipulator hand and remotely pressed against a fuel rod. It was then moved vertically about 20 to 26 cm (8 to 10 in.). The rubber stopper was remotely placed in an open plastic bag. When hands on operations began inside the hot shop, the smear filters were manually removed from the stopper, placed in small shielded pigs, nd sent to LLNL for analyses.
I Crud collection samples were taken in a similar manner. A large diameter rubber . j stopper with double sided sticky tape was remotely pressed against a fuel rod and then retracted. The sample was then placed in a plastic bag, and the bag was manually placed into a shielded shipping pig. 'l
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- P.O. Boa 174
- Lake 32me,NW 88o43
- Phones 613 770 386s
- FAX 770 3976 August 28,1997 PRESS RELEASR FOR lh0 fFD1 ATE REl. EASE RE- PETITION FII.ED WIT 11 THE U.S. NUCLEAR RDOUL ATORY COMh0SSION (NRC) TO SUSPEND NORTifERN STATP.5 POWER COMPANY'S (NSP) LICINSE TO OPERATE ITS H10H.Ll! VEL NUCLEAR WASTD DUMP ON PRAIRIE ISLAND.
NRC R1! OUI.ATIONS llAVE BEEN VIOLA'TT.D DUE TO T!!E INABTLITY OF NSP TO S AFEl.V MAINTAIN ANO UNLOAD STORAGE CASKS CONTAINtNO NUCLEAR WASTE ON PRA!RIE ISLAND. FOR MORE INFORMATION CONTACT: George Crock er, Steering Committee, Pralrie lstand Coalition (612) 770 3861 , Today the NRC was served with a petition Rom the Prairie Island Coalition to suspend ' for cause NSP's Materials License for storing nuclear waste in dry casks on Pralric istand The petition speelfics numerous vlotations ofMRC regulations that require NSP to operate a dry cask storage facility that can be safely maintained, and that can safely unload nuclear waste from dry casks. Now,it is apparent that these functinns are not physically rossible. NRC regulations also require NSP to be truthful and honcst about the information it u. - to get authority to operate a dry cask storage facility. The petition documents instances in wi..ch NSP knowingly used incomplete and inaccurate Informa11on to ad mch authority. Finally, the petition requests formal rulemaking proceedings to set parameters for safe and responsible dry cask storage operstions The fbndamental problem is that, despite NSP assurances to the contrary, the casks storins nuclear waste on Prairie Island cannot be safely unloaded, and required procedures to maintain cask se&Is cannot be safely performed. Storage casks cannot be unloaded, and cask seals cannot be maintalncd because over time, the nuclear wa'ste in the casks can warp, or disintegrate to debris that collects on the bottom of casks. To maintain cask seals, and to untoad casks so the waste can be transported to some other place, the casks must be placed back into the storage pool and be reflooded for radiation protection. But the waste will still be extremely hot Pool water will flash to steam, which must be vented, but t!m steam will be laced with radioactive gases and panicles. NSP does not know how to contain these radioactive materials. And cool pool water on hot nucIcar waste will produce a thermal shock. This shock may cause spent fuel assemblics to totally disintegrate and fall to the bottom of a cask in a pattern that makes the fuel to go crlilcal. This could lead to c.atastrophic radiation releasc. (
"Because of unloading and maintenance problems, each cask on Pralric lstand is a nuclcar time bomb," said George Crocker of the Prairie Island Coalitinn. It is our hope that this petition will help us all find more tcsponsible ways to reanage Mjnnesota's;Tpelear wntsAerort ,
these bombs go off,' bc concluded. I
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