ML23156A064

PRM-072-004 - 63FR12040 - Prairie Island Coalition; Receipt of Petition for Rulemaking

ML23156A064
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Issue date:	03/12/1998
From:	Travers W NRC/EDO
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PRM-072-004, 63FR12040
Download: ML23156A064 (1)

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KEYWORD: ADAMS Template: SECY-067 03/12/1998 PRM-072-004 - 63FR12040 - PRAIRIE ISLAND COALITION; RECEIPT OF PETITION FOR RULEMAKING PRM-072-004 63FR12040 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

PRAIRIE ISLAND COALITION; RECEIPT OF PETITION FOR 63FR12040 PRM-072-004 tMENTNO 2 3 4 5 6 DOCDATE 08/26/1997 03/06/1998 05/12/1998 05/20/1998 05/20/1998 05/26/1998 05/22/1998 05/26/1998 Monday, April 09, 2001 DKTDATE NAME REPRESENT 10/16/1997 03/06/1998 05/18/1998 JUDY TREICHEL, NEV ADA NUCLEAR 05/26/1998 LYNNETTE HEND NUCLEAR ENERGY 05/26/1998 ROGER 0. ANDER NORTHERN STATES 05/26/1998 MICHAEL E. MAS TRANSNUCLEAR, I 05/27/1998 DIANNE R. NIELS ST ATE OF UT AH - D 05/27/1998 GEORGE CROCKE NORTH AMERICAN DOCDESC1 LTR FM GEORGE CROCKE SUSPEND NORTHERN STAT LICENSE TO OPERATE ISFS Page 1 of2

~MENTNO DOCDATE 7 05/27/1998 8 05/18/1998 9 05/25/1998 05/25/1998 II 05/26/1998 06/11/1998 06/18/1998 12 06/18/1998 01/18/2001 Monday, April 09, 2001 DKTDATE NAME REPRESENT 05/28/1998 PRESTON TRUMA DOWNWINDERS, IN 05/29/1998 CAROL A. OVERL SELF 05/29/1998 MARY P. SINCLAI SELF 06/03/1998 KRISTEN EIDE-T COMMUNITIES UNI 06/03/1998 BARBARA WARN SELF 06/15/1998 06/22/1998 06/23/1998 GEORGE CROCKE PRAIRIE ISLAND C 03/01/2001 DOCDESC1 LTR FROM CAROL A. OVE ADDITIONAL COMMENTS ( LTR FM CAROL A. OVERLA THE ELECTRIC POWER RES APRIL 1998 REPORT, NUM FRN-DENIAL OF PETITION FoR-I< LLLtY\\t\\ ll, Al G, Page 2 of2

DOCKET NO. PRM-072-004 (63FR12040) In the Matter of PRAIRIE ISLAND COALITION; RECEIPT OF PETITION FOR RULEMAKING DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 10/16/97 08/26/97 03/06/98 03/06/98 05/18/98 05/12/98 05/26/98 05/20/98 05/26/98 05/20/98 05/26/98 05/26/98 5/27/98 05/22/98 05/27/98 05/26/98 05/28/98 05/27/98 05/29/98 05/18/98 05/29/98 05/25/98 06/03/98 05/25/98 06/03/98 05/26/98 06/15/98 06/11/98 LTR FM GEORGE CROCKER TO CALLAN PETITIONING NRC TO SUSPEND NORTHERN STATES POWER CO. MATERIALS LICENSE TO OPERATE ISFSI AT PRAIRIE ISLAND PLANT FEDERAL REGISTER NOTICE - RECEIPT OF PETITION FOR RULEMAKING COMMENT OF NEVADA NUCLEAR WASTE TASK FORCE, INC. (JUDY TREICHEL, EXEC. DIRECTOR) (

1)

COMMENT OF NUCLEAR ENERGY INSTITUTE (LYNNETTE HENDRICKS, DIR. PLT. SUPP.) (

2)

COMMENT OF NORTHERN STATES POWER COMPANY (ROGER 0. ANDERSON) (

3)

COMMENT OF TRANSNUCLEAR, INC. (MICHAELE. MASON) (

4)

COMMENT OF STATE OF UTAH - DEPT. OF ENVIRONMENTAL QUALITY (DIANNE R. NIELSON, PH.D., EXEC DIR) (

5)

COMMENT OF NORTH AMERICAN WATER OFFICE (GEORGE CROCKER, EXECUTIVE DIRECTOR) (

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COMMENT OF DOWNWINDERS, INC. (PRESTON TRUMAN, DIR., ET AL.) (

7)

COMMENT OF CAROL A. OVERLAND (

8)

COMMENT OF MARY P. SINCLAIR, PH.D. (

9)

COMMENT OF COMMUNITIES UNITED FOR RESPONSIBLE ENERGY (KRISTEN EIDE-TOLLEFSON) (

10)

COMMENT OF BARBARA WARNER (

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LTR FROM CAROL A. OVERLAND, ESQ., SUBMITTING ADDITIONAL COMMENTS (SEE COMMENT NO. 8)

DOCKET NO. PRM-072-004 (63FR12040) DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 06/22/98 06/18/98 LTR FM CAROL A. OVERLAND, ESQ. RE ISSUES RAISED BY THE ELECTRIC POWER RESEARCH INSTITUTE IN THEIR APRIL 1998 REPORT, NUMBER TR-108757 06/23/98 06/18/98 COMMENT OF PRAIRIE ISLAND COALITION (GEORGE CROCKER) (

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10 CFR Part 72 [Docket No. PRM-72-4] [7590-01-~] OOCKETEIT USNRC

• 01 MAR - 1 A 8 : 1 4 OFFIC~ OF SECRETAR'1 RUL 'MAKINGS AND ADJUDICATIONS STAFF Prairie Island Coalition; Denial of Petition for Rulemaking AGENCY: Nuclear Regulatory Commission.

ACTION: Denial of petition for rulemaking.

SUMMARY

The Nuclear Regulatory Commission (NRG) is denying a petition for rulemaking (PAM) submitted by the Prairie Island Coalition (PRM-72-4). The petitioner requested that the NRG undertake rulemaking to examine certain issues regardi~g the potential for thermal shock and corrosion of spent fuel in dry cask storage; amend its regulations governing the storage of spent nuclear fuel in dry storage casks to define the parameters of acceptable oegradation of spent fuel in dry cask storage; amend its regulations to define the parameters of retrievability for spent nuclear fuel in dry cask storage; and require licensees to demonsH"ate safe cask unloading capability before a cask may be used at an independent spent fuel storage installation (ISFSI).

ADDRESSES: Copies of the petition for rulemaking, The Federal Register notice of receipt, the public comments received, and NRC's letter to the petitioner may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. You may also access these documents on NRC's interactive rulemaking website at http://ruleforum.llnl.gov. For information about the interactive rulemaking site, contact Ms. Carol Gallagher, 301-415-5905; e-mail (CAG@nrc.gov). FOR FURTHER INFORMATION CONTACT: Gordon Gundersen, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone 301-415-6195, e-mail (geg1@nrc.gov). SUPPLEMENTARY INFORMATION: The Petition On March 12, 1998 (63 FR 12040), the NRC published a notice of receipt of a petition for rulemaking filed by the Prairie Island Coalition (PIC). The petition consists of the issues presented in paragraphs 13, 14, and 15 of the document attached to an August 26, 1997, letter from George Crocker, PIC, to L. Joseph Callan, Executive Director for Operations, NRC. The issues in the remainder of the August 26, 1997, document were submitted as a petition under 1 O CFR 2.206 of the Commission's regulations. The Director of the Office of Nuclear Reactor Regulation responded to this petition in a Director's Decision (DD-98-02) published on February 20, 1998 (63 FR 8703). Similar issues were addressed by ~he Director in a decision (DD-97-18) published on September 8, 1997 (62 FR 47227). Those issues concerned the degradation, retrievability, and unloading of spent nuclear fuel from dry storage casks. 2

The petitioner requested an amendment of the regulations in 10 GFR Part 72 to define the parameters of spent fuel degradation that are acceptable to the NRG under 1 O GFR 72.122(h). Section 72.122(h) provides that spent fuel cladding must be protected during storage against degradation that leads to gross ruptures or that the fuel must be confined such that degradation will not pose an operational safety conce~n. The petitioner is concerned about the potential effect of spent fuel degradation on the ability of licensees to unload a dry storage cask safely. PIG contended that the NRG has not adequately addressed the possibility of damage to spent fuel caused by thermal shock when cool water refloods a cask that contains dry spent nuclear fuel. The petitioner also contended that the NRG has not adequately addressed degradation of spent nuclear fuel resulting from the loss of helium from failed cask seals or because of the passage of time. The petitioner stated that no procedures have been developed to ensure operational safety or to assess worker or offsite radiation exposure when degraded spent fuel must be unloaded. The petitioner also requested an amendment to the regulations in 10 GFR Part 72 that govern storage of spent nuclear fuel in dry storage casks to define the parameters of retrievability of spent fuel required by the NRG under 1 O GFR 72.122(1). Section 72.122(1) states that spent fuel storage systems must be designed to allow ready retrievability of the spent fuel for future processing or disposal. PIG is concerned that the NRG has not taken into account the potential problems that may be encountered in unloading a cask to retrieve spent fuel. Lastly, the petitioner requested an amendment to the regulations to require licensees to demonstrate the ability to unload spent nuclear fuel safely from a dry storage cask before a cask can be used at an ISFSI. The petitioner contended that if a licensee can demonstrate ability to unload spent nuclear fuel safely from a cask in a pool after long-term storage, the 3

public will have assurance that a spent fuel storage cask can be unloaded. PIG believes that although the NRC's regulations do not require a licensee to be able to unload a cask immediately, the NRG clearly requires a licensee to be able to unload the spent fuel at some point. The petitioner also believes that because in-pool unloading of spent fuel from a dry storage cask (that has contained the fuel for a protracted period) has not been completed, there is sufficient reason to require a licensee to demonstrate the capability to unload a dry cask underwater. PIG stated that it would be satisfied if a licensee can demonstrate the ability to unload spent nuclear fuel from a dry storage cask at some reasonable point in time. Public Comments on the Petition The notice of receipt of the PRM invited interested persons to submit comments. The NRG received letters from 12 commenters: the State of Utah, five private organizations, three associated industries (including one from the Nuclear Energy Institute (NEI)), and three private individuals. The comments focused on the main elements of the petition-degradation, retrievability, and unloading of spent nuclear fuel from dry storage casks. The NRG also received responses from the petitioner and one of the commenters on many of the points raised in the comments. The NRG reviewed and considered comments and responses in developing its decision on this petition. Eight of the commenters supported this petition for rulemaking. Commenters supporting the petition pointed out that a number of degradation modes exist for dry cask storage systems, including flawed cask fabrication, neutron flux and irradiation, liquid metal embrittlement, metal creep, hot metal reactivity, and thermal shock. These commenters believe that any of these modes of degradation may render unloading procedures ineffectual in terms of protecting workers, the public, and the environment from unacceptable radiation exposure. They believe 4

that a rulemaking is needed Jo create procedures that ensure the safe management of the waste during a variety of contingency situations that may arise as a result of degradation. One commenter believes that a rulemaking is needed to provide a direct connection between NRC requirements and concerns about fuel integrity and the procedures necessary for monitoring, retrieving, repairing, and maintaining cask systems. The rulemaking would serve to reduce the uncertainties surrounding degradation, loading, and transfer of nuclear waste and should provide for the development of contingency analysis of the interaction between storage timelines, technologies, and degradation factors. Other commenters supported the rulemaking proposed by the petitioner to address a variety of issues, including the viability of retrieval of spent fuel from dry storage casks, the need to specify procedures for managing cladding degradation, the need to determine how damaged spent fuel will be managed after unloading (since damaged spent fuel cannot be placed in another cask), and whether special unloading procedures are needed for dual-purpose casks (which not only store fuel for an extended period but also transport the spent fuel to a repository or storage facility) because during transport the fuel may be subject to vibration and accidents. The State of Utah believes that a rulemaking is needed because the set of dry storage parameters is too vague and does not provide sufficient guidance for the NRC staff or cask designers. Further, the State believes that dry storage systems have operated with extremely thin safety margins, as evidenced by the welding problems experienced with the VSC-24 casks. In particular, the State notes that NRC's experience with the unloading of spent fuel from transportation casks does not provide a basis for confidence that storage casks can be successfully unloaded because of key differences between the two, primarily that the State believes that the cladding temperature in transportation casks is much cooler than the cladding 5

temperature in storage casks and that transportation casks are used for a brief period, after which the fuel and cask interior can be inspected, which is not true for storage casks. The State also supports the petitioner's request that the NRC's regulations should require a physical demonstration of unloading ability and believes that a physical demonstration is particularly important at an off site ISFSI, given that the reactor and the spent fuel pool that supplied the irradiated fuel may no longer exist. The three industry commenters opposed the petition. The NEI observed that two Director's Decisions (DD-97-18 and DD-98-02) addressed the same issues with respect to the Prairie Island ISFSI and stated that the Director's responses demonstrate that there is no generic issue regarding safely unloading dry spent fuel storage casks, and hence no need for rulemaking. In NEl's view, experience, testing, and computer modeling have all shown that the combination of an inert atmosphere and temperature limits provides a robust basis to conclude that the integrity of the fuel will be maintained during the licensed storage period. NEI also believes that the fact that thousands of spent fuel assemblies have been successfully unloaded from shipping casks without damage - and that most of these shipping casks are designed for fuel temperatures higher than storage casks - provides confidence that storage cask unloading will not result in fuel damage. Moreover, NEI points out that the petition does not present any relevant technical, scientific, or other data to support the need for rulemaking. A cask manufacturer, Transnuclear, Inc., commented that Transnuclear metal casks are designed to store fuel below a maximum allowable fuel cladding temperature in an inert medium (helium) and that this is a well-established method of preventing cladding degradation. This commenter also stated that thermal shock to spent fuel assemblies is not a problem, that spent fuel assemblies have been successfully unloaded from shipping casks without damage, and that most of these shipping casks are designed for higher fuel temperatures than storage 6

casks. Therefore, thermal shock will not present a significant problem when the casks are reflooded with spent fuel pool water before being unloaded. The commenter states that fuel temperature limits as high as 570 degrees Celsius have been approved for transportation packages and that unloading of fuel from a transportation cask into a spent fuel pool without causing fuel degradation has been demonstrated in the Uriited States and France. In the case of unloading fuel from a storage cask, the commenter believed that the thermal shock phenomenon will be much less significant because of the lower fuel temperature (usually less than 300 degrees Celsius). In addition, the thermal shock is minimized by following procedures that allow the fuel to gradually cool down to the boiling point of water (100 degrees Celsius) before being submerged in the pool. The petitioner reviewed the comments received on its petition and provided a response. In the petitioner's view, rules governing procedures for safe management of contingency conditions during unloading do not presently exist and are needed. The petitioner states that the whole point of its request for rulemaking is that lack of actual knowledge about how waste materials will behave during storage and unavoidable management operations makes contingency planning necessary in order to protect against worker and public radiation exposure likely to occur if contingency procedures are not in place. The petitioner believes that phenomena such as high temperature zinc reactivity and thermal shock will allow site personnel very little time to evaluate the situation and initiate corrective action. Reasons for Denial The NRC is denying the petition for the following reasons: The petitioners identified several concerns pertaining to the lack of specific guidance in the unloading procedure to address a scenario in which significant fuel degradation occurs 7

during storage. The NRG staff agrees with the petitioners that such a scenario would complicate the unloading process by requiring additional measures and precautions to limit the release of radioactive materials from the cask into parts of the reactor facility and nearby environs. Currently, unloading procedures used by Part 72 licensees include a hold point to sample the atmosphere within the cask cavity to test for radioactive and flammable gases before venting the cask cavity and removing the cask lid. On the basis of the analysis of the gas sample, the unloading procedure includes a step to allow personnel to determine whether additional measures or precautions are needed to safely unload the cask. While acknowledging many of the petitioner's concerns regarding the potential difficulties in retrieving degraded fuel from dry storage casks, the NRG staff concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to address this unlikely situation. This conclusion is based on the NRG staff's acceptance of current practices and that the required compensatory actions and precautions needed to address such contingency situations may vary significantly, depending on the actual results from the analysis of the gas sample. On the basis of licensees' experiences in developing and implementing plans to address the problem of fuel assemblies damaged during reactor operations, in handling radioactive wastes of various forms, and in resolving other comparable problems, the NRG staff has confidence that licensees could, if necessary, develop plans to retrieve damaged fuel from a storage cask while minimizing the radiological consequences to plant workers and the general public. In addition, the NRG staff is confident that the technical problems associated with retrieving degraded fuel could be overcome. Furthermore, requirements for planning and executing such an activity are contained in the licenses issued for each ISFSI and power reactor, and in NRG regulations at 10 CFR Parts 20, 50, and 72. Therefore, the NRG staff has accepted gas sampling and defined hold or decision 8

points before breaching the cask confinement boundary as an adequate means of addressing concerns pertaining to the unlikely degradation of fuel assemblies during storage. In addition, the NRG inspects loading and unloading procedures during preoperational testing to confirm their adequacy. The NRG believes that the petitioner is incorrect in asserting that 1 O GFR 72.122(h) needs to be revised to define parameters of acceptable spent fuel degradation. The NRG believes that an applicant may store spent fuel without significant degradation in a sate technical manner without additional prescriptive requirements. In the present case, 1 O GFR 72.122(h) specifies the performance-based outcome that must be achieved by the licensee. The applicant must address all relevant considerations to achieve the outcome specified in the regulation. Specifically, paragraph (h)(1) of 1 O GFR 72.122 states, in part that: "[t]he spent fuel cladding must be protected during storage against degradation that leads to gross ruptures or the fuel must be otherwise confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage." Research, experience, testing, and computer modeling have all shown that the combination of an inert atmosphere and establishment of cladding temperature limits provides an adequate technical basis for concluding that the fuel integrity will be maintained during the licensed storage period. Industry experience in unloading transportation casks under water without incurring fuel damage and limited experience in unloading storage casks provides confidence that storage cask unloading will also not result in fuel damage. Additional experience on the long-term performance of spent fuel storage systems has been gained from NRG-sponsored studies. Specifically, the NRG studied spent nuclear fuel assemblies that have been out of the reactor for approximately 20 years. In September 1999, a Gastor-V/21 cask that has been at the Idaho National Engineering and Environmental Laboratory since 1985 was 9

reopened, and the cask internals, fuel assemblies, and several rods were visually inspected. This cask contained 21 spent pressurized water reactor fuel assemblies (with burnup in the 30-35 GWd/MTU range) from the Surry Nuclear Power Plant. These fuel assemblies have been in continuous storage in this cask for approximately 15 years. The examinations found no evidence of significant degradation of the Castor-V/21 cask systems important to safety from the initial cask loading in 1985 to the time of examination in 1999. The fuel examination found no long-term fuel degradation, thus confirming the adequacy of existing practices to protect the fuel. The NRC believes that the petitioner and the commenters have not provided adequate justification for revising the requirements in 1 O CFR 72.122(1) to include specific parameters for retrievability. The NRC reviews an applicant's method of retrievability to determine if it is appropriate for use rather than specifying in the regulations exactly how retrievability is to be accomplished. Each site must have specific procedures in place that are exclusively associated with that site, and the licensee should have the flexibility of achieving the outcome specified in 1 O CFR 72.122(1). Furthermore, Regulatory Guide 3.61, "Standard Format and Content of Topical Safety Analysis Reports for a Spent Fuel Dry Storage Facility contains an outline of the specific information needed, and NUREG-1536, "Standard Review Plan for Dry Cask Storage Systems" provides guidance to the NRC staff performing safety reviews of dry cask storage systems. These documents provide guidance to applicants and the NRC staff to ensure that the safety analysis report (SAR), the safety evaluation report, and the Certificate of Compliance contain commitments to prepare and validate procedures, and to train qualified personnel in their use so that spent fuel can be retrieved safely from a dry storage cask.

The NRC staff agrees with the petitioner's premise that actually unloading a storage cask would likely result in licensees learning lessons that could improve unloading procedures. The staff does not agree that additional demonstration of the unloading procedure is warranted. In addition to the NRC staff's review of the procedure for unloading casks, reasonable assurance that the casks can be sat ely unloaded is provirled by a variety of experiences related to the use and storage of radioactive materials. These experiences include preoperational tests and dry-run exercises that are performed to verify key aspects of unloading procedures for casks; related research sponsored by the commercial nuclear industry, the U.S. Department of Energy, and the NRC; actual loading and unloading of transportation casks; loading of storage casks; handling of spent fuel assemblies under various conditions; and performing relevant maintenance and engineering activities associated with reactor facilities. In addition, as discussed below, there is recent experience from unloading a spent fuel storage cask at Surry. Accordingly, the NRC believes that the request of the petitioner and some commenters' to require a demonstration of cask unloading before a cask can be used at an ISFSI is unnecessary. The NRC staff also believes that adequate assurances are in place to ensure safe cask unloading. As part of the review described in NUREG-1536, the NRC staff verifies that the SAR has requirements for cask unloading procedures. The NRC inspects procedures, training and qualification, and ISFSI operations. Further, requiring a full demonstration of cask unloading could result in unnecessary radiation exposure to workers and the public. The NRC staff's view that adequate assurances are in place to ensure safe cask unloading are borne out by the practical experience in retrieving dry storage casks that have been stored with spent fuel for a number of years. In 2000, two TN-32 spent fuel storage casks at Surry were retrieved from the storage pad because of indications of a failed seal. In one 11

case, the seal monitoring system had developed a leak. The cask was returned to the pool, the seals replaced, the monitoring system repaired, and the cask leak tested. The cask was then returned to the ISFSI pad. The second cask had a leak in the secondary seal. The primary seal was intact. The cask was returned to the pool and the lid removed to replace the seals. Localized corrosion was discovered on the sealing surface of the lid. The fuel was unloaded while repairs were made to the sealing surface. After the sealing surfaces were restored and the seals replaced, the cask was reloaded, leak tested, and returned to the storage pad. During these operations, no releases of radiation to the environment occurred and no spent fuel degradation was found. These two casks were initially loaded and placed in storage in 1996. More information can be found in NRC Inspection Report 72-002/2000-06. The petitioner believes that the NRC has not evaluated phenomena such as high-temperature zinc reactivity and thermal shock that will allow site personnel very little time to evaluate the situation and initiate corrective actions. The NRC staff reviews areas such as thermal loading, inadvertent criticality, and structural or containment failure for normal and abnormal conditions that are addressed by the designer of the storage system. NRC places thermal load limit restrictions on casks approved for use and requires that fuel be stored in an inert atmosphere. Although no adverse effects of zinc on the cladding of the spent fuel stored in NRC certified casks have as yet been identified, NRC has initiated a research project to investigate the possible effects of zinc on spent fuel cladding. The NRC staff believes that the petitioner has identified a valid concern regarding the potential recovery of fuel assemblies that unexpectedly degrade during storage. However, in this unlikely event, the NRC staff has concluded that there is reasonable assurance that a licensee can safely unload degraded fuel or address other problems. This conclusion is based on the NRC's defense-in-depth approach to safety that includes requirements to design and 12

operate spent fuel storage systems that minimize the possibility of degradation; requirements to establish competent organizations staffed with experienced, trained, and qualified personnel; and NRC inspections to confirm safety and compliance with requirements. The NRC staff finds acceptable these procedures for detecting degraded fuel through sampling and, on the basis of the sample results, the implementation of appropriate recovery provisions that reflect the ALARA (as low as is reasonably achievable) requirements. The NRC staff's acceptance of this approach is based on the fact that the spent fuel storage cask can be maintained in a safe condition during the time needed to develop the necessary procedures and to assemble the appropriate equipment before proceeding with cask unloading. The NRC staff also relies on the considerable radiological safety experience available in the nuclear industry in its 13

assessment that appropriately detailed procedures can be prepared for the specific circumstances in a timely manner. For the reasons cited in this document, the NRC denies this petition. Dated at Rockville, Maryland, this 18th day of January , 2001. For the Nuclear Regulatory Commission. William D. irave( Executive Director for Operations. 14

\\ P.O. Box 174

• Lake Elmo. MN 55042
• Phone:

770-3976 KET NUMBER PETITION RULE PAM -r~-4 (t,3FI( I ~O'fo) "98 JUN 23 P 2 :52 June 18, 1998 Secretary, U.S. Nuclear Regulatory Commission Washington, D:C. 20555 Attention: Rulemaking and Adjudication Staff

Subject:

Reply comment regarding petition for rulemaking

Reference:

10 CFR Part 72, Docket No. PRM-72-4

Dear Rulemaking and Adjudication Staff:

This reply comment is in response to comments submitted to the Nuclear Regulatory Commission inthe-above-eapt-ionedmatter by Transnuclear Inc., the Nuclear Energy Institute, and by extension, Northern States Power Company. Michael E. Mason ofTranSJIDcleaT Inc: purports in his May 26; 1998 comment that the seal replacement condition in the NSP ISFSI SAR (Table 5.1-2; Exhibit G of the present PIC Petition for Rulemaking} is not really a condition of the NS-P ISFSI SAR If Mr. Mason is correct, why go to the bother of issuing licenses? IfMr. Mason is correct, any licensee-or technology vender can decide-for itself which operating, maintenance or fabrication conditions and procedures it wishes to conduct, apply or prepare for, and which ones it chooses to ignore-. This permit eondi-tion contains hone of the ambiguity purported by Mr. Mason. It says, point blank, and with no caveat: Major Maintenance ( once in 20 years)

1. Replace-cask lid seals In a way, this assertion by Mr. Mason makes our point. Maybe cask seals will not need replacing after 20 years. Maybe they'll last for 40. Then again, maybe they'll only last for 10 or 15 years. Mr. Mason can speculate, but he doesn't know, and neither do you. That's the problem. Helium, being a very small atom, will, over time, leak out of any container into which it is put. At some point, therefore, either TN-40 casks must be unloaded, or else the seals must be-replaced. Thus the SAR condition. Rules need to be promulgated to ensure that when these necessary functions are performed, they are performed safely, because rules governing procedures for safe management of contingency conditions when these functions are performed do not presently exist. If licensees are allowed to operatewithout knowing proper procedures to perform necessary functions, the NRC is irrelevant from a health and safety perspective, and amounts to little more than a public relations ploy of the nuclear industry.

Adm WI ad JUN 2 5 1998-

UCLEAR REGULATORY COMMISSION AKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSIO Document Statis'&ics Postmark Ca Caples ce A s

As Mr. Mason is in denial about the potential for a variety of phenomenon to cause degradation to waste fuel assemblies over time, it is not surprising that he compares reloading a storage cask with pool water to reloading a transportation cask with pool water. His comparison is meaningless, however, because metal creep, liquid metal embrittlement, neutron flux embrittlement, microbiological corrosion, high temperature zinc reactivity to water and perhaps other degradation modes can impact the integrity of the waste during the storage period. The whole point of this rule-making proceeding is that lack of actual knowledge about how waste materials will behave during storage and unavoidable management operations makes contingency planning necessary in order to protect against worker and public radiation exposure likely to occur if contingency procedures are not in place. Please bear in mind that, contrary to Mr. Mason's imaginings, phenomenon such as high temperature zinc reactivity and thermal shock will allow site personnel very little time to evaluate the situation and initiate corrective action. Mr. Mason states that, "Cask licensees are required to demonstrate the ability to retrieve spent fuel under accident conditions, which are more severe than the long-term normal conditions." He goes on to assert that therefore, the fuel is readily retrievable when required. Again, Mr. Mason misses the point that in reality, neither he nor anyone else knows what "long-term-normal conditions" actually are, or whether they will tum out to be more or less severe than "accident conditions." lfhe does know, where is the analysis? Where is the peer review? What are his assumptions, other than an incredible "no degradation" assumption? What are his bounds of probability that there will be no degradation. What is the probability, and with what confidence can it be objectively stated that cask management events will actually occur as he says they will? These questions remain unanswered, and thus the need for rule-making to develop-safe procedures that cover contingency situations in the event that waste degradation occurs in ways not contemplated by Mr. Mason. The* INEL situation was cited in the petition simply to demonstrate that waste degradation does occur, and that such degradation can result in situations in which existing rules are not adequate. If Mr. Mason is so confident that a TN-40 can be safely unloaded, one would think that he would readily agree to a demonstration of safe unloading procedures-. This would seem to be particularly true considering how demonstrating safe unloading procedures would impact the pub-lic's perception of the industry's ability to safely manage waste assemblies. Instead, however, he attempts to obfuscate the situation with meaningless comparisons to-transport casks. Unfortunately, time is itself a fundamental and dynamic variable that will determine how waste assemblies behave at any specific point in time: Therefore demonstrating a safe unloading procedure after 10 years of dry cask storage may mean little or nothing in terms of an ability to safely unload a cask after 20 years. Hence, again, the need for rule-making to plan for safe management of contingency situations. In its comment dated May 20, 1998-, the Nuclear Energy Institute (NEI) purports that, "... there is no generic issue regarding safely unloading dry spent fuel storage casks, and hence no need for generierulemaking." Like Transnuclear, Inc., Lynnette Hendricks ofNEI bases her conclusion on the belief that waste assembly degradation will not occur

during storage. She believes there is no significant difference between unloading waste never stored in dry casks from transportation casks, and unloading waste from storage casks after a period of dry cask storage. She pretends that because NRC rules require licensees to be able to retrieve irradiated fuel, such an ability is automatically bestowed on licensees without any need to consider procedures that would protect worker and public health and safety in the event that possible, and perhaps very likely contingency situations occur. Ms. Hendricks concludes that because the NRC decided not to suspend NSP's authority to operate a dry cask storage facility by denying two recent petitions, the NRC should have no interest in promulgating cask management rules capable of handling contingency situations. As the NRC denied those two petitions largely on the same set of unfounded beliefs and self-serving assumptions that are contained in the present NEI and Transnuclear comments, she may be right on this point. But her conclusion doesn't necessarily follow, and one would hope that given enough opportunity, the NRC eventually will decide to fulfill its legal obligation to protect worker and public health and safety. This obligation cannot be met without conducting a rule-making proceeding that specifies procedures that will enable on-site personnel to safely manage high-level nuclear waste in a variety of contingency situations that may occur, considering the potential of various phenomenon to degrade dry cask storage systems. In conclusion, it is interesting to note how on the military side, the nuclear industry is intent on the continuous testing of various weapons and weapons systems, because there is so much they don't know about how such systems and devices degrade over time. Yet on the civilian side, even though there is no experience to draw upon, the nuclear industry is perfectly content to rely on speculation, extrapolation, and the same type of blind faith that drives religious fervor, to justify behavior that assumes that degradation will not occur to waste storage systems. Sincerely, .)-/_ c~~~ ---'/6~ George Crocker Steering Committee

(507) 664-0252 June 18, 1998 Secretary KET U BER PETITIO fllE PR 1~-¥ (t,3Fl<. l~o'IO) OVERLAND LAW OFFICE Carol A. Overland Attorney at Law 402 Washington Street So. Northfield, MN 55057 Attn: Rulemakings & Adjudication Staff U. S. Nuclear Regulatory Commission Washington, DC 20555 RE: Rulemaking Docket: PRM-72-4

Dear Rulemakings & Adjudication Staff:

DOCKtTED US,'.RC 098 JUN 22 AlO :18 OFF11. RU' J_1-_,

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u (507) 664-0253 Fax When considering the merits of the please also consider the issues Research Institute in their April To receive this report, call (510) rulemaking petition before you, raised by the Electric Power 1998 report, number TR-108757. 934-4212. This report was issued to promote extension of the 20 year licensing period to 100, a problem in its own for another day. However, it does address several problems also addressed by the Petition, such as fuel and cladding oxidation due to a small cask leak, cladding rupture, and hardware peformance, utility promotion of storage of assemblies with higher burnups, and the zinc problem a la the Explodo Cask at Pt. Beach, although in this report it considers only the zinc and its interactions with fuel and cladding and not the logically explosive reaction of zinc and boric acid. Pay particular attention to those 11 data needs II and 11potential research priorities." Although this is an industry publication, you will likely find that the "data needs and "potential research priorities" cover many of the same issues that evidence the need for rulemaking requested in this Petition. The issues raised by the petition and by the EPRI report are universal to all casks and demand rulemaking. Very truly yours, Carol A. Overland Attorney at Law

ll.S. NUCLEAR REGULATORY OM 1 RUL.E:MJOONGS ADJUD CATI NS AH-OFFIC O THE SECA OF THE COM ISSIO

CAO:dc

NOTE TO: FROM:

SUBJECT:

June 19, 1998 Emile Julian Chief, Docketing and Services Branch Carol Gallagher ADM.DAS ~~ DOCKETING OF COMMENT ON PRM-72-4 Attached for docketing is a comment letter related to the PRM-72-4. This comment was received via the interactive rulemaking website on June 18, 1998. The submitter's name is Carol A. Overland, 402 Washington Ave. So., Northfield, MN 55057. Please send a copy of the docketed comment to David Meyer (mail stop T6-D-59) and Stan Turel (mail stop T9-F-31) for their records.

Attachment:

As stated cc w/o attachment: S. Turel

DOCKET NUMBER PROO. & lJTIL FAc. _2;(?;:;,f,, *~ (lP3F~ I ;;o'/O OVERLAND LAW OFFICE Carol A. overland Attorney at Law 402 Washington Street So. Northfield, MN 55057 DOCICltiD JUfM 1 5 1998 iU.EMAKINGSANO ~ STAFF 860\\'-fHl ,, / / (507) 664-0252 (507) 664-0253 Fax June 11, 1998 Secretary Attn: Rulemakings & Adjudication Staff U.S. Nuclear Regulatory Commission Washington, DC 20555 RE: Response to Comment of Transnuclear, Inc. Rulemaking Docket: PRM-72-4

Dear Rulemakings & Adjudication Staff:

Enclosed please find a copy of Table 5.1-2 from the NSP's ISFSI SAR, Docket 72-10. I have reviewed the comment of Mr. Mason, Transnuclear,Inc., and raise the following response for consideration: Use of Ms. Shankman's letter is not indicative of any misunderstanding regarding seal vs. weld. Ms. Shankman's letter addresses the concern of NRC regarding release of helium and potential degradation. The effect of a failed seal and failed weld are the same if either fails, helium is released with degradation a likely result. Mr. Mason himself admits that the TN-40 monitoring system "virtually guarantees that a helium atmosphere will be maintained in the cask." Comment of Mason, p.l. "Virtually?" And in discussing the seal replacement, his estimation of the potential changes to "unlikely. n But as we learned with VSC-24 welds, the potential for compromise of the helium atmosphere is great. Mr. Mason goes on to say that: Failure of the seals would not result in air entering the TN-40, thereby causing fuel degradation, but rather leakage of helium from the overpressure system into the cask. Comment, p. 1. There are two seals, inner and outer, and if they failed, the pressurized helium would leak into the atmosphere. Mr. Mason, the Prairie Island Coalition, Ms. Shankman, and myself are in absolute agreement that maintenance of helium is a necessary I

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• .:: l:J~l.l U.S. NUCLEAR REGULAiORY COMMISSIOti RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OFlffE SECRETARY OF THE OOMMISSION Doc:IIUltSt!dlslk:8 Postmmk ll8111."'ll ~ 1, ~ i-fi'tt -°"fr//<-- &,_,,L ~

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comment that: [s]toring the fuel in a helium environment and at the relatively low temperatures discussed above assure that the condition of the fuel at the onset of storage is maintained throughout the storage life, and thus the fuel is readily retrievable when required. Id. Can that helium environment be maintained? When/if the seals fail, it is immaterial whether the seals are metallic (TN-40) or a weld (VSC-24/Transtor). The fuel is, as Ms. Shankman's letter states, subject to degradation, which is unacceptable. If it is degraded, how will it be readily retrieved? If it cannot be retrieved, the fuel must not be loaded into casks. Mr. Mason also states that the misapprehension regarding maintenance states: Petitioner is under a of the TN-40 seals. He [T]he reference to the NSP SAR "requiring" seal replacement seems to be a misunderstanding of the TN-40's dual seal and monitored overpressure system. Id. Please refer again to Table 5.1-2 of NSP's ISFSI SAR, which lists "~nticipated Time and Personnel Requirements for Cask Handling Operations." Under "Major Maintenance" is the notation "once in 20 years" and "Replace cask lid seals." The task will require 3 persons, 1950 minutes, total time to transfer cask to spent fuel pool, replace lid seals, and return cask to ISFSI pad." There is no misapprehension -- Mr. Mason is likely stating that maintenance is not necessary because the required physical task of changing a seal is presenting a problem to NSP and Transnuclear -- it simply cannot be done. Yet seal maintenance must be completed. Any idiot knows that seals must be maintained and that they are subject to failure. If they were not subject to failure, a complex monitoring system would not be necessary. UNDER THE SAR, THE CASK SEALS MUST BE REPLACED EVERY 20 YEARS AS MAJOR MAINTENANCE. The seals must also be replaced when/if they leak. NSP and Transnuclear, and the NRC as well, cannot ignore the SAR. In December, 1994, when I asked, during a tour of NSP's facility, what NSP's plans were for seal maintenance, I was told by Michael Wadley that they had no plan. One month later, I was told in a public meeting that the seals have to be replaced every 20 years under the SAR. When I asked how this would be done, I was told the plan had not been completed. The "unloading plan" was subsequently developed and approved, and the casks were loaded, but as we know, the unloading plan, essentially the reverse of loading, is inadequate. Addressing thermal shock, Mr. Mason states that "Thermal shock will not present a significant problem when the casks ar reloaded with pool water prior to unloading. Unloading of spent fuel is typically part of a fuel transportation cycle... Unloading of fuel

from a transportation cask into a spent fuel pool without causing fuel degradation has been demonstrated in both the U.S. and extensively in France at the LaHague reprocessing facility." However, there is no record of any demonstration of the unloading of a cask that has been used for storage of irradiated spent fuel for anylength of time. I put Mr. Mason to his proof -- produce the documentation of unloading irradiated spent fuel that has been stored for a great length of time. For years NSP has been making this claim, but then it disingenuously cites the TN-24P study! The TN-24P study is not proof of unloading capability. Mr. Mason dismisses the problems in unloading the TN-24P, and offers that the TN-24P prototype was not reviewed or licensed by the NRC, and that canisters were used. However, Mr. Mason fails to explain why metal canisters would warp under the conditions of dry cask storage (short term in that case) and assemblies would not! Further, as we all know, NRC reviewal and licensing does not guarantee adequate construction or safety. The NRC only sets standards and reviews and inspects licensees to assure they meet these standards. Federal Tort Claim of General Public Utilities Corp., et al., 13 NRC 773, CLI-81-10 (1981). Note that Mr. Mason makes the claim that "Transnuclear knows of no case in which retrieval of spent fuel assemblies from transport casks, which are very similar to storage casks, has been a problem." He cannot make that direct claim for storage casks, because as we all know, no storage cask in use for any length of time has been unloaded. Non-irradiated fuel that has been transported to the plant and spent fuel transported for a brief time in 3-assembly casks are quantitatively different. Similar is not close enough. Transnuclear's smoke and mirrors comment begs the question. The Petition is calling into question the standards and requesting rulemaking to address obvious cask problems universal to all casks and to those problems specific to the TN-40, and a rulemaking proceeding is necessary. Very truly yours, ~,4,~ Carol A. Overland Attorney at Law CAO:dc

ISFSI SAR TABLE 5.1-2 (Continued) ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS Operation Storage Area

23.

Unload from vehicle position in location (Dl, D2, D3)

24.

Check surface dose rate (D6)

25.

Connect pressure instrumentation (D4, D5) Periodic Maintenance

1.

2.

3.

4.

Visual surveillance (NA) Repair surface defects (NA) Instrument testing and calibration (NA) Instrument repair (NA) No. of Personnel 5 5 5 2 2 2 2 3 Time (min) 60 30 30 15 60 180 60 1950** Avg. Distance (ft) from Cask 5 3 5 5 3 5 3 8 No measurable dose associated with this activity. Therefore, the number of personnel, time and distance are not significant. Parenthetical information corresponds to Table 5.1-1 activity numbers. Total time to transfer cask to spent fuel pool, replace lid seals, and return cask to ISFSI pad. TABLE 5.1-2 REV. 2 9/91

DOCKET NUMBER TIO RU PR 7 :l-L/ ( '131=~ l'-o'l-o) DOCKETED USNRC Acknowledged by card JUN - 4 9 8

tJ.S, NUCLEAR REGULATORY CO nu!LJC"""""'NGS & ADJUDICATIO OFFICE OF THE SECRETAR 0 THE CO MISSIO ~/:s/&J'il 1?~ ~ ~evu(_ 11?~ I ~ ~~ ~~ /t).s

DOCKETED us. me David L. Meyer, Chief .98 JUN Rules and Directives Branch -3 Pl :46 Division ofAdministrative ~~ryices Office of Administration F,, United States Nuclear Regwlr_atory Commission Washington, D.C. 20555-001

Dear Mr. Meyer,

DOCKET NUMBER PETITION R PRM 7.2- '/ {t.,3F~ l~O'/-o) r.. r:, 1998 JUN - 2 P'-1 2: I 3 RULES 8, Oii-. 13HL\\NCH US NRC -,:.f [£--'(.(ov Jc, c µ./,.,: j I (, ( 2.-(c; J ]x_yU~e/ The following comments are in reference to Docket Number PRM-72-4, Prairie Island Coalition's petition for rulemaking. The form and content of C.U.R.E.'s comments have been primarily shaped by the NRC public communication documents from the NRC Strategic Assf.!Ssment and Rebaselining initiative as well as the documents available on Public Involvement in the Nuclear Regulatory Process available on the NRC website. C. U.R.E. is a grasssroots citizen group which advocates for responsible nuclear waste storage. We have found the NRC documents clear and useful. They have aided our understanding about how the public can be more involved in nuclear waste policy and regulation. Misunderstanding of public concern often results of from a failure to recognize that it is the long-term risks and costs associated with nuclear waste storage that are primary to citizens. The interests associated with this temporal perspective are quite different than those that are central to the utility or even to regulating agencies. Issues of maintenance, monitoring, management and funding, health and safety - all look different from a long-term perspective. It is imperative that the 'long term' begins to be more effectively factored into cost-benefit, health and safety, risk and performance standards. r TT P P <'nnnr.rtc rnlpmr:ilr;no th-:it,u,11 f~r-1l;t-:it-<> or<><>t<>r rP011lr:itnr" -_,.~ *.&."".&....J* "--4>,t-'~.._,..,... w ......_......,._.....a.~.a..~........ b ........ _.......,._.._..._....._ __........................ _::,... --... -.a.. ........,-J coherence between commercial reactor operation, decommissioning and storage; also within the nuclear waste storage framework. Uncertainties increase risk. We feel that the purpose and effect of this particular rulemaking should be to reduce the uncertainties surrounding degradation, loading and transfer of nuclear wastes and should provide for the development of contingency analysis of the interaction between storage timelines, technologies and degradation factors. NRC should use the full extent of resources available to the agency from public, professional, governmental and industry sources. 'U - 4 9 8 by

LJ. NUCLEAR REGULATORY COM, SION AULEMAKINGS & ADJUDI ATI ~ STAFF OFFICE OF THE SEC EI OF THE COMMISSIO

In addition, C.U.R.E. encourages NRC to shift from a 'confidence' based to risk-informed and performance-based analysis and decision making in waste storage regulation. In its committment to the best and most efficient use of public resources NRC should consider ways of bringing the public into partnership, as it has found ways to partner with the industry in efforts to fulfill its mission. We hope to see an increased use of non-adversartial models such as those mentioned as part of NRC's "enhanced participatory rulemaking", We hope that NRC will consider building into its rulemakings, avenues for public input that enhance rather than drain the resources of the agency. Please include this letter and the questions which follow our comments in your on line service. It is not necessary to include the body of our comments or the appendix which we developed in response to your SDI paper on Public Communications Initiatives. I can be reached at 612-331-1430 or (612) 345-5488

~);frN~

Kristen Eide-Tollefson for C. U.R.E. Communities United for Reponsible Energy* p.o.box 130 Frontenac, MN 55026

Following is a list of questions that members of C.U.R.E. helped to develop for a related forum, around the questions of degradation and its impact on unloading, transfer, transportation and storage strategies. We hope that this rulemak:ing will either address these questions or bring us closer to being able to address them. Thank you

1) For such purposes is it desirable that a distinction be made by NRC regarding standards for the integrity of cladding in:

1) storage and

2) any transfer of materials in case of emergency or need to transport the fuel.

2) Under what circumstances could failure of integrity of the cladding affect safety?

1) What is the potential for criticality a) under long term storage conditions, b) in combination with transfers of fuel in case of accident, cask breach or for transportation.

2) How might a disintegrated state of the fuel affect worker exposure during transfers under circumstances ref erred to above?

3) How could disintegrating cladding produce unanticipated conditions or temperatures in otherwise normal storage conditions?

a) What might these conditions be? 4)What are the factors that could produce criticality conditions? 5)How will these be anticipated, addressed, planned for? NRC analysis NRC rulemak:ing Utility regulation

3) Will DOE accept degraded fuel at a permanent repository?

How will this concern be investigated? How will NRC responsibilities for not only short, but long term Health and Safety a) How will DOE determine condition of materials prior to acceptance and/or transportation of fuel? b) What criteria for 'acceptable' condition of fuel might DOE have; or could NRC recommend? c) If fuel is not 'acceptable' what will happen?

In addition, C.U.R.E. encourages NRC to shift from a 'confidence' based to risk-informed and performance-based analysis and decision making in waste storage regulation. In its committment to the best and most efficient use of public resources NRC should consider ways of bringing the public into partnership, as it has found ways to partner with the industry in efforts to fulfill its mission. We hope to see an increased use of non-adversartial models such as those mentioned as part of NRC's "enhanced participatory rulemaking", We hope that NRC will consider building into its rulemakings, avenues for public input that enhance rather than drain the resources of the agency. Please include this letter and the questions which follow our comments in your on line service. It is not necessary to include the body of our comments or the appendix which we developed in response to your SDI paper on Public Communications Initiatives. I can be reached at 612-331-1430 or (612) 345-5488 Thank you, Kristen Eide-Tollefson for C. U.R.E. Communities United for Reponsible Energy* p.o.box 130 Frontenac, MN 55026

REQUEST FOR RULEMAKING BASED UPON NRC DOCUMENTATION OF AGENCY'S STRATEGIC ASSESSSMENT OF REGULATORY ACTIVITIES: The request for rulemaking outlined in Docket No.PRM-72-4 Prairie Island Coalition, Receipt of Petition for Rulemaking was based, in part, upon citizen review of NRC documentation of its Strategic Planning Framework - Strategic Assessment and Rebaselining project (September, 1996-). It was inspired by NRC's acute observation of the exacerbating effects of public mistrust upon the many uncertainties of nuclear waste management. This rulemaking petition is a specific response to NRC's inquiry into the agency's role: "how it may be able to effect change to reduce the uncertainties". It is, in some sense, a test of NRC's willingness and ability to effectively engage specific uncertainties of persistent public concern: the relationship of degradation to the ability to safely unload, transfer, transport and store nuclear waste. This rulemaking is needed to provide a direct connection between NRC requirements and concerns about fuel integrity, and the procedures necessary for monitoring, retrieving, repairing, maintaining, as well as assuring effective and safe tranfer and transportation of dry cask storage systems. The responsiveness of regulators and utilities to public demands for procedural and technical integrity will be of increasing importance to the success of interim storage strategies. In order to maintain the bases of its regulatory authority* ensure that the uncertainties of storage term and technologies do not undermine the Ion~ or short term public health and safety, and maintain and promote public trust and confidence, C.U.R.E. encourages NRC to proceed with this rulemaking. The stakes seem high for communities and public interest groups who are monitoring the viability of various storage schemes. Without mechanisims that provide standards for ongoing research, development, maintenance, refitting and backfitting and of the necessary technologies, current storage sites face the specter of becoming defacto 'permanent' storage sites. Especially in a dergulated environment they could find themselves underfunded and ill-prepared to meet any long term storage challenge. These concerns further suggest the necessity of addressing certain strategic challenges, fundamental to the dry cask storage enterprise:

• NRC ability to guarantee decommissioning and restoration of sites.
• Can public concern regarding the uncertainties of degradation, retrieval, and unloading of spent nuclear fuel in dry storage casks, be effectively addressed without considering a broader analysis of the interactive contingencies of containment materials, term of storage, monitoring, transfer and transportation technologies? If not at the federal level, where will such essential analysis be carried out?
• Can such an analysis be effective without applying rigorous and scientifically based risk and performance base standards, such as are applied to the technical details of nuclear operations? Can the extensive materials testing and research done during the period of reactor development of the l 970's be utilized in the further development of storage technologies?
• Given the fact that the nuclear fuel storage enterprise is, in effect, a vast experiment, what kind of rules, proceedings and standards will ensure that :

1) the public is protected from physical, psychological or economic exploitation (key: public participation);

2) regulation can keep up with changing circumstances; regulation will be able to meet the challenges of decommissioning and waste storage in a deregulated environment;

3) funding will continue to be available for the long term challenges of responsible nuclear waste management.

Since this rulemaking request arose, in part, from observations made upon NRC's strategic assessment papers, some of these observations will be incorporated into our comments on the petition for rulemaking submitted by Prairie Island Coalition. References will be to the goals and strategies sections in the introductory and background pieces. We will not be referring to the options outlined in the strategic assessment issue papers unless it is to clarify our own comments. This rulemaking request and comment also utilized NRC document: Public Involvement in the Nuclear Regulatory Process available at nrc.gov/OPA/gmo/tip/publicin.htm#intro.

2.

These comments are submitted by C.U.R.E. Communities United for Responsible Energy, is comprised of citizens from Florence Township, Lake City and surrounding communities in Southeastern Minnesota. We have been involved in nuclear waste issues since February of 1995 when NSP announced that it had chosen alternative sites in Goodhue County for Prairie Island Waste, as directed by the 1994 Minnesota Legislature. C.U.R.E.'s mission is advocating for responsible long and short term management of nuclear waste. C.U.R.E. also advocates for increasing the role of public participation in permitting and policymaking proceedings that involve nuclear waste management. (Please see Appendix I & II) C.U.R.E.'s advocacy of NRC rulemaking on the petition at hand, is based upon the following convictions: I. Nuclear fuel will degrade; so will containment materials. II. Public fears are not irrational. III. Environmental isolation for nuclear wastes should continue to be a priority. IV. The goal is -socially, environmentally, technically, economically-responsible nuclear waste management. V. Successful regulation requires a balance of interests, of long and short term goals. VI. Technical advances alone will not 'solve' the problem. VII.. The national nuclear waste predicament requires that both the public and the industry become more accountable; NRC is currently in the best position to facilitate. VIII. We need to create innovations in public policy and government regulation that will allow for timely integration of new information, and best utilize the resources of all stakeholders in order to meet the long term demands of responsible nuclear waste management. C.U.R.E.'S EXPERIENCE OF UNCERTAINTY FACTORS IN NUCLEAR WASTE SITING AND QUESTIONS OF LONG TERM MANAGEMENT During the Minnesota Environmental Quality Board Citizen Site Advisory Task Force process, the citizens involved grappled with a number of the fundamental uncertainties of nuclear waste storage. They made a recommendation to the state agency that a timeline be developed as part of a scoping mechanism. This timeline was intended to evalute items of potential "impact" such as cost, transportation, health and safety risks. "The timeline is intended to 'scope' both known and,as yet, unknown dimensions of potential impacts of the storage of high level nuclear radioactive waste "for an unknown duration".

3.

The subcommittee preparing the recommendation required that the timeline extend to at least 1,000 years or @ 1/10 the half life of radioactive materials that must be isolated. "This (1,000 year) extension should include, at least, a charting of the information about the known and conjectured effective life of radioactive and containment materials: transformations decay etc." - Legislative and Regulatory Review in appendix to Report of the Site Advisory Task Force Goodhue County Dry Cask Storage Alternate Site Project January, 1996 A contingency analysis of fuel and containment degradation along a timeline was an essential feature of the scope. Despite Task Force recommendations to MEQB, we found that it is hard to create such an analysis at the state level. Pressures from the industry not to deal with problems that may draw public attention or concern on nuclear waste storage issues is strong at all levels. While this is understandable, carried too far, cooperation with such resistance constitutes a betrayal of the public trust and makes regulating agencies less effective than they need to be to meet challenges that do not answer to political pressures, such as the physical realities of fuel and materials decay. Without the development of contingency analyses, the unavoidable uncertainties of long term nuclear waste storage are certain to create hazards to public health and safety - over the long term. Without such information, technical waste management research and development will become prohibitively expensive as it tries to cope with management strategies that are insufficiently informed about the basic materials that they are utilizing. NRC HAS IMPORTANT ROLE IN REDUCING UNCERTAINTIES: The uncertainties of long term storage and utility, state and national storage schemes are a prime source of pubic concern and distrust. The DSI 6 paper on High-Level Waste and Spent Fuel suggests that NRC has considered the possibility that greater clarity and proactive initiative on its part could play a significant role towards reducing uncertainties and enhancing progress (p. 14). Indeed, from the public perspective, there may be no other road to resolution. NRC should focus upon opportunities to engage issues that can, if handled properly, serve to reduce pivotal uncertainties. This rulemaking provides an excellent opportunity to bring clarified NRC Strategies, goals and principles to bear upon the persistant public concerns represented in PIC's petition.

4.

CONFIDENCE BASED DECISION MAKING INADEQUATE: The strategies of 'confidence' based decision making upon which NRC has heavily relied in what it sees as 'lower risk' (non-operations) regulatory catagories, tends to dismay and infuriate a public which is struggling to responsibly inform itself, and grapple with complicated and technical details. Decisions which rely upon 'confidence' (in past 'confidence decisions'; in utility and technological ingenuity etc.) will not assuage public concern. Avoidance has not been effective. Reassurance has not been effective. Discounting of public fears and concerns provokes frustrated reactivity. APPLYING RISK-INFORMED, PERFORMANCE-BASED REGULATION: In discussing its philosophy of "Risk-Informed, Performance-Based Regulation" in Strategic Assessment Issue Paper DSI-12, the staff considers the application of this philosophy in materials handling/waste disposal situations (e.g. as in this rulemaking). For the purpose of this and other nuclear waste handling and storage standard development and rulemaking procedures, NRC's suggestion that it apply higher standards of a risk and performance based model that have formerly been reserved for nuclear operations, is highly advisable. The need for a more systematic way to deal with long term, economically viable nuclear waste management strategies, is critical. NRC's role is key. The value of regulatory coherence is central to a satisfactory application of any model. Regulatory coherence would itself contribute a great deal to reducing the 'uncertainty' which surrounds nuclear waste management issues. The public participation and communication initiatives should be involved in the application of risk/performance models in order to ensure that a more stable, systematic and effective regulatory environment is the result. We hope that this rulemaking will provide an opportunity for using risk-based and cost-benefit criteria to engage the areas of public concern articulated in PIC's petition. Properly applied risk-informed and performance-based models is likely to radically improve public perception of "reliabiliy" in NRC decisionmaking - as defined in NRC's Principles of Good Regulation. In addition, since the public can be counted on to identify specific issues in the waste conundrum that most need addressing, public pressure can be a resource, helping NRC to identify priorities in regulatory activity, while enhancing its credibility as an agency that is responsive to

5.

public concern. The purpose of utilizing such models should be to more effectively engage, not to avoid, issues where public perception of risk is persistant. For the purpose of applying risk-informed, performance-based models to the ongoing creation of standards and rules for nuclear waste storage:

1) NRC's "Defense in Depth" policy may need to be adjusted to the specific concerns of long term isolation of nuclear wastes from the natural and human environments. Such an application of the Defense in Depth philosophy to long term nuclear waste management must consider as primary catagories of analysis:

a) The natural and necessary degradation of both fuel and containment materials b) The timeline along which such degradation may either increase and/or eventually reduce the hazards of exposure, release etc.

2) Industry resistance to the application of such models and standards should be balanced by serious consideration of items which the public has identified as an area of risk and concern.

Economic &/or regulatory incentives might be engaged to encourage the cooperation of the industry.

3) Timelines for long-term and short-term considerations should be built into all risk and performance assessments. One source of misunderstanding of public risk perception is the failure to recognize the concerns that arise for the public over the long-term health and safety effects of nuclear waste maintenance & storage.

4) The differences between the cost and benefit interests of the industry and the public should, by extension, be clearly articulated in any cost and benefit analysis. The agency need not become immediately entangled in the perception that it must take one side or another, but should engage a variety of investigative strategy.

PUBLIC CONCERN WITH LONG TERM 'COSTS' The public has a primary concern about the long term viability of storage models. Therefore the public is especially concerned about the natural course of degradation, and the funding of monitoring, maintenance and technical systems that will assure long term isolation of nuclear wastes from the natural and human environments. The long term 'costs' to the public are more difficult to evaluate than the short term business strategies that drive the industry. Because of NRC's mission, to "ensure adequate protection of the public health and safety" it is essential that the agency find ways to evaluate and effectively incorporate long term 'costs' and 'benefits' into their analyses.

6.

THE PUBLIC AS A RESOURCE: C.U.R.E. does not have technical expertise to apply to this rulemaking, although it is committed to engaging the technical issues that are identified and elaborated by the petitioner and NRC. What we do bring to bear on the rulemaking process is our own experience in dry cask storage issues and a number of hard won insights about what kinds of processes are more or less effective in creatively engaging the public as a resource in nuclear waste storage issues. We feel that greater public involvement and accountability will be crucial to any effective nuclear waste storage strategies. Appendix I contains a selection of our comments upon OSI 14, the NRC strategic assessment issue paper on Public Communication Initiatives. The essence of these observations is that, given NRC's description of the regulatory climate and resource limitations it seems time to shift NRC' s perspective on public participation to recognize the public as a significant resource factor. NRC descriptions of "enhanced participatory rulemaking" is an excellent example of engaging public participation to increase the range of insights, information and resources available to the NRC in decisionmaking. CONSIDER NON-ADVERSARIAL AVENUES FOR PUBLIC INPUT CURRENTLY AVAILABALE ONLY BY PETITION: We hope that, along with this rulemaking, NRC will consider that the rapidly changing nature of nuclear waste management issues calls for avenues of public input other than the petition format. We understand that the processing of 2206 petitions is perceived as a burden and a distraction to NRC staff, who are already overwhelmed.

1) When public issues arise, how can NRC use resources already in place such as the public liason resources of the Office of The General Counsel or the OPA to engage public interest and debate?

2) How could the skills of the NRC be extended to more easily accomodate "enhanced" public participation by such methods as it has used in "enhanced participatory rulemaking"?

3) When does NRC utilize "enhanced participatory rulemaking"?

Will NRC be utilizing such methods in this rulemaking?

4) Is there any other structured public feedback mechanism that NRC could put into place that would allow NRC to respond from a less adversarial perspective, and engage public concern as a resource?

Could this be put into place as part of a rulemaking?

5) How can a clear and realistic articulation of differing interests enhance pubic trust and increase clarity in NRC decisionmaking &

communication?

7.

Communities United for Reponsible Energy* p.o.box 130 Frontenac, MN 55026

APPENDIX I: CURE COMMENTS ON: OSI 14-Public Communication Initiatives The Public as partners in regulatory concerns and process NRC increasingly depends upon the entities, primarily utilities, that it regulates as partners in the regulatory process. However, because of the nature of radioactive materials, it is

federal, state and local governments and the public who will bear the ultimate risks, costs, responsibilities and impacts of nuclear industry operations. They will bear any burdens not assessed to or planned for by the utilities for many generations to come. It is therefore appropriate, if not incumbent upon the NRC to review its relationship with the public in light of this fundamental fact, and to distribute its own resources to adequately protect long-term as well as short-term public interests and concerns.

It is to the advantage of the NRC as well as the public that the public be fully informed and involved; its concerns adequately developed and considered, and public interest and expertise recognized as a resource. Only opportunities for genuine public participation will strengthen public support and trust in NRC. NRC acknowledges this in the review of the BRC policy process where it lists, as "one of the principle lessons learned", to "provide an opporunity for public involvement at a time and in a manner that makes clear to all that they have the ability to influence the outcome". Support for Commission's preliminary views. The Commission recommends that NRC "place a priority on early identification of public concerns and methods for public interaction in making regulatory decisions that are likely to generate substantial public interest or concern"... In addition, the Commission recommends that "NRC interpret the term "public" in its broadest sense, understand who our various publics are, and focus on what they need in order to facilitate interaction and dissemination of information." We would like to see NRC address potential mutual benefits of public participation in the regulatory processes in its public policy statements. Regulation of the management, ownership, and funding of nuclear generation and waste operations is a crucial factor in the health and safety of present and future generations. NRC's strategic assessment and rebaselining project should take a proactive approach, in general, towards short and long term planning. While NRC may consider that it is 52

more difficult to assess public interests than utility interests, NRC should recognize that there is no better ally than the public itself in identifying public concerns. Right use of public interest and expertise, will increase rather than deplete NRC resources. Support for interactive models The issue paper correctly identifies significantly increase the importance regulatory activities: of public communication several of the factors that will of public communication in NRC "As deregulation heightens competition, the population of regulated facilities ages, and the waste disposition policy remains unresolved. public attention to NRC regulatory matters will continue and may increase". (p. 7, paragraph 3) However, it fails to adequately consider public participation as a resource in its portrayal of costs and benefits. The title of the paper [public communication rather than public participation] indicates the non-interactive bias current in NRC public policy. We support an interactive model for public communication and participation. NRC should regard public participation as a resource The definition of "public" put forward. by the Commission in its preliminary views is excellent. However, the definition of "public resources" reflected in the paper is one sided and shortsighted. When NRC speaks its "fundamental obligation"to attend to costs and benefits of approaches to public communication" (Discussions, III, p. 13), it clearly refers to the agency's obligations regarding efficient use of public monies and agency resources. This notion of costs and benefits fails to recognize citizen expertise and interest as a 'benefit' and a 'resource' to the government and its agency in the execution of its responsibilities to the 'public interest' and general welfare.A slight but signficant shift of perspective can clear the path of some of the obstacles that NRC sees as a limiting its options to involve the public meaningfully in its activities. A shift in perspective towards genuine public participation that NRC has begun to take, both clarifies and justifies NRC resource committment to public communication initiatives. Consider public agency" sources (p.16) as a cost effective atlernative for "non-of information and methods of distribution 53

CURE: WHO ARE WE AND WHAT ARE WE UP TO? For many years, the state of Minnesota, its public interest groups and the communities surrounding NSP's Prairie Island plant, have committed substantial resources trying to anticipate and respond to a rapidly changing environment on nuclear issues. We have struggled together and with each other to address a myriad of issues and items of public concern, generated by approximately 25 years of nuclear power operations in the state. These include the disposition, management and transporation of nuclear wastes; public health, safety, social and environmental impacts; and more

recently, deregulation, industry

merger, decommissioning and the economics of the nuclear industry.

A substantial amount of public participation and expertise has been one of the positive byproducts of this struggle. But another effect that this struggle has had upon the state is the alienation of communities, the embitterment of the legislative process, deepening wounds and division between the public interest and nuclear industry concerns, charges of environmental racism, and the fear and sense of persecution that have been an unintended but everpresent attendant of nuclear power throughout the nation. Our small community in the Southeastern corner of Goodhue County was one of the more recent, but not the last, to be afflicted. We responded to the crisis in 3 ways. l) EQB CITIZEN SITE ADVISORY TASK FORCE. Citizens from surrounding communities were nominated to the State Environmental Quality Board's Citizen Advisory Task Force. The EQB task force was supported by the state agency and produced a fairly extensive record as well as the report which was submitted to the agency in February, 1996. The Task Force thoroughly reviewed the history of nuclear waste management in Minnesota and read all the pertinant documents. Task Force members used their report to lobby the legislature and agencies, urging them to effectively address the public concerns raised in their process. While the Task Force recommendation was not to establish an alternate site, it took responsibility for the implications for Prairie Island and made several recommendations.

2) FLORENCE TOWNSHIP, as local government body, held informational meetings, formed township task forces, and raised and appropriated monies for legal defense. We used the crisis, after an initial round of divisiveness, to educate and sensitize the larger community to the concerns of our neighbors in Red Wing and Prarie Island. We used our experience to encourage greater accountability to public concerns in nuclear waste permitting and policymaking processes. We suddenly found that our 'back yards' extended far beyond the borders of the township: from Prairie Island, downriver and now-to Utah. While the site may change, the fundamental issues have not.

3)

C.U.R.E. (Communities United for Responsible Energy). Citizens from Florence Township and Lake City formed a citizen's group. The focus of our mission statement was, and still is, the responsible (long/short term) management of nuclear waste. C.U.R.E has been active in the legislative process over the last 3 years and has submitted numerous public comments: to the EQB, the PUC, the DPS and the NRC. C.U.R.E has worked with a wide range of other public interest groups to promote responsible nuclear waste management at the state and federal levels. C.U.R.E. is a member organiza-tion of Prairie Island Coalition. 49

Secretary DOCKET NUMBER PETITION RULE PAM 72-/ DOCKETED USNRC ( 4> 3 FR. I Z. O 'f tJ) 5711 Summerset Dr. Midland, MI 48640 '98 MAY 29 p 3 :2QMay 25, 1998 Nuclear Regulatory Commission Washington, D. C. 20555 OFFICL=.,..ir: _<~F_ *,:: , ~ { RULt.:.,.',,,I , l() ! ',. ~ ) Re: Docket'W~~~ ~12-~S SIAFF To the Secretary, NRC: The following are my comments on the petition for rulemaking filed by the Prairie Island Coalition and assigned to Docket No. PRM-72-4: In evaluating the impact of thermal shock during any unloading process of dry cask storage, the types of changes that the fuel has undergone during the dry storage process has not been adequately addressed. These changes have been described in NRC's consultants' report, CNWRA 93-006, "Characteristics of Spent Nuclear Fuel and Cladding Relevant to High Level Waste Source Term", prepared by the Center for Nuclear Waste Regulatory Analyses, San Antonio, Texas, May, 1993. That report states that "the dry environment has the potential of producing such problems as further fuel cladding oxidation, increased cladding stresses and creep deformation as a result of rod internal pressure, and volume expansion of the fuel due to air permeating through any pinholes and incipient cracks in the cladding. These possible spent fuel and cladding alteration modes could be quite accelerated under dry storage conditions, since the temperatures are much higher than in wet storage. Temperatures in the range of 300 to 400 degrees C. are being considered for the extended dry storage of spent fuel for periods up to 100 years following discharge from the reactor." (Exhibit 1) Since no cask that has been in dry storage in a helium environment for any length of time has ever been unloaded, the public can have no confidence in the claims made for unloading procedures, given the serious mistakes and surprises that have been a part of dry cask storage operation thus far. This can only be resolved by unloading a cask through the spent fuel pool process-- such as Consumers Power Co. promised when a defect in a weld was discovered in Cask #4 in Aug., '94. But at that point, the utility realized it did not have an adequate unloading process even for the newly loaded undamaged fuel in Cask #4. Having an adequate unloading process was required as a condition of loading a cask in the first place. Although it is claimed that there is now an adequate unloading process, (NRC Report 50-255/96201; 50-7 /96201) that report does not address the types of changes in dry storage described in NRC's consultants' report CNWRA 93-0006. This report JUN - 4 1998 Acknowledged by card........ _ _,.,."".,.,.,..,.,.,..,

f *,.' r,. ',,, e

states that "the unloading procedures provides direction for coping with damaged fuel", but it does not say how such fuel will be managed after unloading since damaged fuel cannot be placed in another cask. This is a serious omission in the consideration of the effects of thermal shock. Concern for crevice corrosion in what is now the VSC-24 cask design has not been addressed or resolved. A letter by F. Sturz of the NRC to J. Massey of Pacific Sierra Nuclear (now Sierra Nuclear Corp.), Aug. 28, 1990, states that crevice corrosion could lead to perforation of the MSB. It further states that the ASM Handbook, Vol. 13 on corrosion identifies this design as one that must be avoided since it permits corrosion cells to be established along the long, flat surfaces. "Air, water vapor and iron are all that is needed to initiate crevice corrosion. The use of coatings could exacerbate the problem since any defects would establish localized corrosion cells." This problem may apply to other cask designs also. (Exhibit 2) Following the explosion in the VSC-24 cask being loaded at Point Beach in May, 1996, Don't Waste Michigan, a non-profit citizens coalition in Michigan, retained a noted corrosion engineer expert, Dr. Rudolf Hausler, to review the documentation that the NRC required of all utilities with dry casks following that event in its Bulletin 96-04. (Dr. Hausler had been previously retained by the E.P R.I. to identify a corrosion problem that was plaguing all U. S. reactors. He not only identified the problem, but designed a corrosion inhibitor for it.) It was determined that the explosion at Point Beach was caused by a corrosion reaction between the zinc coating and the boric acid of the spent fuel water during the loading procedure. Dr. Hausler provided a report with new insight into the hydrogen evolution in that cask. (Exhibit 3) We incorporated his report into a 2.206 petition to the NRC. Dr. Paperiello responded to the petition by dismissing the significance of Dr. Hausler's observations. Dr. Hausler responded with some important observations on the potential corrosion problems in this cask. He also noted that the comments by the NRC staff on previous experiments with zinc "reveal a stunning ignorance on the part of the NRC staff of the difference between intermetallic chemistry and the chemistry of ionic solutions in contact with a metal." (Exhibit 4) I appreciate the opportunity to comment on this rulemaking procedure. Yours sincerely, lair, PhD on' t Waste Michigan

/ t *. * -~!.>** Preparod tor Nuclear Regulatory Commission Contract NRC-02--88-005 I _-C-(~nicr for Nuclear Waste Begulatory* Analys~s San Antonio 1 Texas rJ.1()~,;';i[1;*,1 *, ,;,:1(1'.<1 l Plif.? w~-.:-, r i-wM

• I l

, [1/<

. CNWRA 93*006 I I I . /

-",\\.\\_ J~':4t-vel~ U> give die activity as a f\\mction of time after discharge (Locke. 1975). 1be aJQJJatiom~~;1}~:,~~l. '\\ *uic *code have been validaied api:osr experimental data on ~pent fuel Bdionuclide iDvemories (Red> ei'**:: *:."*......

• al.* 1990).

4.3 CHARACTERISTICS OF SPENT FUEL lmdiated tud still contains most of us orilinal 2310, about one-cbird of its original 235U., * .umost all.the fission product&. ~ tbt tra:murwC (DU) isotopes, and mmy aaivadon prod:o.ets. Tables 4-7 *and 4-& sw:nmarize the most important isotopic feazura of typical L WR. spent fuels.. The volume of ~laSteS generated during the service life of LWRs is given in Table +-9. 4.4 STORAGE OF SPENT L\\VR FUEL Prior to permanczit disposal in a repositoty. the Spent we1* ii upecud to be stored for some years (perhaps 10 years or,nore) to allow the assemblies to cool down as a result of dissipation of the

• *. decay ht.at. Two principle mdhods are available for atoring the 1pent fuel for extended periods of time?

namely. wet storage in water pool$ and dey storage in air or gases (EPRI. 1986; &ilcy et al., 1986; EPRI, 1984; Johnson, 1979). Wau:r pool storq;e has been used in the U.S. since the fint reac:tOtS were i,~,built in 1943 (Johnson. 19n). HowJYet, experience with dry storaae, an atccmauve to exU8ded wet

f. storage,, is rather limited. Toe Nd may be stored WlCODIOlidated '1t comolidatod. By defulition, unconsolid2ted fuel means intact assemblies; consolidated means tbcfuel rods are ranoved from the fuel assembly and placed in a grid with closer spacing than that of an Intact assembly. or the rods are plac<<l in a close-p&eked may inside a eanistet (Zach&, 1988; Jobmon, 1986;.EPRI, 1989). Volume AVlngs of 2: l by consolidation have been demonstrated by many utilitiea (Zachy. 1988). The ad\\tantages of fuel rod consolidation are obvious: (1) storage capacity would bo almost doubled, and (ll) tho number of spem-

. fuel shipping casks can be halved. In the wet storage mode, the decay heat from Che spem fuel is removed by deionized water (DIW) at a temperature bcJow 40°C. Zircaloy claddin1 does not experience any siJQlficant additional corrosion or hydriding under such coDditlom compared to that experienced while in c.ore. In the dry 4 "5* * * * *,,_ storage mode, the decay beat is removed by usin, air or an inert ps (usually btlium or nltro&en) under forced ~nvcctiou. The dry storage facility is simplc-z and cheaper to maintain compared to wet storage. f , l { On the other hand, the dry environment has the potential of produclna such problems as further fuel cladding oxidation. increased cladding mcsaes and c:reep deformatioo u a rcault of rod inWna1 pressure. and volume expansion of the fuel due to air ~ 1hrouJh any pinholes and incipient rolCb in the ~'

• ' *1,* cladding. These possible spent fuel and c:laddin& alteration mode, could bo quite accclented undet dty
• storage conditions. since Ult tetriperanu-es are much blgber tbm in wet 5tarqc. Tempa:aiwes in the nuigc l/ of 300 to 400ac are being considered for the ntcadcd dry storage of spent fuel for periods up to
• 100 years following discharje from the reaaor. Bccausc of me swcd reasons, any extended dry storqe

(( may require evaluation of any additional pro<<$S of materials degradation or alteration of the cladding propenies which may influence iu subsequens. behavior in a repository. Long-cam bc:bavior of n.tco.ded (r wet stored spent fuel may llced to be modeled differently from.~lies that expericnu dry storage following discharge from the rQCt.Or. Allowable long-term storage temperatures and times for dry SU>rage of spent LWR fuels will depend upon a number of factors including claddin& stress levels. fuel type and assembly design. materials condition of the cladding at the time of dis.charge. ~ecay beat history of the spent fud, and hear 4-ll

COPt'\\ENTS ON PSN 8/14/90 LETrER The change to Section&.*. page 8-8. which deletes the parenthetical rtfftak, M(only once per wee~ security patrols),* frOffl the discussion of Yisual surveill1nce requirements is acceptable.

• The chinge to Section 9.2, page 9-2, which ch1nges the reconrnended maintenance schedule for inspection of a'r vents for detection of blockage from Rweekly* to Rdaily or weeklyw is not acctptab1e. The original weekly frequency is also unacceptable.

This 1s bec1use the report indicates th1t un1ccept1ble concrete temperatures (over t~e 350° shori.. tenn temperatul"e permitted by ACI 3C9) MIY o~cur in l~ss than a week if the vents are blocked. The suggestion thlt the ut111ty conduct 1nd subftlit results of I strength ttst cf the proposed eonerete follo~1ng high temper&ture exposure (at the revised section 12.2.3,l) i$ not an adequ,te basis for the extended surve111uc, frequency (further discussed belo.,). -.* The change to Section 11.2.9.2 th1t adds *or weekly* to tne desc;rtpt1on of tht iurveillanct wh,ch would revt11 possible vent blockage by* d1il1 patrol 1s un1ccephble. The th1nge to the first paragraph of Sectton 12,2.3,1 that follows *the interval for this surve111ance is to be 1 day (24 hours)* is un1cctpt1ble. The piragr1ph should end lt that point. This 1s because no evidence is provided that th! suggested testing wn 1 nt isfy the cc,ncerns for both the negat he dur&b111ty 1nd negative strength effects of elev1t(d temper1tures on roncrett. The staff cons1d~rs th1t NRC 1ccept1nce of departu~e fr°"' the Atl 349 code limits on high temperature exposure of concrete would r,!QU 4re convincing empirical evid~nce to counttr the curl"ent ACI position {ACJ 3-i9), ind to va1tdlte tests which C<.,uld prove the high temper1ture c1p1b111ty of the individual concrete mix. Neither has bten submitted. crRROSION CONCERNS IN TH£ PSH DESIGN This was I topic of dhcunion at 11*rev1ous meeting 1n Rockville. Although PSN has provided idequate discussion of the overall Nrg1n af the MSB not to f&fl by gener1l torrosion, t~ tddftional co1roston conterns are still outstanding. The NRC staff's concerned ttiat PSN hu no-:. tnod1f1!d 1ts design to avotd metal*to~ mthl cor1tact betwt,n \\_he MS8 ind the VCC liner. Thh corrosio.* during the 11feti~ of the VSC whi-di could ~ake 1t difficult, if not 1,npois1b1t, to r~ve the HSB fr~ the YCC. The N~_~tlilff hu not been abh to quant H)' the corrosion

• bond 11 strength. _ Hovever, lt -i value as low u 200 psi h usul'l'ltd, tt.)ie1ds 1 load 600 1000 pounds of vert1cal force to pull the two surfices apart. This far exteeds the c1p1c1ty of the lifting lug 1-.

Another concern 11*for crevice corro,ion. Crevice torroiion could 1ead to perfont1o" of the Nse*~~- A911n, the NRC stiff hu not been ab1e to quantify tht tffect. However, the ASM Handbook, Volunt 13 on C9rrouon. 1dentif1u the design .t,s,gtf-thlJ must 6i_avo1dtd. fhe curnn\\ duign h v@ry Susceptible since 1t pem s corros1on ct11l to 6e est1b11$h,~ 1\\ong the 1ong. flit surfaces, Air, water vapor, 1nd tron art 111 that is ne,~M to tn1t11tt crevice corrosion. The~ use of c01tit'tgs *could uaeerbatt the prt1ble111 s1nct any defects would establish localiud corrosion cells. In 1ddttto1,, the onset of gtnen1 corrosion wi11 generate water-1bsorbtn9 oxides that could 1cc,ler1.te crevice c;orrosion. Cr-eviu

Dr. Mary Sinclair 5711 Summerset Dr. Midland, MI 48640 Rudolf H. Hausler 7804 Pencross Lane, Dallas, TX 75248 Tel: 214 490 8605 Fax: 214 490 8878 October 6, 1996

Subject:

Preliminary Review of Documentation relating to the Safety Evaluation of the Use of the Sierra Nuclear Corporation's Dry Fuel Storage System I. Scope This analysis is based on the documentation available at the time of the analysis (see . references attached as Appendix II). The primary focus is corrosion and materials peeformance both during the loading process of the dry storage sy~tem components and during long term storage. It is assumed that the reader is familiar with the components of this system (Ventilated Storage Cask System, VCS) and its components (Multi-Assembly Sealed Basket, MSB; Multi-Assembly Transfer Cask, MTC, and the Ventilated Storage Cask, VSC). Therefore a detailed description of these components is omitted here. II. Background The issues relating to corrosion and materials performance of the ISFSI (Independent Spent Fuel Storage Installation System) arose when an explosion occurred during the welding of the lid onto the MSB. Subsequent to this incident, the NRC issued Bulletin 96-04 requesting information on the "Chemical, Galvanic, or other Reactions*in Spent Fuel Storage and Transportation Casks" dated July 5 1996. The responses to this request, as well as the responses to subsequent Confirmatory Action Letters (CAL's) by the NRC are reviewed. The immediate concerns relate to the nature of the explosive gas which apparently was generated inside the MSB, possible means to avoid such incidences in the future, and questions relating to the behavior of materials associated with the MSB and the VSC on long term exposure to high temperatures. III. The Evolution of Hydrogen. The MSB, consisting of a cylindrical storage can containing 24 rectangular sleeves (designed to hold one fuel rod assembly each) is fabricated from heavy steel sheet. The outside of the MSB is coated with a zinc primer covered by an epoxy coating. All inside steel surfaces are coated only with a zinc primer. This includes the inside wall of the MSB as well as all surfaces of the sleeves. As a consequence, the Zircaloy fuel rods are in electrical contact with primer coated surfaces and form a galvanic corrosion element. 10/9/96 rhh/sinclair/final

III. 1 Boric Acid Chemistry The spent fuel pool (SFP)water contains 2850 ppm of boron or 16,310 ppm of boric acid. (There is no indication in the available literature that the solution might be buffered with sodium borate). The pH of this solution can be calculated as 4.86 (see Appendix I). Contrary to a comment by Wisconsin Electric Power Company (WEPCO) the boric acid in pure water does not increase the pH but reduces it from the pH of water; therefore the corrosion of zinc is.accelerated, not reduced (see ref. 11, pg. 2, last par.) It appears however, that the low pH of the borated SFP water has been recognized a.nd is quoted as being about 4.5 in several other documents. III. 2 Corrosion of Zinc in Borated *water The corrosion rate of zinc primer applied to steel coupons appears to have been measured by NWT and _by Entergy (Ref. 7) These tests seem to have focused on the rate of hydrogen generation and precipitate formation. No corrosion rate data are available, nor have the tests been described in any detail in the available documents. It is being said that the results from NWT Corporation (San Jose, CA) "appear fo be. fairly consistent with those from Entergy Company", but about 8 times higher than the rate indicated by WEPCO. WEPCO assumed a zinc corrosion rate of 0.028 inches per . year (ipy) based on literature data quoted in Uhlig's Handbook of Corrosion. ,:-:H;owe.ver, no effort was made by WEPCO to translate that into a hydrogen evolution rate. Rather, WEPCO chose to assume a hydrogen evolution rate based on some measurements (Ref. 9) These measurement indicate that the solution was saturated with hydrogen at a partial pressure of one atmosphere. The reported concentration of 15 cc H/kg HiO corr~nds to saturation of the water with hydrogen gas at of 1 ata and a temperature of 25 C. One does not know from the report, where the water was sampled and how the analysis was made. However, if the water in the MSB became saturated with hydrogen at 1 ata, the hydrogen evolution was no doubt quite fast and the atmosphere above the water must have been very rich in hydrogen. None of these factoids are useful for the assessment of the zinc corrosion rate and the associated H2-evolution rate under use conditions. There are several effects which must be taken into consideration in the design and evaluation of corrosion tests, or when comparing literature results.

a. The data quoted by WEPCO (from Uhlig) were obtained on solid zinc samples and hence apportioned to a measurable surface area. The zinc primer consists of zinc powder in an inorganic matrix. Depending on the particle size and the zinc content in the primer, a surface area many times the apparent area may be exposed to the corrosive medium. Because of the small size of the particles, corrosion is very likely accelerated over that measured on solid coupons. Furthermore, the hydrogen evolution rate would be proportionately larger at the ratio of the apparent, or geometric, surface area to the real surface area of the zinc particles. Literature data, such as those quoted by Uhlig, have no relevance with respect to zinc particle corrosion in paint primers.

b. Galvanic Corrosion: The Zircaloy surfaces of the fuel elements inside the sleeves of the MSB are in electrical contact with the zinc primer, with which the inside surfaces of the sleeves are coated. It has been amply demonstrated, e.g., that the Inconel tubes in a steam generator galvanically accelerate the corrosion of the carbon steel walls of the generator during chemical cleaning (Ref. 12). (The effect on the welds is even larger).

There is no reason to assume that Zircaloy will not similarly accelerate the corrosion of 10/9/96 2 rhh/sinclair/final *

---

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zinc. As a consequence, this fact must be considered in tests which purport to evaluate the behavior of the zinc primer when in contact with the SFP water during loading of the MSB and subsequent operations. The NRC has repeatedly called for comments regarding galvanic reactions. This question has not been answered to date. WEPCO indica,tes (ref. 9) that the "rate of galvanic corrosion of the zinc from the Carbo Zinc 11 SG may be three times higher than in neutral (pH 7) water". Oearly this comment does not answer the question since the galvanically accelerated corrosion rate at the low pH is not specified. The same reference, pg. 2, indicates that "Published data were found for the kinetics of zinc corrosion in boric acid". However, no reference is given and no data are quoted.

c. Corrosion Kinetics: All metals which can form a solid corrosion product when corroding in an aqueous medium exhibit passivation behavior. This means that the corrosion rate observed on the bare metal upon immersion into the corrosive fluid gradually decreases with time. The initial corrosion rate, immediately upon immersion, may be orders of magnitude higher than the average rate for the duration of the test. The extent and the rat~ to which this decrease of the rate occurs depends on the n~ture of the corrosion product. Therefore, an average corrosion rate result (such as weight loss) which has been obtained in ~ long term test is not representative of the possible short term corrosion rate and is a function of the test duration.

• ,*Furthennore, a test such as the one performed by ANO (Ref. 10) may be misleading in other ways. The test protocol called for increasing the temperature from 120 °F, following immersion of a primer coated coupon into the borated water, to 200 °F, followed again by a slow cool down to the starting temperature. The duration of the test was 75 hrs. It is known that increasing temperature promotes passivation. The corrosion rate in borated water may actually decrease with increasing temperature. (The phenomenon is well documented for CO2 corrosion). Hence the temperature cycle performed in ANO tests may not be representative of the worst case condition. In view of the importance, physically and financially, of the question of the rate of hydrogen generation and associated SFP water contamination, it would seem that a full evaluation of the zinc corrosion under all possible conditions of actual use ilJ. the application under consideration should be conducted.

d. The Nature of the Zinc Primer.

Ref. 8 (WEPC0) discusses the use of zinc primer from the point of view of contamination introduced into the SFP water. Here it is stated that the primer contains "25,000 ppm of zinc and 530 ppm of lead". All other specifications for the zinc level in the primer indicate 85 % (850,000 ppm). If the lead contamination of the zinc in the primer were indeed 2 %, as indicated by the WEPCO numbers, then the the amount of lead introduced into the SFP water through corrosion of the zinc would be 34 times larger than what WEPCO assumed it to be. This discrepancy, it would seem, calls for a re-evaluation of the lead impurities which could be introduced into the SFP water, in light of the limiting amount of lead which can be tolerated according to the Westinghouse specifications (ibid). III. 3 Conclusions It would appear from the above that the question of corrosion of the zinc in the primer has not been resolved quantitatively, neither from.the point of view of hydrogen evolution, nor from the point of view of SFP water contamination. 10/9/96 3 rhh/sinclair/final

i The NRC request for review of the materials performance questions by a corrosion specialist does not seem to have been fulfilled. IV. Additional *concerns Relating to Materials Performance IV. 1. Temperature and Epoxy Coating Specifications for the Epoxy coating to be applied to the outside surface of the MSB are presented in Ref. 11. It is noted that no temperature performance limits are specified. In view of the fact that the temperature in the MSB on long term storage may be as high as 800 to 900 °F (in the extreme as much as 1000 °F) the performance of the Epoxy top coat becomes a concern. It is recognized that the space between the MSB and the VCC is air cooled. This suggests that the lowest overall heat transfer coefficient is given by the natural convection of the air through the annular space. As a consequence, the temperature of the epoxy would be expected to be very little below the values quoted. Degradation by carbonization of the ~ganic material can, therefore, ~e expected. No surface tem~ture has been calculated for the outside of the MSB. Since the unloading of the VCS-is part of the safety evaluation, and since it would also have to be performed in the SFP, the corrosion of the residual zinc on the outside of the MSB in contact with charred epoxy paint becomes a real concern. While extensive simulations and calculations of the heat transfer rates from the MSB to the environment have been made it should be recognized that the heat transfer coefficients for radiation, conduction, and natural convection are afflicted with uncertainties. It is therefore suggested that confidence limits be given for all such calculations and be reviewed by a third party. Oearly the heat transfer cannot be very efficient if the temperatures in the MSB can rise to 1000 °F. Common uncertainties in the coefficients of 20 to 50% (particularly with respect to thermal conductivity through coatings) can alter the calculations significantly. IV. 2. Zinc-Steel Interactions The melting point of zinc is 786 °F, or well below the anticipated temperatures inside the MSB. This raises a serious concern about liquid metal embrittlement. While such phenomena are not generally observed following the galvanizing process of steel, it must be recognized also that during galvanizing, steel is in contact with molten zinc for a very short time only. The exposure of steel at 800 * °F to molten zinc over a period of years is totally uncharted territory and needs to be seriously investigated. IV. 3. Zinc-Zircaloy Interactions The continued integrity of the Zircaloy cladding over the 20 year licensing period is imperative if the fuel rods are to be unloaded any time in the future. In this context at least two issues need to be reviewed relating to stress corrosion cracking and liquid metal embrittlement. It has been reported (Ref. 14) that concentrated methanol and organic solvents containing small amounts of chloride can cause stress corrosion of zirconium. Generic specification of cleaning solvents should therefore not be permissible, since many such solvents either contain methanol, are chlorinated hydrocarbons, or contain chlorinated hydrocarbons. 10/9/96 4 rhh/sinclair/final

Similarly, liquid metal embrittlement has been reported for zirconium in contact with molten cesium or when exposed to liquid sodium or cadmium. The effect Qf zinc metal in contact with Zircaloy at 800 to 1000 °F must therefore also be ~xamined since zinc, like cadmium and mercury is also a 2b' transition metal. Liquid embrittlement of metals by mercury is well known. Most Zircaloys behave very similar to zirconium because the alloying constituents are present in the Zircaloy in small amounts only. N. 4. Performance of Steels at Low Temperatures

a. Manganese Steels: The generic specification of structural steels such as A-36 and A-516 is unsatisfactory from the point of view of cold temperature behavior. The embrittlement temperature is very dependent on the manganese content in the steel as well as on other alloying elements which are not included in the generic compositional specifications of these steels (Ref. 13). A-516, e.g., has a Mn spec of O.S-to 1.2 %. In view of the fact that the NRC has specified a temperature 50 °F above the embrittlement temperature, efforts must be made to d~velop more precise specifications for the steels to be used µi Q.rder to assure safe handling of the storage cask under all possible conditions including emergencies.

b. Welds: It is noted that no specifications relative to the embrittlement temperatures have b~h_ de_fined for welds. The absence of such specifications is considered very dangerous.

IV. 5. Residual Corrosion of the Zinc in the MSB and Resulting Pressure on Long Term Storage It has been proposed that following closure of the MSB (welding of the seal lid) the water is drained from the inside of the basket. Complete drying is then attempted by a vacuum process. There appears to be no verification of the residual water which may remain in the basket. Such verification is essential in order to determine the pressures which might develop in the basket, particularly if these pressures need to be used in the evaluation of the -structural integrity of the basket on long term storage. V. Summary Some safety aspects of the Sierra Nuclear Fuel Storage System have been reviewed in the wake of Point Beach (Wisconsin Electric Power Company) hydrogen explosion. Both short term and long term concerns are discussed.

• The storage configuration allows for the fuel elements to be in contact with zinc (metal) primer which is applied to all internal steel surfaces of the MSB. This creates a galvanic element between the zinc and the Zircaloy which in tum accelerates the corrosion of the zinc. No tests have been performed simulating this galvanic situation. Calculations relating to the hydrogen evolution rate and the SFP water contamination rate, which are based on the available tests, have therefore little relevance.

Numerous discrepancies in the responses submitted by the various companies to the NRC request for additional assessment of safety concerns reveal an alarming lack of consensus.

• The epoxy coating applied to the outside of the MSB cannot possibly withstand the temperatures said to develop on long term storage. This will result in an expensive SPF water contamination during unloading of the radioactive material.

10/9/96 5 rhh/sinclair/final

The NRC has published temperature specifications relating to low temperature embrittlement of the structural carbon steels used in the Sierra Nuclear spent fuel storage system. These specs have not been translated properly to the corresponding material composition specifications (for the steels in question), thereby creating additional hazards in handling the storage casks under possible emergency situations. The zinc-steel interaction at 800 to 1000 °F and possible steel embrittlement over a period of 20 years has not been considered, again creating a n additional hazard for handling the MSB in the future. Similarly, the effect ( or the absence thereof) of molten zinc on Zircaloy has not been verified experimentally. The possible failure of the cladding could make unloading of the sp~t fuel rods impossible. The vacuum drying process does not seem to have been experimentally verified relative to residual water remaining in the MSB. All calculated data relating to pressure in the MSB during storage, and continued integrity of the seal welds are therefore open to questioning. None of the temperature calculations and heat transfer assessments have been experimentally verified. There has never been a field test of this storage sy-stem. It is not apparent that the storage system has been instrumented in order to verify the design assumptions.

• o onsulta f

R.H. Ha 10/9/96 6 rhh/sinclair/final

APPENDIX I Determination of pH of Boric acid solution: Solution: Therefore: since: 10/9/96 Distilled water @ room temperature Boron content 2850 ppm; 16,310 ppm boric acid; 0.264 mol/L boric acid (KY (boric acid)= 7.3x10*10 [H+] = [H2B03] [H+ J2 = 7.3

• 10-10
• 0.264 pH =4.86 7

rhh/sinclair/final

APPENDIX II.

References:

1. Safety Analysis Report (SAR), October 1989; Section 3: Principal Cask Design Criteria

2.

Safety Evaluation Report; NRC March29~ 1991 3

• Safety Evaluation Report; NRC April 1993

4.

Notice of Non~Conformance; NRC, June 7 1996

5.

Initial Responses to CAL 4-96-002; Entergy Operations, June 2, 1996 6. Request for additional Information regarding Bulletin 96-04, "Chemical, Galvanic, or other Reactions in Spent Fuel Storage Casks", Letter from NRC to WEPCO 9/4/96 \\

7.

SNC response to NRC Bulletin 96-04 Jdr VSC-24 Dry fuel Storage System; Sierra.. Nuclear Corp., August 16, 1996 .>; c, 8,.,

• Use of Carbozinc Paint as a primer for the Multi-assembly Storage Basket, Internal

-

• Memo, Wisconsin Electric Power, July 18 1995

9.

WEPCO response to NRC Bulletin 96-04 for VSC-24 Dry fuel Storage System,: WEPCO August 16, 1996

10. Consumers Power response to NRC Bulletin 96-04 for VSC-24 Dry fuel Storage System:* Galvanic, or other Rea.ctions in Spent Fuel Storage and Transportation Cask;

• August 19, 1996

11. WEPCO response to NRC request for additional information relating to: Bulletin 96.:.

• 04 fQr VSC-24 Dry fuel Storage System: Galvanic, or other Rea,ctions in Spent Fuel Storage and Transportation Cask; September 6 1996

12. Non Proprietary Corrosion Inhibitors for Solvents to Qea,n Stea,m Generators; R.H.

Hausler, EPRI report NP 3030, 1983.

13. Metals Handbook; 9th Edition, Vol. 1, Properties and Selection of Irons and Steels, American Society for Metals, 1978, pg. 689

14. Metals Handbook; Desk Edition; American Society for Metals, 1985, pg. 20-35 10/9/96 8

rhh/sinclair/final

CORRO-COr\\JSULT A, RH HAUSLER Dr. Mary Sinclair, Ph.D. Co-Chair Don't Waste Michigan 5711 Summerset Drive Midland MI 4864 via fax Dear Dr. Sinclair 'i2' (214) 490 8878 ~i12/29/96 Rudolf H. Hausler 7804 Pencrol.\\S Lane, Daile:.~, TX 75248 Tel: 2:i.4 490. 8605 Fax: 214 490 8878 December 29, 1996

Subject:

NRC Response to Don~t VVaste Michigan Petition I have reviewed at your request NRC Director Paperiello 's response (December 10, 1996) to your Petition of October 18, 1996 and my.letter of October 6, 1996. The following are my comments in part. It is difficult, as I am sure you are aware or: to formulate ft consistent assessment relative to the inherent safety of the VSC-24 nuclear spent fuel storage system in view of the fact that only fragmentary information emanates from the NRC relating to the design of the device, the Utilities' responses to the various CAL's (\\.~onfirmatory action letters) and the NRC's subsequent evaluation of the responser.. Neve-rtheless, from what I understand and have learned from the NRC's review of EOI's (Entergy Operations, Inc., ANO units I and 2) response to the NRC Bullet111 96--04, EOI did not specifically address the question of the accelerated hydrogen evolution Ct'.used by the galvanic element fon11ed by the zinc con.ting and the fuel assembly. However, Director Papcriello addressed himself to this question in a manner which requires some comments. As we understand it, a fuel assembly consists of the foel encased in sealed zircaloy tubing (cladding). Director Paperiello states that the fuel assemblies are furthermore fitted with stainless steel tmdcaps and that these endcaps are in contact witfr the zinc primer (pg 2. par. i). He then invokes contact resistances between the fuel assembly and the stainless steel endcaps, and the stainless steel endcaps and the primer, to rule out ~u1y effectiveness of the galvanic couple. I wollld like to point out the following: Cl215

..-----~~~-------~.-c~-7 -q

CORRO-CONSULTA, RH HAUSLER "Gil' (214) 490 8'-178 !if:12:/29/96 (~'10:-37PM

a. If the endcaps are fitted sufficiently loosely over the ends of the fuel assemblies such that electrical contact resistances exist which are sut1icientlv high to negate the flow of a low voltage galvanic current, then cre*vices ~nt~hf~xist. wh;re borated water can be held by capiHary action. Such a design could lead to crevice corrosion similar to the well known steam generator phenomenon which causes denting.

Much more lik,~ly, howeve.r, is a di;isign vvhereby the endcaps are fitted tightly over the ends of the. foe.I assemblies. In this latter case no contact resistances can possibly exist unless ru1 electrically isolating material was fitted between the end of the fuel assembly and the stainless steel endcap.

b. The fuel assembly (st.ainless steel endcap) has to be necessarily in contact with the zinc coating for a galvanic effect to cause, acceleration of hydrogen evolution.

For this to occur a complete electrical circuit has to be established as fol]O\\vs: The potential differences bet,veen the hydrogen evolution reaction (approximately- 0.4 V) and the zinc oxidation reaction ( approximately -0. 9 V) causes an dedrical cum~nt to flow from the zinc coated carbon steel to the fud assembly. The-current then exits the fuel assembly as an ionic cunent ( electrolytic current) travels through the electrolyte layer and reenters tbe zinc coating. The entry reaction causes the zinc to oxidize, while the current leaving the fuel results in hydrogen evolution. The hydrogen evolution from the foeJ assembly adds to the hydrogen evolution which will occur by the reaction of the zinc with the water even when not in galvanic contact. Thus the phe-nomenon is called galvanic acxeleration. (The mechanism of galvanic acceleration and the parameters which control it have been studied in depth, for instance, during EPRI 's extensive Skam Generator Chemical Cleaning projc,.::t.) Once a galvanic element has been established the thermodynamic potential difference between the zinc ( carbori steel) and the zircaloy (approximately 0.5 V) decays to near zero and equals the sum of the various IR (current times resistance) drops encountered in electrical circuit. These IR drops are defined as foll.ows:

1. eledrolyte resistance (]Rel.ct,)

2. cathodic polarization re-sistance (IRp,,L.:ath., non linear)

3. anodic polarization resistance (IRpoI.a,d.' non line,ar)

4. electrical resisLmce in the metal (IRm,:.a1)

5. contact resistances (IR"""ct)

The sum of the.se 5 voltage drops must equal approximately 0.5 V. In order to judge the effect ofthe contact resistance one would have to able to assess its value [J3/5

C:)~RO-CONSULTA R:-1 Hft.USLER 'Ir (214) 49Cl 8876 lii:12/29,96 relative to the other resistances. Only if the contact resistance is substantially larger than any of the oth.=,rs, or in fact the sum of the. others, can a galvanic effect be ruled out. A difficulty in this analysis is the fact that the polarization resistances r.re a function of the current. They are in fact by definition zero if no current flows. It, therefore, stretdu;-:s the imagination to rule out a galvanic effect simply on the assertion of contact resistances. Considering that the full \\\\*e:ight of the fuel assembly rests on the bottom of the sleeve within :he basket i..11.to which it has been plact~d, would, in my estirnation, result in a very good and intimate contact with the zinc on the bottom of the basket, and insure a very low contact resistance. It is not so much the fa,t that galvanic acc:elt:.ration 0f hydrogen evolution may occur, but ratherthe uncertainty of the-magnitude of this galvanic effect which requires a more thorough analysis of this aspect of the VSC-24 in order to prevent or rule out any future surprises. We,,..1ill need to review all the other points raised and objected to by Director Paperiello as hofefully the pertinent literature becomes available. It is, however, noted that Sierra Nuclear Corporation is actively investigating an alternate coating and that the NRC staff "bdieves it prudent to use an alternate coating'*. Current pr~jections are for an alternate coating to be available in six months. We would also like to point out that the question of the interaction of zinc with zircaloy at high temperature~ has not been satisfa,:torily been laid to rest. The comments by the NRC staff that previous experiments, in which "zinc was injected into boiling water re.actor systt~ms without a detrimental effect on the cladding of the-foel assemblies, re*veal a stunning ignorance on part of the NRC staff ofthe. difference betv-.*een intermetallic chemistry and the chemistry of ionic solutions in contact with~ metal. In view of the above it v-.*ould appear thl'lt a temporary halt to spent fuel loading into the VSC-24's pending additional reviews is definitely indicated. Please call me if you have any additional questions Sincerely Rudolf H. Hamler 2;415

DOCKET NUMBER PETITION RULE PRM ?~,::._.If (b3 F,q. 120To) OVERLAND LAW OFFICE Carol A. overland Attorney at Law 402 Washington st. so. Northfield, Minnesota 55057 DOCKETED USNRC (507) 664-0252 (507) May 18, 1998 Secretary Attn: Rulemakings Staff u.s. Nuclear Regulatory Commission Washington, DC 20555 On August 26, 1997, the Prairie Island Coalition filed a §2.206 Petition, Docket No. PRM-72-4, portions of which have been deemed a petition for rulemaking under 10 u.s.c. §2.802. The issues in this petition subject to rulemaking are those of retrievability, degradation, and unloading of fuel stored in casks; issues universal to all dry cask storage. These and similar issues have been raised regarding the TN-40, VSC-24, and NUHOMS casks, and rulemaking is best suited to develop a solution to the cask problems and to assure that these safety issues are addressed. It is the mandate of the NRC to protect the health of the public and to minimize danger to life and property, and the mandate authorizes use of rulemaking to accomplish these goals. For these reasons, I submit this cover letter and the attached comment to the Nuclear Regulatory Commission, and ask that the Commission initiate formal rulemaking on the issues of retrievability, degradation and unloading of casks used for storage of nuclear waste. If you have any questions, or require further information, please let me know. Very truly yours, ~~ carol A. overland Attorney at Law CAO:dc . JUN = 4 1998 Acknowledged by 001@,o**-*000000~

BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION IN THE MATTER OF: ) ) ) ) ) ) ) ) ) ) ) ) ) PRAIRIE ISLAND COALITION, Petitioner,

v.

COMMENT OF CAROL A. OVERLAND UNITED STATES NUCLEAR REGULATORY COMMISSION, Respondent. Docket No.: PRM-72-4 The Nuclear Regulatory Commission has vast powers set out in the federal code *. 42 U.S.C. §2201; Cotter Corp. v. Seaborg. 370 F.2d 686 (10 Cir. 1966). Those duties of the Nuclear Regulatory Cqmmission include the authorization to: establish by rule, regulation, or order, such standards and instructions to govern the possession and use of special nuclear material, and byproduct material as the Commission may deem necessary or desirable *** to protect health or to minimize danger to life or property **. 42 u.s.c. §2201(b)(emphasis added). Protection of public health and safety is the responsibility of the Nuclear Regulatory Commission. Id.; see also e.g., Federal Tort Claim of General Public Utilities Corp., et al., 13 NRC 773, CLI-81-10 (1981)(explaining that the NRC does not certify to industry that approved design and procedures will protect industry property, but that its mandate is protection of public health and safety). This rulemaking is necessary and appropriate because the Commission has a specific mandate to: make, promulgate, issue, rescind, and amend such rules and regulations as may be necessary to carry out the purposes of this chapter. 42 u.s.c. §220l(p); see also Rockland County v. U.S. Nuclear Regulatory Com'n., 709 F. 2d 766 (2nd Cir. 1893), certiorari denied 104 S.Ct. 485, 464 U.S. 933 (1983). The NRC itself recognizes the importance of its regulatory function in its Strategic Planning Framework, where it states:

l DEFENSE IN DEPTH ensures that successive measures are incorporated into the design and operating procedures for nuclear installations to compensate for potential failures in protection or safety measures, wherever failures could lead to serious public or national security consequences..* REGULATORY EFFECTIVENESS emphasizes the* approach that, because safety is paramount in the c*ommission' s regulatory program, certain standards and practices to ensure adequate protection will be required, whatever the cost. NRC Strategic Assessment and Rebaselining, Strategic Planning Framework, Appendix II, P. 22. This same documents sets out the NRC "Commitment to protecting public health and safety:" Id.

• Hold the health and safety of the public as our first priority
• Identify and thoroughly assess health, safety, and environmental issues, and ensure that they are resolved in a timely manner
• Work to achieve a reasonable balance between risks and benefits to the public in our regulatory activities The umbrella safety philosophy of the NRC is clear:

ACCOUNTABILITY TO THE PUBLIC dictates that just as licensees are accountable to the NRC, so too is the NRC accountable to the American people and their elected representatives, the Congress. For the NRC, part of accountability entails being candid with the public about what it is doing and why, as well as acknowledging the public's interest in safety issues and its right to know. In addition, the NRC recognizes that the Atomic Energy Act ensures that the public has an important role to play as the agency addresses issues of safety and health. For members of the public to perform that role, they need sound, complete and up-to-date information from NRC. A key element of the NRC's safety philosophy is that nuclear regulation is the public's business. Id. That said, the issues of dry cask storage that have been raised by the Prairie Island Coalition and others are screaming for the attention of the NRC. Many parties have presented issues of whether spent fuel can be retrieved from a dry cask or whether the changes to fuel and assemblies in storage and the unloading process itself render retrievability impossible. They present issues of whether degradation of spent fuel is inherent in dry cask storage, and whether the +evel of degradation present violates the present

I regulations. They also present issues of whether a dry cask used for storage for any length of time can in fact be unloaded in a pool, and whether, absent a true demonstration of cask unloading, further dry cask storage can be authorized. Thus far, the concerns voiced in formal petitions, suits, and interventions before the NRC have been abjectly disregarded. The valid and thoughtful concerns of the public and scientific community can and must be addressed through rulemaking. All §2. 206 petitions regarding dry cask storage or casks generally must be regarded as a petition for rulemaking, and must be added to the balance when the NRC weighs whether the Prairie Island Coalition petition advances to formal rulemaking. There has been such an extensive body.of scientific and legal information and concerns presented that there is no question that rulemaking is necessary for the protection of the health and safety of the public. See § 2. 206 Petition of Prairie Island Indian Community, Docket Nos. 50-282, 50-306, 72-10 and Director's Decision, 46 NRC 35, DD-97-18 (1997)(Transnuclear ~N-40 casks in use at Prairie Island); §2.206 Petition of Don't Waste Michigan and Lake Michigan Federation and Director's Decision, 45 NRC 4 75, DD-97-15 ( 1997) ( Sierra Nuclear VSC-24 Casks in use at Point Beach, Palisades, and Arkansas Nuclear one); §2.206 Petition of Citizen's Utility Board and Director's Decision, 45 NRC 328, DD-97-9 (1997)(Sierra Nuclear VSC-24 Casks in use at Point Beach, Palisades, and Arkansas Nuclear One); §2.206 Petition of Fawn Shillinglaw and Director's Decision, 45 NRC 135, DD-97-5 (1997)(Sierra Nuclear VSC-24 Casks in use at Point Beach, Palisades, and Arkansas Nuclear One); §2.206 Petition of Toledo Coalition for Safe Energy and Director's Decision, 45 NRC 71, DD-97-3 (1997)(VECTRA Technologies NUHOMS-24P Casks in use at Davis-Besse); §2.206 Petition of Don't Waste Michigan and Lake Michigan Federation and Director's

Decision, 45 NRC 33, DD-97-1 (1997)(Sierra Nuclear VSC-24 Casks in use at Palisades).

The NRC must fulfill its mandate of protection of public health and safety and its philosophy of public participation by establishing a formal rulemaking proceeding to upgrade the regulations governing retrievability, degradation, and to require demonstration of ability of a licensee to unload of dry casks for licensure. Respectfully submitted, C1-d Carol A. Overland

1. 254617 OVERLAND LAW OFFICE 402 Washington st. So.

Northfield, MN 55057

DOCKE: r r,~ 1 DOCKETED USNRC Wednesday, May 27, 1998 Office of the Secretary PE I 101 H1i ~ PAM 1~-t/ ( v3 F~ l~0'/-0) .98 MAY 28 Pl2 :04 ATTN: RULEMAKINGS STAFF U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Rulemakings Staff:

(j) As an environmental organization concerned with the proposed Private Fuels Storage LLC ISFSI on the Goshute Indian Reservation here in Utah, Downwinder's would like to comment on the Prairie Island Coalition's petition for rulemaking regarding the retrievability, degradation, and unloading of spent nuclear fuel stored in casks. We are interested in this petition because the nuclear fuel stored on Prairie Island will eventually be stored in Utah if PFS builds it's proposed ISFSI. We strongly assert that the recent problems of Sierra Nuclear's VSC-24 cask are UNIVERSAL TO ALL DRY CASK STORAGE. The new Sierra TRANSTOR cask is especially vulnerable because of it's dual purpose use. As we are asked to live next to vast amounts of spent nuclear fuel in Utah, we are very concerned that Sierra has not solved the welding problems (See NRC letter to Sierra, February 17, 1998)or adequately addressed the viability of retrievablitiy of the spent fuel rods. Also, the issue of degradation of the casks over time has not been addressed in a meaningful, scientific manner. Sierra has not had a good record with the VSC-24. This hardly speaks well for it's sibling, the TRANSTOR. We fully support the Prairie Island Coalition's petition to initiate formal rulemaking on the issues of retrievability, degradation and unloading of casks used for storage of nuclear waste. Sincerely, Preston Truman Director Winston Weeks Education Director Downwinders, Inc. 239 E. So. Temple,#201 SLC, UT 84111 (801 )521-6128 http://www. downwinders. org E-Mail: wweeks@aros.net

U.S. NUCLEAR REGULATORY COMMISSION RllEMAKINGS&ADJUDICATIONS STAFF OFFICEOFTHESECRETARY OF THE COMMISSION Doc *., t Str.tis res Postmark Oat -2, ;J9/g_r;- £,mtYL__~ ~ l~ ~ Copies Receiv.d --- '---,---- Add'I Copies Pepr<.1 JCed _ _,_ __ s~ ~ IA5 I i.

North American Water Office P.O. Box 174. Lake Elmo, MN 55042 (6 12) "n0-3861 DOCKETED us J May 26, 1998 Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Rulemak:ing and Adjudications staff RE: NRC Docket Number PRM-72-4 Comments of the North American Water Office

Dear Rulemak:

ing and Adjudications staff: °98 MAY 27 P 3 :1 0 0 f I f r,*... C r,.......... Existing rules governing dry cask storage of high-level nuclear waste assume that the waste will not degrade in any significant way during storage, or during necessary maintenance and unloading procedures. With this rational, existing rules define cask maintenance and unloading procedures that are essentially the reverse of loading procedures. There is significant evidence that a number of degradation modes exist for dry cask storage systems, including flawed cask fabrication, neutron flux and irradiation, liquid metal embrittlement, metal creep, hot metal reactivity in water; thermal shock, and perhaps others. Many of these degradation modes are addressed and discussed in the attached memorandum by RudolfH. Hausler ofCorro-Consulta to the North American Water Office (NAWO), which is hereby incorporated into the comments ofNAWO for purposes of this proceeding. Any of these modes of degradation may render reverse loading procedures ineffectual in terms of protecting workers, the public and the environment from unacceptable radiation exposure. This docket is therefore obligated to modify, and accountable for modifying rules, so that they adequately describe and define procedures that ensure the safe management of the waste during a variety of contingency situations that may arise as a result of degradation. To facilitate rule modification, it is now incumbent upon the NRC to produce a public list of all contracts, contract amounts, internal NRC programs and budgets, and the general topic of each contract and program related to high-level nuclear waste degradation occurring during dry cask storage operations. MAY 2 8 1998 Acknowledged by card HHHMt1,n*t uw WWII

U.S. NUCL... r\\R REGULATOiW CO,"1~lSS10N RULE,;1AKiNGS & t, D,JUDICAT!Oi' S S ;_..wF 0, fiCI.: O,-iHi= SECRETARY Or TiiE coi:M:SSION Ooc.umcnt S.atistics

North American Water Office Comments NRC Docket Number PRM-72-4 May 26, 1998 Page two Finally, pending the adoption of adequately modified dry cask management rules, the NRC should not allow licensees to load casks with irradiated fuel. Sincerely, ~~~ George Crocker Executive Director

George Crocker Executive Director North American Water Office P.O. Box 174 Lake Elmo, MN 55042

Dear George,

Rudolf H. Hausler 7804 Pencross Lane, Dallas, TX 75248 Tel: 972 490 8605 Fax:. 972 490 8878 The following are comments with respect t-o-the questions you-asked: The TN-40 spent fuel storage assembly used at Northern States Power has a stainless steel basket assembly for 40- spent fuel rods. This assembly is inserted-int-o-a- ca-rbon steel shell the-internal surfaces of which are coated with Zn/Al (70%/30%). The fuel elements, which are inserted into the rectangular slots of the basket are madefrom-zirca-Hoy cla-dding (Ref. 1): The-storage assembly is to be licensed for 20 years. A number of serious concerns arise with this and similar proposed storage assemblies which µave not been publicly discussed and answered. Interaction between the Zn/Al coating and the borated SFP water This interaction has been evaluated in the laboratory where it was concluded that the evolution rate of hydrogen at SFP water temperature - or slightly higher - was nqt sufficient to raise any concern about the generation of an explosive atmosphere during loading procedures. It was then concluded that "the-small surface area-initially exposed to the borated water during refilling (unloading procedure after 20 years), along with the small generation rates of hydrogen and previous passivation due to the develop~ of a protective oxide layer is not expected to yield a hazardous concentration of hydrogen" (Ref. 3). This conclusion is highly speculative a-nd certainly not verified. It is known that active metals (less noble than hydrogen: E0 Me < E0m) can react explosivelywith liquid or vaporized water at elevatedtemperatures-with-thegenerationofhydrogen: Therateofthis reaction depends on the temperature as well as the rate of which water may be added to the basket during the unloading procedure. Sinre the interior of the basket can easily have a temperature in excess of 600 °F, there has to be specified a precise unloading procedure which must include a cooling process and procedure to safe temperature prior to introducing SFP water into the storage assembly. Furthermore, since there is currently no provision to monitor the interior temperature of the storage assembly, safety measures must be in place, and evaluated, should an explosive reaction occur unexpectedly. Zn/Al and Zircalloy during Storage

Since the TN-40 is surrounded by an outer neutron absorption shield, 4.3" thick and composed of borated polyesterresin contained in long slender aluminum containers, the main resistance to heat transfer is in this shell (the heat conduction of organic resins being much smaller than that of carbon steel}. This means that the inside temperature of the storage assembly is very high and the thermal gradient from the fuel assemblies to the Zn/ Al coated inner carbon steel containment wall is small. This means that when the basket is filled with 40 fuel assemblies, each generating a residual heat flux of 1 KW, the temperature of the Zn/ Al coated carbon steel container wall can reach temperatures of 600 to 800 °F. The cavities will fill with Zn vapor of sufficient partial pressureto absorb on the Zircalloy. This can lead to "liquid metal" embrittlement of the fuel cladding. It has not been shown conclusively to date, other than by idealized modeling, that such cannot happen, even though the potential problem has-been widely discussed*in many places and is currently being worked on again. The crucial question is whether the chemical potential of a Zn/Zircalloy complex is smaller than the chemical potential of Zn at the same temperature. This question cannot be answered by studying the bulk properties of the two metals. Rather, the chemical potential of a Zn/Zircalloy complex at the Zirealloy grajn boundaries must be taken into consideration. If µz.n > µZircalloy at the grain boundaries, then accumulation of Zn can occur at the grain boundaries, and embrittlement will be the consequence. There is no question that the processes involved are exceedingly slow and perhaps barely observable over a one year period. Never the less, the risk ofliquid i;netal embrittlement over a 20 year period is unacceptable at any level because of the damage which can be caused by the disintegration of the fuel elements* during unloading of the basket. It is indeed difficult to understand why the interior of the storage basket is not clad with a non-corroding alloy in order to avoid even to have to consider explosive generation of hydrogen when water comes into contact with the hot Zn/ Al coated surfaces; orthe hiwi temperature degradation of the fuel element cladding when in contact with Zn vapor over a long period of time. General and Conceptual questions The above discussion raises a number of conceptual* questions with* respect to testing and risk assessment:

a. Corrosion tests are generally performed over short periods of time and the results are subsequently extrapolated linearly over time periods 1000 to 5000 times longer without any thought given to the relevance of the linearity assumption. Many corrosion phenomena, however, are known to accelerate with time, such as pitting, crevice corrosion, and intergranular disintegration. More generally, it can be asserted that all metal deterioration processes must be assessed based on known kinetics over long time periods for the results to be relevant. Hence such relevancy has to be demonstrated in all testing, modeling and extrapolation of the deterioration processes which will occur and may occur in connection with spent nuclear fuel storage devices.

b. Test specimens used in materials property evaluations are all too often irrelevant and non-representative of worst case conditions. This requirement of relevancy is essential in risk assessment. In the case of the fuel cladding it is essential that used fuel assemblies be considered. It is already known that the cladding of the fuel elements can deteriorate during normal operations of the energy generation cycle. Embrittlement has been observed at the end caps. How-will such damaged or partially damaged assemblies hold up to loading, storage and unloading procedures? What are the effects of thermal shock on already weakened claddings?

The need for relevant risk assessment and remediation procedures Because of the above considerations it is imperative to detail and submit for independent review all risk assessment algorithms, and accident prevention and remediation procedures in all areas of operation, such as loading, storage and unloading. While the individual SAR' s and specifically Bulletin 96-04 have attempted to do this, it is clear that many assumptions made to date have not been verified, that insufficient effort has been exerted toward such verification, and that many test procedures have been totally insufficient in the interest of rapid commercialization of such spent nuclear fuel storage devices. It is therefore considered reasonable to request of this docket complete and specific answers to the following questions:

a. While the top of the cask is bolted onto the body, the bottom is* welded onto the cylinder which constitutes the side wall (9.5"). Massive welds of this kind have a tendency to crack because of residual stresses. The same holds true for the trunions.

Hence: - What are the residual stresses; how are they measured; and how do they ~ffect performance under hot and cold storage conditions. - What are the NOT inspection procedures and how accurate are they. Since thermal fatigue is much more pronounced in the storage vessels the question arises whether boiler vessel codes are adequate design* criteria for the st.qrage module. - If weld cracks, fatigue cracks, etc. open from the atmospheric side, what specific measures have been taken to verify the absence of crack growth due to atmospheric conditions during 20 year storage.

b. What are the effects of neutron flux and irradiation on materials' properties. Have these effects (or the absence thereof) been verified? If so, how, and what are the long term extrapolation algorithms?

c. The questions of the liquid metal embrittlement and the Zn vapor effects on intergranular deterioration of the Zircalloy need to be reviewed and verified. Outline current work (objective and methodology) and procedure to establish risk assessment for long term storage. Special attention must be given to relevancy of procedures and test specimens. The major objection is to language such as " --these effects are

deemed to be minimal". Risk assessment procedures have been developed in various industries which are much more precise than fuzzy language of this kind.

d. The problem of thermally quenching a hot storage basket in fuel pool water, and the concomitant reactions of reactive metals with liquid water or the vapors thereof, must be scientifically detailed with explicit procedures and quantitatively verified reaction rate assessments. Similarly, safety measures must be explained in-detail for the situation that runaway (explosive) reactions occur despite adequate controls of the quenching procedure.

e. The docket should outline assessment and verified counter measures of all possible thermal shock events.

f. Did NRC or the vendors assess the composite risks of nuclear waste leakage of any kind from structurally impaired waste cask storage systems. (Monte Carlo simulations of probabilities of all known and discussed risk factors). Such assessments must be made public with all assumptions detailed and verified and subject to third party review.

REFERENCES

1. Information Handbook on Independent Spent Fuel Storage Insatallantions; M.G: Raddatz, M.D. Waters, US Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, NUREG-15 71, December 1996

2. Transnuclear Inc. TN-32 Topical Report, NRC Questions 07/19/96, E-15011, Aug. 22 1996.

Response to NRC Bulletin 96-04

3. Northern States Power Company. Final Response to Bulletin 96-04. November 41996,

DOCKET NUMBER State of Utah PETITION RULE PRM 12-~ DEPARTMENT OF ENVIRONMENTAL QUALITY {t 3 Fl<\\ z. D'fo) 0 OFFICE OF THE EXECUTIVE DIRECTOR \\.....::./ Michael 0. Leavitt Governor Dianne R. Nielson, Ph.D. Executive Director Brent C. Bradford Deputy Director May 22, 1998 Secretary 168 North 1950 West P.O. Box 144810 Salt Lake City, Utah 84114-4810 (801) 536-4400 (801) 536-0061 Fax (801) 536-4414 T.D.D. www.deq.state.ut.us Web U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Secretary:

Re: Prarie Island Coalition Petition for Rulemaking (Docket No. PRM-72-4) Enclosed are the state of Utah's comments on the above-referenced petition. These comments are also being forwarded by e-mail to Emile Julian, Office of the Secretary, at the following e-mail address: ELJ@nrc.gov. Thank you for your consideration of the State's comments. Best regards, Executive Director enclosures MAY 2 8 1998

I .... ~ --- ---~- -- ----

Before the Nuclear Regulatory Commission In the Matter of Prairie Island Coalition Petition for Rulemaking Docket No. PRM-72-4 May 22, 1998 Comments by the State of Utah Introduction and General Comments The State of Utah generally supports the amendments to 10 CFR Part 72 requested by the petitioner Prairie Island Coalition (PIC). The intent of the requested amendments is to more specifically define parameters related to dry cask storage including fuel degradation and fuel retrievability and unloading. On these issues, the present regulatory framework is quite undefined. Neither cask manufacturers nor the public is quite clear about the intent and effectiveness of Part 72 regulations. The public perception is that this an ongoing experiment taking place in the field. Furthermore, the public lacks confidence that the regulations protect the public health and safety, which is the ultimate goal of Title 10 and the underlying federal legislation. Thus, there is an important need to put Part 72 regulations on a better framework. The State of Utah appreciates the Commission's willingness to open these issues to public debate. As you are aware, the State of Utah has intervened in the licensing of the proposed Private Fuel Storage, LLC (PFS) dry cask storage facility in Skull Valley, Utah. Clarifying and strengthening Part 72 regulations will better protect Utah residents should the PFS facility be licensed. Strengthening Part 72 storage regulations will also better protect people along the transportation routes. The State believes the present Part 72 regulations and set of dry storage parameters are too vague and do not provide sufficient guidance for the Staff or cask designers. Dry storage systems have operated with extremely thin safety margins. Recent safety problems with the VSC-24 casks were never anticipated when Part 72 regulations were first adopted, and thus the regulations need to be changed to cope with the present dry storage reality.

The Staff has attempted to model the regulatory approach to dry storage casks after that for transportation casks. But the operating environment is entirely different. In almost all cases, fuel shipped in transportation casks has started out in spent fuel pool. Pool storage is a relatively benign environment. Fuel is stored at 50°C in irradiated fuel pools, before being shipped to reprocessing plants or alternate storage pools. The cladding temperature is therefore relatively cool. Transportation casks are used for a brief period, after which irradiated fuel and the cask interior can be inspected. Damaged fuel can be containerized and faulty casks can be repaired or replaced. Transportation casks are subjected to hypothetical testing conditions. Dry storage conditions are vastly different. Under dry cask storage, the maximum cladding temperatures can approach 380°C. The high heat places fuel cladding under continual stress for prolonged periods. This stress, without a countervailing pressure as exists in a reactor, can lead to degradation of cladding and make retrieval difficult. Unloading a dry storage cask has never been carried out, although 10 CFR § 72.122(1) and the Certificates of Compliance require this ability. It has been over four years since the NRC directed Consumers Power to unload VSC-24 cask #4 and this has still not been carried out. Except for minor checks of the exterior of dry storage casks for radiation and temperature, irradiated fuel is never retrieved and inspected and fuel cladding, in particular, not checked for cladding defects following cask closure. Following an undefined storage period, irradiated fuel with stress-weakened cladding must be transported to a final high-level waste repository. The final transportation conditions will therefore also have changed. Reports on transportation risk prepared for and by the Nuclear Regulatory Commission in the '70's never anticipated these changed conditions. Specific Comments Degradation of Spent Nuclear Fuel The PIC petition requests an amendment of Part 72 regulations to define the parameters of fuel cladding degradation. 10 CFR § 72.122(h) states that spent fuel cladding must be protected during storage against degradation or the fuel must be configured such that degradation will not pose an operational safety concern. But 10 CFR § 72.122(h) does not define the parameters. 10 CFR § 72.236 further requires cask designers to specify parameters of spent fuel that can be stored in a particular design, and requires spent fuel to be stored safely for a 2

minimum of 20 years(§ 72.236(g)). The cask must be designed to provide adequate heat removal capacity without active cooling systems (§ 72.236(f)). But the word "adequate" has been left for the Staff to define. The staff has in tum required cask designers to follow a prescription laid out in a 1987 Battelle report. 1 The Battelle report estimates an 0.5% cladding failure under Battelle's methodology which takes into account the fuel bumup and fuel assembly type to predict the appropriate post-operation cooldown period (p. 3.9). The cladding is one barrier to the release of radioactivity to the external environment. Cladding also provides structural integrity. Cladding which is severely degraded cannot easily be retrieved and handled. At Palisades the grapple atop a spent fuel assembly that remained in the reactor for five fuel cycles, broke off when the fuel assembly was moved. For some fuel assemblies, the Battelle methodology translates to a maximum fuel cladding temperature of 380°C. The PIC petition points out that if such a cask is opened and relatively cool water is inserted, the thermal shock may further degrade spent fuel cladding. The Battelle methodology does not account for subsequent problems which have occurred. For example, the VSC-24 has had major welding problems. If through these welding flaws, helium is allowed to escape from the cask, and helium is replaced by air, primarily nitrogen, the temperatures may rise precipitously since helium is approximately five times more heat conductive than nitrogen. The Battelle report shows that internal temperatures may rise an additional 80°C. The implication is that many more fuel rods than 0.5% will have degraded cladding. Therefore, 10 CFR § 72.122(h) must be amended to increase the margin of safety. Ready Retrievability If a greater percentage of fuel cladding than assumed by the NRC and Battelle has degraded, this has important implications for transportation and handling accidents and sabotage. All accident consequences would be increased, particularly if cask welds are faulty. Moreover, cask handling accidents could result in damaged fuel cladding. Although previous studies have tended to minimize the consequences of such accidents, they have not given adequate consideration to the effects of fuel degradation on accident consequences. In particular, cask manufacturers have not properly examined the consequences of cask drop accidents on spent fuel cladding. For example, the HI-1 Levy, IS and BA Chin, "Recommended Temperature Limits for Dry Storage of Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas," PNL-6189, May 1987. 3

STORM SAR refers to a report by Lawrence Livermore Labs2 to show that spent fuel cladding will remain intact during a cask drop accident.3 Despite the title of the report, it does not deal with spent fuel assemblies, but only with non-irradiated fuel assemblies. Irradiation within a reactor makes fuel assemblies more brittle and less resistant to impact. Thus, the Holtec analysis is not conservative. If irradiated fuel were involved in a cask drop accident, the cladding could shatter, making retrievability exceedingly difficult. Further, storing irradiated fuel in a high temperature environment for extended periods will degrade the cladding and make it more vulnerable in transportation accidents. Similarly, previous studies on the impacts of sabotage erroneously underestimated the consequences of sabotage events on dry storage casks. This is because the studies evaluated non-irradiated fuel, rather than irradiated fuel, which is more brittle and degraded due to prolonged heating in a storage cask. Sabotage tests conducted by Sandia and Battelle employed non-irradiated fuel. 4 Unloading Irradiated Fuel The Prairie Island Coalition also requests an amendment to require licensees to demonstrate the ability to unload irradiated fuel safely before a cask may be used at an ISFSI. The State of Utah supports this request. Unloading may be required under several circumstances. As at Palisades, NRC may discover, from Quality Assurance documentation, faulty welds and require unloading of spent fuel and transfer to another storage cask. NRC Staff have uncovered numerous recent examples of welding problems with VSC-24 casks.5 Or a cask may be damaged in a storage accident or by sabotage. Unloading may also be required in order to transport and dispose of irradiated fuel at a federal repository. All Certificates of Compliance issued by the NRC require this unloading capability. However, there is a great deal of difference between demonstrating unloading ability on paper and actually being able to carry out this unloading ability in 2 Lawrence Livermore Laboratories, "Dynamic Impact Effects on Spent Fuel Assemblies," UCID-21246, 1987. 3 Holtec International, HI-STORM SAR, Section 3.5. 4 Schmidt, EW et al, "Shipping Cask Sabotage Source Term Investigation," NUREG/CR-2472, December 1981. 5 Letter from CJ Haughney, NRC to K Moekel, Sierra Nuclear Corporation, February 12, 1988. 4

reality. The events at the Palisades reactor make this abundantly clear. Thermal shock may degrade fuel assembly cladding and steam, under pressure, may leach radionuclides from irradiated fuel and into the pool area. Thus, the request by PIC for an actual physical demonstration of unloading ability before a cask may be used at an ISFSI is eminently reasonable. For a stand alone ISFSI away from a reactor pool, such as at the proposed PFS facility in Utah, demonstrating the capability to unload casks is especially important given that the reactor and the spent fuel pool which supplied the irradiated fuel may no longer exist, be inoperable or decommissioned. For this reason, unloading ability should also be required at an independent stand-alone away-from-reactor ISFSI. Included in each reactor SAR is a calculation showing that a fuel pool accident cannot result in radiation exposures greater than 5 rems at the site boundary. These calculations assume a cask or fuel assembly is dropped into the fuel pool. However, these calculations in reactor SAR's do not assume a driving force moving radionuclides out of damaged fuel assemblies. Thus, only gases in the rod plenum are assumed to be released to the external environment. But if an accident involves a dry storage cask, water would be converted to steam within the cask. If the cladding is degraded, the steam would mix directly with irradiated fuel and leach out radionuclides. This radioactive steam, under pressure, may be forced out of the cask and be carried into the fuel pool area and ventilation system. Doses at the boundary may exceed 5 rems. This accident is serious enough for the Commission to require a detailed analysis and for utilities to ensure that this accident cannot take place. NRC regulations should be drafted that require a physical demonstration of this unloading ability to ensure that the public health and safety is protected. Conclusion The State of Utah supports the petition by PIC to amend Part 72 regulations. The regulations should be amended to provide a greater safety margin in dry storage. If welds are faulty and degrade, helium may be lost and internal temperatures may rise by as much as 80 °C. The number of fuel rods with degraded cladding will thereby rise. To allow for this contingency, the NRC must require a longer cooling time or greater spacing between fuel assemblies. Irradiated fuel with degraded cladding may not be easily retrieved. Finally, the NRC should require an actual physical demonstration of unloading ability before a storage cask can be employed. 5

DOCKET NUMBER PETITION RULE PRM 7 Z-1/ (fo3 FP. l2o'{o)

• TRANSNUCLEAR, INC.

(!}) Secretary, U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Rulemakings and Adjudication Staff

Subject:

Comments regarding petition for rulemaking.

Reference:

10 CFR Part 72, Docket No. PRM-72-4

Dear Sirs:

In the March 12, 1998, issue of the Federal Register (Volume 63, Number 48), the U.S. Nuclear Regulatory Commission (NRC) solicited public comment on a petition for rulemaking submitted by the Prairie Island Coalition. Transnuclear, Inc. has reviewed the information submitted by petitioner and believes that there is no need for the NRC to proceed with rulemaking.

1. Degradation of spent fuel during storage (10CFR72.122(h))

The use of a cladding temperature limit and inert gas cover has been well established as a means of preventing both cladding rupture and oxidation of UO2 to U3O8 in fuel rods with minor breaches. (See for example documents PNL-4835, PNL-6189, and UCID-21181.) Transnuclear metal casks are designed to store fuel below a maximum allowable fuel cladding temperature in an inert medium (helium). The approved temperature limit accounts for the effects of cladding temperature, decay time, burnup and fission gas build-up. This method limits the possibility of failure of the maximum temperature fuel rods and also addresses the issue of spent fuel degradation during storage. The TN-40 cask is not subject to failed welds and to fuel degradation due to cask seal failure as a result of helium gas release. The TN-40 cask is designed with a sophisticated closure monitoring system which virtually guarantees that a helium atmosphere will be maintained in the cask. The citation of Susan Shankman's letter to Sierra Nuclear seems to imply a misunderstanding of the TN-40's containment boundary which uses metal seals rather than a welded lid. Similarly, the reference to the NSP SAR "requiring" seal replacement seems to be a misunderstanding of the TN-40's dual seal and monitored overpressure system. Seal replacement is not required to maintain containment; it would be a response to the highly unlikely event of seal leakage. Failure of the seals would not

Page 2 of 3 result in air entering the TN-40, thereby causing fuel degradation, but rather leakage of helium from the overpressure system into the cask. Furthermore, since the overpressure system is monitored, a loss of pressure in the system will trigger an alarm that will permit adequate time for site personnel to evaluate the situation and initiate corrective action, if necessary. Thermal shock will not present a significant problem when the casks are reloaded with pool water prior to unloading. Unloading of spent fuel is typically part of a fuel transportation cycle. Fuel temperature limits as high as 570°C have been licensed for transportation packages. Unloading of fuel from a transportation cask into a spent fuel pool without causing fuel degradation has been demonstrated in both the U.S. and extensively in France at the La Hague reprocessing facility. In the case of unloading fuel from a storage cask, the thermal shock phenomenon will be much less significant due to the lower fuel temperature (usually less than 300° C). In addition, the thermal shock is minimized by following procedures that allow the fuel to gradually cool down to the boiling point of water (100° C) before being submerged into the pool. Thereafter, it can be cooled down to the pool water temperature without generating significant thermal stresses in the fuel cladding.

2. Retrieval of spent fuel (10CFR72.122(1))

Cask licensees are required to demonstrate the ability to retrieve spent fuel under accident conditions, which are more severe than the long-term normal conditions. Storing the fuel in a helium environment and at the relatively low temperatures discussed above assure that the condition of the fuel at the onset of storage is maintained throughout the storage life, and thus the fuel is readily retrievable when required. The incident with the TN-24P at INEL cited in the petition concerned a consolidated rod test program which stored fuel rods in metal canisters. This situation is not germane to at reactor spent fuel storage of bare assemblies-first because fuel rods were placed in canisters, and second because one canister was loaded close to the thermal limit and thermal expansion of the canister caused the fit to be so tight that removal of the canister was difficult. It is important to note that the TN-24P cask at INEL was a prototype which was not designed or analyzed by Transnuclear for the storage of canistered rods. Nor was the TN-24P reviewed or licensed by the NRC. Transnuclear knows of no case in which retrieval of spent fuel assemblies from transport casks, which are very similar to storage casks, has been a problem.

3. Unloading of Spent Nuclear Fuel It is unnecessary for regulations to be amended to require demonstration that storage casks such as the TN-40 can be unloaded. The TN-40, and other Transnuclear metal storage casks, are based on the well proven technology of metal transport casks of which hundreds of unloadings have been demonstrated on a world wide basis with none of the

Page 3 of 3 problems postulated by the petitioner. It is sufficient that the NRC requires licensees to prepare unloading procedures which are then reviewed and approved by the NRC prior to licensing. In conclusion, none of the points raised by petitioner require amendment ofNRC regulations. The NRC should not devote its limited resources to rulemakings which do not substantively contribute to protecting the health and safety of the public. s:pt}' t hael E. Mason Chief Engineer

DOCKETED USNRC /W'orthem S1"tes Powe, Company 414 Nicollet Mall Minneapolis, MN 55401 p "98 NAY 26 P 3 :28 Telephone (612) 330-5500 OFrl( f

• . )

RULi~ * -'~ ;::; DOCKET NUMBER ADJ - I I ~l) U01r:. 1,. : *. _,T,.i:"FF PETITJON RULE PRM 7~-t/- May 20, 1998 Secretary Rulemaking and Adjudications U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Petition for Rulemaking PRM-72-4 ( r,,3F,< t ;;io,; o) @ Northern States Power would like to offer the following comments regarding the recent Prairie Island Coalition Petition for Rulemaking regarding 10 CFR Part 72 (Docket No. PRM-72-4). The petition requests that the Commission undertake rulemaking related to storage and unloading of spent nuclear fuel in dry storage casks. Northern States Power endorses the comments recently submitted by the Nuclear Energy Institute (NEI) on this petition. NEI is correct in noting that current NRC regulations have been proven to adequately address all the issues raised in the petition. The petition presents no new information to support the contention that the current regulations are inadequate. NEI is also correct in noting that two recent NRC Director's Decisions also provide sufficient information to conclude that the Petition for Rulemaking should be denied. These recent Director's Decisions 1*2 were in response to 2.206 petitions which raised identical issues as PRM-72-4. Both of these 2.206 petitions were denied in their entirety by the NRC. These Decisions provide a factual basis to conclude that the current NRC regulations regarding fuel storage and unloading are more than adequate to protect the public health and safety. The regulations have been proven through experience to be effective and complete, and no additional NRC rulemaking is warranted at this time. 1 Director's Decision DD-97-18, August 29, 1997 2 Director's Decision DD-98-02, February 11, 1998 MA¥-2 8 1998-- Acknowledged by card......... -

U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Oocum I t... ti r. s Postmark Date ---~~-~_,_LCJ...:.8 ___ _ Copies Received _ _ _,_.....;/~ --- Add'! Copies Reprod ..,d _ __., ___ _ SplJia~~t~ ~~ ~

?~ -~IDs

Secretary Rulemaking and Adjudications 05/20/98 Page 2 of 2 In summary, Northern States Power is confident that current NRC regulations adequately address all the issues raised in the petition and that no additional rulemaking is warranted. If you have any questions related to our comments please contact Jon Kapitz at 612-388-6758, ext. 4819. /btJ Roger O. Anderson Director, Nuclear Energy Engineering

May 20, 1998 Secretary ~CKETEO USNRC NUCLEAR ENERGY INSTITUTE

• 98 HAY 26 P 1 :38 Of-! :.

RUL. ADJUC1 DOCKET NUMBER Lynnette Hendricks

DIRECTOR, PLANT SUPPORT NUCLEAR GENERATION PETITION RULE PAM 1 ~ -1 Rulemakings and Adjudications

( (p3F~ /:JOIIO) U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Prairie Island Coalition; Receipt of Petition for Rulemaking 63 Fed. Reg. 12040 (March 12, 1998) Request for Comments PROJECT NUMBER: 689

Dear Sir:

The Nuclear Energy Institute (NEI)1 submits the following comments on the Prairie Island Coalition Petition for Rulemaking regarding 10 CFR Part 72 (Docket No. PRM-72-4). The petition requests that the Commission undertake rulemaking to examine certain issues related to the potential for thermal shock and corrosion in dry cask storage. The petition also requests that the Commission amend its regulations to define the parameters of acceptable degradation of spent fuel in dry cask storage. Finally, the petition requests that the Commission define the parameters ofretrievability of spent nuclear fuel in dry cask storage and require licensees to demonstrate safe cask unloading ability before a cask may be used. The petition for rulemaking was included in a 10CFR 2.206 petition that requested the NRC to revoke the Prairie Island Independent Spent Fuel Storage Installation material license (SNM-2506). This petition raised the same issues regarding 'thermal shock' of spent fuel assemblies during unloading and other general issues related to cask unloading as the current petition for rulemaking. The NRC has 1 NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including regulatory aspects of generic operational and technical issues. NEI members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry. MAY 2 8 1998 Acknowledged by card.................................. 17/6 I STREE T, N W SUITE 400 WA SH IN GT O N, DL 20006-370 8 l' H O N E 20 2 739 8000 f AX 2 0 2 78 5,40 I 9

U.S. NUCLEAR REGUI.ATORY COMMISSION AULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OFTHECOM.1S~IO,J Copies Received Add'! Copies Repr d0 ed _ Spec~ ~ / 'is/Ds

Secretary, Rulemakings and Adjudications May 20, 1998 Page 2 issued a Director's Decision denying the 2.206 petition in it's entirety. This petition for rulemaking argues that the specific problems alleged at Prairie Island exist in the industry as a whole. The NRC response not only clearly documents that there is no specific issue with the ability to unload a TN-40 cask at Prairie Island, it also provides sufficient evidence to conclude there is no generic issue regarding safely unloading dry spent fuel storage casks, and hence no need for generic rulemaking. The nuclear industry is confident that current NRC regulations adequately address all the issues raised in the petition and that no additional rulemaking is warranted. The petition presents no new factual basis to question the adequacy of the existing NRC regulations. The petition does not present any relevant technical, scientific or other data to support the need for rulemaking. Nor does the petition identify any specific case(s) were the NRC rules are deemed deficient. Most importantly, these issues have all been addressed in detail in two recent NRC Director's Decisions2,3 denying petitions filed under 10CFR 2.206 concerning the Prairie Island ISFSI license (SNM-2506). Specific Comment on PRM-72-4

1. Paragraph 13 requests NRG to solicit and review information regarding thermal shock and corrosion inherent in dry cask storage and usage and to define the parameters of degradation of spent nuclear fuel in dry cask storage acceptable under 10 CFR 72.122(h).

The existing rules of Part 72 adequately protect against the potential for fuel degradation. The requirements of 10 CFR 72.122(h) contains sufficient requirements for fuel integrity during storage. Experience, testing, and computer modeling have all shown that the combination of inert atmosphere and temperature limits provide a robust basis to conclude that the integrity of the fuel will be maintained during the licensed storage period. The petitioner presents no factual information to suggest otherwise. There is also no factual evidence presented to suggest that current unloading procedures will result in damage to the spent fuel. On the contrary, there is a wealth of experience to demonstrate that the current equipment and procedures will not result in fuel damage. While the petition notes the limited experience in unloading spent fuel storage casks, it ignores the vast experience base in unloading spent fuel shipping casks. Thousands of spent fuel assemblies have been successfully unloaded from shipping casks without damage. It is important to note that most, if not all of these shipping casks are designed for fuel temperatures 2 DD-98-02, director's Decision Under 10 CFR 2.206, February 11, 1998 3 DD-97-18, Director's Decision Under 10 CFR 2.206, August 29, 1997

Secretary, Rulemakings and Adjudications May 20, 1998 Page 3 higher than storage casks. The higher temperatures in the shipping casks results in greater thermal stress during cooldown, and the experience unloading these shipping casks without fuel damage provides confidence that the lower stresses experienced in storage cask unloading will certainly not result in fuel damage. The NRC staff has presented similar conclusions in the recent Director's Decision2 "The NRC staffs judgment that there is reasonable assurance that the TN-40 casks can be safely unloaded comes from a variety of experiences related to the use and storage of radioactive materials. Among these experiences are the dry-run exercises that were performed to verify key aspects of unloading procedures for the TN-40 cask; related research sponsored by the commercial nuclear industry, the U.S. Department of Energy, and the NRC; actual loading and unloading of transportation casks; loading of storage casks; handling of spent fuel assemblies under various conditions; and performing relevant maintenance and engineering activities associated with reactor facilities." Given the fact that (1) fuel integrity will be maintained during normal storage by the inert atmosphere and (2) the return of a cask to the spent fuel pool for will not result in fuel degradation that would result in operational safety problems, there is no need to amend 10 CFR 72.122(h), per the petitioners request.

2. Paragraph 14 requests NRG to define the parameters of retrievability required under 10 CFR 72.122(l).

The argument pertaining to the inadequacy of unloading procedures presented by the Petitioner centers on the lack of an actual example of the unloading of a dry-storage cask at a commercial reactor facility. The petition ignores the vast experience with successfully unloading shipping casks. As discussed in DD-97-18, 'Among these experiences are the dry-run exercises that were performed to verify key aspects of unloading procedures for the TN-40 cask; related research sponsored by the commercial nuclear industry, the U.S. Department of Energy, and the NRC; actual loading and unloading of transportation casks; loading of storage casks; handling of spent fuel assemblies under various conditions; and performing relevant maintenance and engineering activities associated with reactor facilities.' The NRC staff has reviewed the information submitted by the Petitioner and has determined that the licensee could, if necessary, unload a TN-40 cask and has not, therefore, identified a deficiency in 10 CFR 72.122(1) or failure to comply with 10 CFR 72.122(1).

Secretary, Rulemakings and Adjudications May 20, 1998 Page 4

3. Paragraph 15 requests NRC to require demonstration of safe cask unloading ability before a cask may be used at an ISFSI.

10CFR 72.122 (f) requires that systems and components that are important to safety must be designed to permit inspection, maintenance and testing. 10CFR 72.24 (p) requires that a license applicant describe the preoperational testing and initial operations program for the facility. These NRC requirements already provide the necessary assurance that a licensee demonstrate the ability to safely retrieve the spent fuel. Therefore, no additional NRC rulemaking is required in this area. Summary The current NRC regulations adequately address the safety concerns outlined in the petition. The petition does not present any relevant technical, scientific or other data to support the need for rulemaking. Nor does the petition identify any specific case(s) were the NRC rules are deemed deficient. For these reasons, NEI believes that these issues do not warrant rulemaking, and the Petition for Rulemaking should be denied. We appreciate the opportunity to comment on the proposed petition. If you would like to discuss our comments further, please contact me at (202) 739-8109 or Alan Nelson at (202) 739-8110 or by e-mail (apn@nei.org). Sincerely, Lynnette Hendricks LH/APN/tnb

NEVADA NUCLEAR WASTE TASK FORCE, INCORPORATED Alamo Plaza 4550 W. Oakey Blvd. DOCKET NUMBER ~:o~ c.c..,, Suite 111 C r---. Las Vegas, NV 89102 p T p ~ 1P-1 Orr r; ,..i:;:~- 702-248-1127 FAX 702-248-1128 800-227-9809 Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 RE: Petition for Rulemaking 10 CFR Part 72 [Docket No. PRM-72-4] George Crocker Prairie Island Coalition ( l,'3FI< /~o'/o) (J) Cf) May 12, 1998 Enclosed are comments to the referenced petition for rulemaking. Also included is one attachment. The Task Force appreciates the opportunity to comment on this important matter. ~

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~ / ~~/4L_ udy Tfe'ichel Executive Director MAY 2 1 1998 c:, c::O (,/) ("")

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U.S. NUCLEAR REGULATORY COMMISSfO, RULEMAKINGS & ADJUDI ATIONS STAFF OFFICE OF THE Er~,: OFTHECOMMI Do., n+"" Postmark Date 6 I 'I 1~ Copies Recelv.., __ l Add'I Cop'es Rei:: o _.!i. s~ ~%os

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[Docket No. PRM-72-4] ".:;:_ t ;1-1 .. - - *)' i ?:: George Crocker ~ U1 Prairie Island Coalition } > t:-:~, 77 I..,.) -n The petitioners are requesting that the NRC amend or revise its regulations governing high-level nuclear waste cask certification -- specifically, that any cask for the storage, transportation or disposal of irradiated fuel be required to have proof of contingency plans to deal with the potential for fuel and fuel cladding degradation in procedures for unloading. This petition for modification of licensing requirements would apply to :-ill containment systems for the storage of irradiated fuel outside of reactor pool storage as well as the certificate of compliance (COC) for any storage, management or disposal facility requiring an NRC license. It must be assumed that any cask either could or would require unloading procedures. Casks while at a storage facility at the reactor site will require unloading in the event that any failure is detected in cask integrity. They will also be unloaded if or when fuel is to be transferred to a transportation cask to be shipped away from its original location. Ultimately, it is assumed that fuel will again be unloaded from transportation and storage casks and transferred to disposal containers. All are required to be licensed by the NRC and all casks and maintenance facilities must be able to demonstrate unloading and management procedures that can adequately and safely manage degraded fuel and/or fuel cladding. The current certification requirements do not satisfactorily address the unloading of such casks. When licenses were granted it was apparently assumed that unloading would simply be a matter of reversing the loading procedures. This is not the case. At Palisades where a VSC-24 has been known to have a defective weld since shortly after it was loaded in July 1994, there is still no approved method for removing the fuel. Even though a workable document for unloading a cask was required by the COC, no such plan was available when the VSC-24 was certified or when casks began to be loaded, or indeed even now. Had the NRC requirements for dry cask l.:ertification been as stringent as proposed now and also by petitioners previously, this design would not have been approved and the dilemma with the Palisades VSC-24 and possibly other casks would likely have been avoided. This rulemaking for modification of 10 CFR Part 72 is urgently needed and will have broad implications. The licensing and performance of all systems for nuclear waste and spent fuel management, storage, transportation, and disposal are of concern to the general public. Communities with operating commercial reactors as well as those being considered for waste sites have serious doubts about the safety of existing and future facilities. The Nevada Nuclear Waste Task Force is submitting these comments because if either the proposed Yucca Mountain repository or a Congressionally mandated interim CJ C:O U) ('") z::::s::

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storage facility become a reality in Nevada, the NRC licensing process will be the ultimate public forum. The NRC license for such facilities as well as for the casks arriving or to be used at either or both locations will establish the final, irreversible decision regarding public safety and environmental protection. One of the primary functions to be performed at both the above ground facilities at a repository or an interim storage site will be cask unloading. This will occur after the casks have traveled perhaps thousands of miles and may or may not have been subject to vibration, jarring, shaking, or accidents. Tested and approved unloading procedures, in the event that unforeseen cladding or fuel degradation have occurred, are essential. Nevadans are aware of and paying increasingly close attention to operation and performance at currently licensed NRC facilities, and especially to the effectiveness of the licensing process NRC officials must be aware that the issuance of a license for cask designs and cask maintenance facilities under the regulations of 10 CFR Part 72 "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste" has direct implications for all communities under consideration for interim storage and disposal sites now or at a future date. Because of the dangers of radioactive contamination, the licensing process must include far broader public input and involvement and be impeccably thorough, complete, and trustworthy. 2

DOCKETED USNRC [7590-01-P] NUCLEAR REGULATORY COMMIS~N MAR -6 p 3 :20 10 CFR Part 72 O r-.- 1** [Docket No. PRM-72-4) Prairie Island Coalition; Receipt of Petition for Rulemaking DOC E U BER p R 'lfl-'f AGENCY: Nuclear Regulatory Commission. ( ~3Ft<. 1~0¥0) ACTION: Petition for rulemaking; Notice of receipt.

SUMMARY

The Nuclear Regulatory Commission (NRC) has received and requests public comment on a petition for rulemaking filed by the Prairie Island Coalition. The petition has been docketed by the Commission and has been assigned Docket No.

PRM-72-4. The petitioner requests that NRC undertake rulemaking to examine certain issues addressed in the petition relating to the potential for thermal shock and corrosion in dry cask storage. The petitioner requests that the NRC amend its regulations that govern independent storage of spent nuclear fuel in dry storage casks to define the parameters of acceptable degradation of spent fuel in dry cask storage. The petitioner also requests an amendment to the regulations to define the parameters of retrievability of spent nuclear fuel in dry cask storage and to require licensees to demonstrate safe cask unloading ability before a cask may be used at an Independent Spent Fuel Storage Installation (ISFSI).

U.S. NUCLEAR REGULATORY COM -.,S10 RULEMAKINGS & ADJUDICATIONS STAFF OFACEOFTHESECRETAR OF THE COMMISSIO Oocllnent Statistics Postrnmk Date _______ _ CoplasRGCGIIV8d ______ _ Add'I Raprocktced ___ Special Distribution, ______ _

2 f'r)OA.j ~ (, I / <j l:f 8 DATE: Submit comments by (=t'S deys followit 19 publiel!ltion in u,e Fede, al ~egister). Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before this date. ADDRESSES: Submit comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Rulemakings and Adjudications staff. Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4: 15 pm on Federal workdays. For a copy of the petition, write: David L. Meyer, Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. You may also provide comments via the NRC's interactive rulemaking website through the NRC home page (http://www.nrc.gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905 (e-mail: CAG@nrc.gov). FOR FURTHER INFORMATION CONTACT: David L. Meyer, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone: 301-415-7163 or Toll Free: 1-800-368-5642 or E-mail: DLM1@NRC.GOV.

3 SUPPLEMENTARY INFORMATION:

Background

The Nuclear Regulatory Commission received a petition for rulemaking submitted by George Crocker on behalf of the Prairie Island Coalition (PIC) in the form of a letter and an attached document addressed to L. Joseph Callan, Executive Director for Operations, NRC, dated August 26, 1997. Most of the issues presented in Mr. Cracker's letter and the attached document pertain to a petition filed under 10 CFR 2.206 regarding dry storage cask regulations that has been reviewed by the NRC Office of Nuclear Reactor Regulation (NRR). See 62 FR 53031. The resolution of these issues is presented in a decision published by the Director, NRR (DD-98-02; 2/11/98). This notice pertains to paragraphs 13, 14, and 15 on page 3 of the document attached to the August 26, 1997, letter from PIC. These paragraphs contain a request for rulemaking under 5 U.S.C. 553(e) of the Administrative Procedure Act (APA). The NRC has determined that the issues presented in paragraphs 13, 14, and 15 of the PIC document constitute a petition for rulemaking under 10 CFR 2.802. Paragraph 13 requests NRC to solicit and review information regarding thermal shock and corrosion inherent in dry cask storage and usage and to define the parameters of degradation of spent nuclear fuel in dry cask storage acceptable under 10 CFR 72.122(h). Paragraph 14 requests NRC to define the parameters of retrievability required under 10 CFR 72.122(1). Paragraph 15 requests NRC to require demonstration of a safe cask unloading ability before a cask may be used at an ISFSI. These requests do meet the sufficiency requirements for a petition for rulemaking under

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

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

6 addressed degradation of spent nuclear fuel due to the loss of helium from failed seals or due to the passage of time. Retrievability of Spent Nuclear Fuel The petitioner also requests an amendment to the regulations in 1 O CFR Part 72 that govern storage of spent nuclear fuel in dry storage casks to define the parameters of retrievability of spent fuel required by the NRC under 10 CFR 72.122(1). Section 72.122(1) provides that spent fuel storage systems must be designed to allow ready retrievability of the spent fuel for future processing or disposal. PIC is concerned that the NRC has not taken into account the potential problems that may be encountered in unloading a cask to retrieve spent fuel. In support of its claim, PIC cites an April 16, 1997, memorandum from Jack Roe, NRC, to Cynthia Pederson, NRC Region Ill, and asserts that this memorandum is evidence that NRC has not taken into account possible problems with retrieval of spent fuel. The petitioner also cites a study of the TN-24 cask conducted by the Idaho National Engineering Laboratory (INEL) in 1990, which involved opening TN-24 casks that contained canisters of spent fuel assemblies that had been stored for several years. The petitioner contends that the INEL study found the thermal damage so great that some canisters containing spent nuclear fuel could not be retrieved from the cask. The petitioner believes that the INEL study and the cited NRC memorandum, copies of which are attached to the petition, demonstrate that spent nuclear fuel cannot be reliably retrieved from dry storage casks.

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

8 The Petitioner's Conclusions The petitioner has concluded that NRC regulations in 1 O CFR Part 72 that govern independent storage of spent nuclear fuel in dry storage casks must be amended. PIC has concluded that thermal shock and associated degradation of spent nuclear fuel during the unloading of dry storage casks has not been adequately addressed in NRC regulations. The petitioner requests an amendment to the regulations to define the parameters of acceptable degradation of spent nuclear fuel in dry storage under 10 CFR 72.122(h). The petitioner has also concluded that NRC regulations do not adequately address issues related to the retrieval of spent nuclear fuel from dry storage casks. The petitioner requests an amendment to the regulations to define the parameters of retrievability of spent fuel from dry storage casks required under 10 CFR 72.122(1). Lastly, the petitioner has concluded that NRC regulations do not adequately address issues pertaining to unloading of spent nuclear fuel from dry storage casks. The petitioner requests an amendment to the regulations to require licensees to demonstrate the ability to unload spent nuclear fuel safely from a dry storage cask before the cask may be used at an ISFSI. Dated at Rockville, Maryland, thist~day of March, 1998. For the Nuclear Regulatory Commission. ~ef6.G__ Secretary of the Commission.

P.O. Box 174

• Lake Elmo, MN 55042
• Phone: 612-770-3861
• FAX 770-3976 August 26, 1997 DO KET UMBE IL PR 7 /). ~ q p

L. Joseph Callan Executive Director of Operations US Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Callan:

( (,3F~ J :}.O l/0) 21.J..(.J A. " DOCl(ETED OCT 1 - 1997 -~ RULEMAKINGS AND AOJU ICA"ii01,!S STAFF ~ Please find enclosed a petition pursuant to Section 2.206, Title 10 of the Code of Federal Regulations (CFR). The Prairie Island Coalition (PIC) hereby petitions the Nuctear Regulatory Commission (NRC) to suspend for cause the Northern States Power Co. (NSP), Materials License No. SNM-2506 oeeded to operate an Independent Spent Fuel Storage Installation (ISFSI) at the Prairie Island Nuclear Generating Plant (Pl). ~ 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 proceedings under 5 U.S.C. 553(e) to examine the issues addressed herein. Thank you. Sincerely, --11"~~ George Crocker, Steering Committee Prairie Island Coalition EDO -- 8970632

1' BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Docket 72-10 IN THE MATTER OF: ) ) ) ) ) ) ) ) ) ) ) ) ) PRAIRIE ISLAND COALITION, Petitioner, DOCl<ETED vs. nuu::W1KINGS AND ADJUDICATIONS STAFF UNITED STATES NUCLEAR REGULATORY COMMISSION, Respondent. \\ \\ SECY-NRC

)

Petition Pursuant to 10 CFR Part 2.206 of the Commission's regulations, the Prairie Jsland 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 these casks cannot be placed back into the pool to perform these functions due to unresolved problems with fuel degradation during storage, flash steam, thermal shock, and the resulting potential for radiation dispersion.

3.

Determine that NSP violated 10 CFR §72.122(h) by using a . cask that must be placed into the pool for necessary maintenance and/or unloading procedures, while such placement will prematurely degrade the fuel and pose operational safety problems with respect to its ultimate and necessary removal from dry cask storage.

4.

Determine that NSP violated 10 CFR §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 ISFSI is permanently decommissioned. This failure may prevent decommissioning. 6

• Determine that NSP violated 10 CTR §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 delioerately , ( and knowingly submitting incomplete and inaccurate material *, 1, 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.

10.

Administer such other sanctions for the above violations of NRC regulations as the NRC deems necessary and appropriate. Provide Petitioner the opportunity to participate in a public review of maintenance, unloading, and decommissioning processes and procedures in question and an opportunity to comment on draft findings after investigation by the NRC.

11.

Order modification of NSP's Technical Specifications to ensure a demonstrated ability to in fact safely maintain, unload, and decommission TN-40 casks.

12.

Review NSP's processes and procedures for maintenance, unloading, and decommissioning, and if NSP does not possess capability to unload casks, order NSP to build a "Hot Shop" for air unloading of casks and transfer of the fuel. ii

13.

Under 5 u.s.c. 553(e), Petitioner requests a formal . rulemaking proceeding to solicit information and review current information regarding thermal shock and corrosion inherent in dry cask storage and usage and to define the parameters of degradation acceptable under 10 CFR 72.122(h).

14.

Under 5 U.S.C. 553(e), Petitioner requests a formal rulemaking proceeding to define the parameters of retrievability required under 10 CFR 72.122(1).

15.

Under 5 u.s.c. 553(e), Petitioner requests a formal rulemaking :proceeding for amendment of current licenses and rules for prospective licensing proceedings to require demonstration of a safe cask unloading ability before a cask may be used at an ISFSI. Preliminary Matters and Facts

2.

4.

The Prairie Island Coalition (hereinafter "PIC") incorporates herein by reference the facts, argument, and conclusions of the Prairie Island Indian Community's §2.206 Petition dated May 28, 1997. PIC was established in 1990 for the purpose of loGation and dissemination of information regarding dry cask storage, and/ opposition to NSP's plans to construct and operate an Independent Spent Fuel Storage Installation (hereiijafter "ISFSin) at its Prairie Island Nuclear Generating Station (hereinafter "PI"). PIC is a coalition of 30 environmental groups, tribal and urban Indian organizations, peace and justice groups, businesses, religious groups, and urban and rural citizen organizations. It is a project of the North American Water Office. 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. At the federal level, PIC has.an active relationship with the NRC regarding PI nuclear operations. PIC filed a §2.206 petition with the NRC on June 5, 1995 regarding failure of reactor components and waste management problems, including cask unloading problems. PIC participated in the NRC public meeting in Red Wing, MN regarding NSP and Transnuclear TN-40 cask fabrication quality control problems. PIC petitioned for intervenor status in NSP's licensing proceeding before 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.

7.

w 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 irrqdiated 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 th~ 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. 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 developed, and there has been no assessment or evaluation regarding the actual ability of such measures to adequately protect worker and pubic health, and the environment. Safety margin references may also be assumed to refer to the question of whether the disintegrated fuel could be physically unloaded at all.

8.

Also in this letter, Ms. Marcus reaffirms that SARs "over-

9.

10.

. simplify matters" when they state that unloading is basically the reverse of loading, because such statements do not reflect that the unloading process introduces different conditions and complications compared to the unloading process. In a letter dated July 10, 1997, from Beth A. Wetzel of the NRC staff to NSP, Ms. Wetzel requests additional information regarding the PI spent fuel special ventilation technical specifications. Exhibit B, July 10, 1997, Letter from NRC's Beth A. Wetzel to Roger O. Anderson, Director of Licensing and Management Issues for NSP. In this request, Ms. Wetzel has clearly acknowledged the importance of the considerations which she raises, taking these concerns a step further than Ms. Marcus in her letter (Ex. A), particularly regarding concerns about steam pressurization when the cask is initially filled with radioactive pool water prior to'loading. This request raises valid questions about the ability of the pool ventilation system to adequately vent, contain, and filter radioactive material coming out of the cask as the water enters. Ms. Wetzel acknowledges the potentiAl for thermal shock, and that a cask unloading procedure which produces this effect may result in significant radioactive contamination of the environment. Degradation of the fuel and/or assemblies due to thermal shock is equally croubling. 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 storage terms. Exhibit E, April 16, 1997 NRC Memo from Jack Roe, Director, Division of Reactor Projects III/IV, Office of, Nuclear Reactor Regulation, to Cynthia D. Pederson, Director, Division of Nuclear Materials Safety, Region III. This memorandum offers "clarifications regarding the terms ready retrieval and structural defects." In this memorandum, Mr. Roe defines "ready retrieval" to mean that the regulations do not require licensees to be able to immediately retrieve waste. See 10 C.F.R. §72.122(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 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:

13.

(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 where a spent fuel cask has weld flaws, i.e., Palisades, where welds have failed. Mr. Ro~ did not address the issue or assurance* that the utility can in fact unload the casks. There are other reasons to unload a cask that have not been addressed in Mr. Roe's letter. The NRC has clearly stated that: [S]hield-lid weld failures affect the integrity of a cask confinement boundary. The root-cause of the shield-lid failures and the potential for 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. e 14 *. Another reason casks must be placed into the pool and opened is obligatory cask maintenance which must be. completed on Transnuclear's TN-40 cask. Exhibit G, NSP SAR for Prairie Island ISFSI, Table 5.1-2. Seals must be replaced, or again, there will be a helium leak and unacceptable degradation. It also does not address whether NSP can replace a seal on a.cask 20 years after it was loaded or when a seal fails. And seals do fail. Again, Mr. Roe's rationale does not address whether the cask can, in fact, be unloaded.

15.

16.

17.

Another reason the TN-40 casks at Prairie Island would require unloading is that state law requires that they be moved off of Prairie Island. This state requirement anticipates that the casks must be moved after a term of temporary storage, in Minnesota defined as eight years. In the matter of Sperit Fuel Storage Installation, 501 N.W.2d-638 (Minn. Ct. App. 1993). Even if.the spent fuel were to stay for the life of the NRC license, it would have to be unloaded to move to a federal interim site or repository, as provided in the NRC's Waste Confidence Decision and upon which all nuclear waste storage facilities are premised. September 11, 1990, Waste Confidence Decision Review, 54 CFR 39767. Again, this is another scenario where the.NRC's anticipation of the necessity of unloading is inadequate. Yet another scenario where unloading is required is for decommissioning. NRC authority rests on the requirement that it license only facilities that can be constructed, operated, and decommissioned. NRC regulations require that the facility "be designed for decommissioning," and that the licensee make provisions to "facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI... is permanently decommissioned." Because.there are unaddressed unloading issues such that it is unreasonable to assume that the TN-40 cask can indeed be unloaded, NSP has violated the rule by failing to make the required provisions that assure it can decommission the licensed facility. 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 utility demonstrate that it can unload a cask. If the . utilities can demonstrate that a cask can be unloaded in a pool after long-term storage, we can rest assured with the knowledge that, although they may not be able to unload it as soon as the need to unload appears, they will in fact be able to unload it at some reasonable point in time.

18.

No dry cask that has been used for storage for some time, i.e., over a year, has been unloaded in a pool. There are issues that remain unaddressed, and NSP has not demonstrated that it is able to unload a cask in its pool. It has no other facilities for unloading.

19.

The NRC itself declares that cladding degradation, because it could lead to future fuel handling and retrievability problems, is unacceptable. Ex. F, 4/15/97 NRC's Susan Frant ShanJanan Letter to Sierra Nuclear. In that particular case, the letter writer is concerned with degradation due to escape of helium, and emphasize~ that: Since one of the design requirements of the cask is the long-term protection of the fuel cladding (10 CFR 72.122(h), such degradation would be unacceptable. Loss of helium from the TN-40 cask is an anticipat°ed event, hence NSP's seal pressure monitors. Exhibit H, June 30, 1995, Notice of Violation, Inspection Report, 7.1, p. 23. However, the degradation that a helium leak would cause is not addressed, nor is the method by which NSP would replace the defective seal. NSP's TN-40 cask runs the significant risk of degradation due to thermal shock, loss of helium through failed seals, and most.importantly, degradation due to the passage of time. NSP's TN-40, its seal maintenance program, thermal shock inherent in placing the cask in the pool, and degradation over the passage of time make this cask unsuitable for storage. NSP is therefore in violation of 10 CFR 72.122(h).

20.

In a study of the TN-24P, which NSP claims is so very similar to the TN-40, conducted by INEL in 1990, INEL experienced serious thermal problems, not related to cladding, but to the structure of the inserted canisters. Exhibit C, INEL Letter, October 18, 1990; Exhibit D, INEL Letter, November 21, 1990. It is important to note that these were canisters containing assemblies, which allowed less room in the basket. It is equally important to note that these casks were unloaded in air in a Hot Shop. These canisters _had been stored for several years, and the thermal

• damage was so severe that the canister could not be unloaded.

In the October 18, 1990 letter, the writer . declared: [T]he canisters had "setup" in some fashion: thermally, twisting, bowing, corrosion or other... " The canisters had apparently taken on a set most probably thermally induced although possibly including other factors such as bowing, twisting or other~ The laminated makeup of the TN-24 basket may also be involved... It should be clear, nevertheless, that the experience encountered should receive future focus since the inability to extract at lest one of the assemblies with existing equipment is apparent. In the November 21, 1990, letter, in the "Review of Stuck Fuel Assembly Issue," the writer said of the damage: (T]hermal expansion of the canister is the most probable cause, bowing, twisting or other mechanisms cannot be eliminated as possibilities; we presently have little capability to determine the root cause because accessing the assembly or the basket is not feasible with fuel in the cask. For the 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 i~ no longer stuck because of thermal unloading of the basket following the removal and placement in the VSC-17 cask of 17 fuel canisters. Id. The letter noted that an attempt could be made to remove the stuck canister, but a major consideration was that it "may become stuck in a partially withdrawn position or that canister damage might be incurred." Clearly, fuel stored in the TN-24P is not retrievable.

21.

NSP's SAR for the Prairie Island ISFSI provides that the TN-40 cask seals must be replaced every 20 years, or sooner if there is a seal failure. Exhibit G. The SAR states that as a part of the cask seal replacement, the TN-40 must be placed in the spent fuel pool, and that replacement of the seals is completed in the pool. Yet, as demonstrated by Beth A. Wetzel's 7/10/97 Request, there are unresolved safety considerations recognized by the NRC, primarily ventilation of the flash steam produced by introduction of the cooler pool water into the hot cask. Exhibit B, 7/10/97, License Amendment, Request to NSP. Secondly, there remain unresolved thermal shock issues, where introduction of cooler pool water would crack zircaloy cladding or assemblies. 22.. NSP consistently claims that casks can be unloaded, _and that "thousands of Transnuclear casks have been unloaded worldwide." Exhibit I, Environmental News, August 1997. NSP has also made this statement under oath in an Affidavit, and in its legal argument. Exhibit J, In the Matter of a Request by Northern States Power Company for. Certification of Compliance, Cl-96-2189, CS-96-2190, Respondent's Response, p. 5-6; Aff. of Jon Kapitz, p.2. In Mr. Kapitz's Affidavit, he first states that: The unloading procedure and the relevant design features for the TN-40 casks approved for use at the PI' Plant are based upon features and procedures common to existing Transnuclear casks used worldwide, including shipping casks and storage casks like the TN-24P cask. Exhibit J, Aff. of Kapitz, p. 2. (emphasis added). He goes on to say that: v While NSP has not needed to unload any of the five TN-40 casks that have been loaded at the PI plant to date, a comparable Transnuclear storage cask (a TN-24P cask) has been successfully unloaded as part of a g~oject jointly sponsored by the Electric Power Research Institute and the United States Department o.f Energy. Id. (emphasis 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 even specifically cite the study!

23.

Mr. Kapitz's statements are false. He claims that the' procedures developed for Prairie Island are the same as those for the TN-24P. However, a fundamental element in NSP's unloading procedure is that it is a pool transfer. A quick review of the study provides a reason it may not have. been included with Mr. Kapitz' Affidavit. Exhibit K, EPRI, "The TN-24P PWR Spent-Fuel Storage Cask: Testing and Analyses" EPRI NP-5128, April 1987. The cask to cask transfers in this study were completed in a "Hot Shop" and were AIR transfers. These were not pool transfers as are required at Prairie Island. Hot Shop transfer procedures are inapplicable to pool transfers and do not substantiate any claim that NSP can unload a TN~40 in a pool.

24.

NSP's claims that the casks can be unloaded based upon past experience with similar casks, but this is false. NSP claims that it has based its unloading procedures on experience with similar casks, but the casks are not similar because the loading and unloading procedures are distinct. 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 NSP cite and produce that demonstrates that a TN-40 cask can be unloaded in a pool? Conclusions NSP has violated 10 CFR 72.122(f) because it cannot maintain casks. NSP has not addressed or resolved this problem and has provided inaccurate and incomplete information regarding this issue. NSP has violated 10 CFR 72.122(h) because the fuel is subject to degradation in the maintenance and unloading process specified by NSP. NSP has not addressed or resolved this problem, and has provided inaccurate and incomplete information regarding-this issue. NSF has violated 10 CFR 72.122(1) because the fuel is not retrievable, it cannot unload casks. NSF has not resolved this problem and has provided inaccurate and incomplete information regarding this issue. NSF HAS violated 10 CFR §72.130 by using the TN-40 cask and failing to make provisions that facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI is permanently decommissioned. This may prevent decommissioning in so far as a TN-40 cask that cannot be unloaded can therefore not be decommissioned. NSF has 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. NSF has received actual and constructive notice that there are cask unloading issues, has even received requests from the NRC th~t it address some issues, and rather than take steps to correct its unloading problem, it has instead refused to directly address these continuing problems. NSF has 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 through its continual insistence that_ it can unload TN-40 casks despite substantive information otherwise, and by the knowing use of inapplicable studies to back up its false claims. NSP must be held accountable for these violations. It must not be allowed to load further casks until it has demonstrated its ability to unload them before an independent third party and has modified its Technical Specifications to reflect any changes in procedures or equipment to effect this change. Further, NSP must pay a substantial penalty for its knowing submission of incomplete and inadequate information regarding cask unloading issues, particularly that it is not possible to unload a cask; that no cask used for long term storage has ever been unloaded in a pool; that because necessary cask seal maintenance requires that the cask be opened, placed into the pool and submerged, which cannot be accomplished, NSP cannot properly or adequately maintain the TN-40 casks; that introducing radioactive pool water into a hot cask can cause radioactive flash steam that poses a health and safety threat to workers and the public; that introducing radioactive pool water into a hot cask can cause thermal shock that would damage cladding and assemblies and bend or warp metals with which it comes in contact; that thermal shock would impermissibly degrade fuel and make it irretrievable; that fuel is also irretrievable because NSP cannot unload a TN-40 cask at any time in the foreseeable future; that NSP cannot decommission the casks and site 'because it cannot unload the fuel to move it to another location; for these reasons, NSP has violated NRC regulations and must be substantially fined. The NRC must prevent an erosion of public confidence in the NRC's ability to safely regulate the nuclear industry, particularly on waste management issues. The NRC must open a complete and thorough re-evaluation of dry cask storage operations at the ISFSI on Prairie Island and at the many other sites where the issues raised above remain unresolved. Until such time as this evaluation has been conducted, changes made, and problematic processes and procedures demonstrated that assure the NRC and the public of the licensee's ability to safely manage irradiated fuel in dry storage casks through the life cycle of the fuel and casks, the Materials License for ISFSI operations on Prairie Island must be suspended. During the term of suspension, no further casks shall be filled at the Prairie Island site. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20M5-0Q01 February 25, 1997 George Crocker, Steering Committee Prairie Island Coalition P.O. Box 174 Lake Elmo, MN 55042

Dear Mr. Crocker:

Exhibit A 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 (SAR) for the independent spent fuel storage installation (ISFSI) at the Prairie Island Nuclear Generating Plant provides various estimates of radiation exposure associated with tne 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 signifi~antly 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 1 ess than 0~ 27 person-rem. Regulatory limits for maximum radiation exposures to plant personnel are defined in Part, 20 of Title 10 of the Code of Federal Regulations (10 CFR 20). In general, licensees are.required to control the occupational dose to individual adults to less than five rems per year. The offsite release of radioactive materials during the unloading of a dry storage cask is expected to be negligible. In regard to the worst case scenario, the SAR for the Prairie Island ISFSI includes an analysis of a hypothetical loss of confinement barrier which assumes the total inventory of radioactive gases within a cask are released. This hypothetical scenario results in a maximum individual whole body dose of 0.15 rem for a member of the public. Any credible accident involving a dry storage cask at Prairie Island would result in less exposure to the general public than does this hypothetical scenario. The possible generation of steam during the refilling of a storage cask would not be a significant factor in offsite release since the steam would be vented into the spent fuel pool. In addition, the loading -and unloading of casks are performed within the auxiliary building which has additional design features that minimize the release of radioactive materials. As part of its assessments of licensees' procedures for unloading dry storage casks, the NRC staff considers the dry-run exercises performed to verify key aspects of unloading procedures, as well as licensees' actual experience in the loading and unloading of transportation casks, loading of storage casks, handling of spent fuel assemblies under various conditions, and performing various activities associated with reactor facilities. In the absence of actual experience in unloading spent fuel from a cask following a long period

of storage. a general understanding of technical capabilities and related experiences enables the NRC staff to assess the adequacy of a licensee's procedures for unloading dry storage casks. For those examples of cask unloadings mentioned in the staff's letter of January 7, 1997, to Representative Jennings, the activities were performed without significant releases of radioactive material and within regulatory limits pertaining tij occupational exposures of plant personnel. In order to ensure that the fuel assemblies in dry storage casks have maintained their integrity during storage, a gas sample is taken from the (-.:- _. early in the unloading process. In the case of Prairie Island, the licens~~:1 unloading procedure (Enclosure 1) requires personnel to determine if additional steps or precautions are warranted based on the analysis of the g"': 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 have not involved the temperature differences between fuel and coolant thqt may occur t if a cask was unloaded after a period of storage, engineering evaluations and_, i experiences with transportation casks have shown that "thermal 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 normal fuel handling practices. In the unlikely event that fuel degradation has occurreQ during storage, the unloading may require additional filtering and even vacuuming debris from the bottom of the cask. Such steps would be developed and implemented, as necessary, following the discovery of fuel damage as a result of samples and surveys required in the unloading procedure. Licensees o have experience in handling damaged fuel assemblies, including the need tr retrieve fuel pellets, as a result of several cases of fuel assembly damage that occurred during reactor operation. Although licensees would be able tr develop*means to retrieve degraded fuel assemblies from a dry storage cask, the accumulated occupational dose to perform this activity may be increased from the previously mentioned estimates. Fuel reactivity for criticalitv . considerations could increase only under very idealistic and highly unlike1, disintegration patterns in the fuel. Upon detection that fuel disintegrati~L had occurred, special measures would be developed and implemented to assure~ adequate safety margin is maintained during unloading. Some SARs do state that unloading is basically the reverse of loading and H'

• statement, in a general sense, is true. However, such statements may tend{,

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 pl an for dry cask storage and related statement:-:~ made by the NRC staff, including those by Mr. Kugler, the staff was emphasizing that licensees need to identify the conditions and complicati:~

G. Crocker . that are associated with the unloading process and ensure that unloading

• procedures address those concerns.

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

Enclosures:

As stated (2) cc w/encl: The Honorable Loren G. Jennings Minnesota House of Representatives Box 27 Rush City, MN 55069 cc w/o encl: see ~ttached page

UNITED STATES NUCLEAR REGULATORY COMMISSION Exhibit B WAIHINGTON, D;C. IDCC: 0001 Mr. Roger O. Anderson, Director Ucnsing and Management laauea Northern States Power Company 414 Nicollet Mall Minneapolis, Minnasota 55401 July 10, 1997

SUBJECT:

REC\\UEST FOR ADDITIONAL INFORMATION ON THE PRAIRIE* ISLAND NUCLEAR GENERATING PLANT, UNITS 1 ANO 2, AMENDMENT OF SPENT FUEL POOL SPECIAL VENTILATION TECHNICAL SPECIFICATION CTAC NOS. M987!52 AND M987153)

Dear Mr. Anderson:

By Jettera dated May 7, 1997, and supP'ementad May 30, 1997, Northern States Power Campany aubmitted a request to amend th. Prairie Island Technical Specifications pertaining to the spent fuet pool special ventilation system. In order to review the proposed changes the staff requires some additional information. Our request for additional information (RAU Is enclosed. In order.to continue our review of your submittal on an expedited basis, please provide your response to the staff's RAI as soon as practical. If you have any

• questions regardi'lg the c:nntant of the RAf, please contact me at (301) 41fi-1355.

¥ Docket Noa. !50-282. 50*306

Enclosure:

As stated cc w/encl: See next page Sincerely, &a~ 0. ~ Beth A. Wetzel, Project Manager Project Oireatorata 111-1 Civision of Reactor Projects

• Ill/IV Office of Nuclear Reactor Regulation

REQUEST FOR ADDITIONAL INFORMATION FOR REVIEW OF THE AMENDMENT OF THE SPENT FUEL POOL SPECIAL VENTILATION ZONE TECHNICAL SPECIFICATIONS

1.

Step 8.27 of 095.2, "TN-40 Cask Unloading Procedure* directs the cask to be filled with water *. The caution prior to step 8.27 reads, *The water/steam mixture from the vent port hose may contain some radioactive gas. The area directly above where the hose is discharging shall be closely monitored to determine tf there is a radiological hazard.* Is the spent fuel pool pecial ventilation system operable during the performance of this 1tep of the unloading procedure? If the spent fuel pool special ventilation avstem is inoperable during this step and other portions of the unloading procedure because the overhead crane 41 supporting th* cask through the open spent fuel pool enclosure slot doors, discuss why an Inoperable ventilation system does not pose a radlologlcal hazard and give any . precautions and protections that ensure that 10 CFR Part 20 and Part 100 requirements ere not exceeded.

2.

Section 5.5 of the Prairie Island ISFSI (independent spent fuel storage lnstallatlon] safety analysis report (SAR) states in part,

• After moving the cask into the fuel pool area, -the cavity will be depressurlzed and..the cask lowered Into the spent fuel pool.* However, Step 8.4 of procl;jdure 095.:

directs the cask to be depressurized while it Is still located In the rail bay area. 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 depressurizatlon? Does the SAR assume that tha spent fuel poof special ventilation system will be operable during the cask depressurization evolution 7

3.

When th* spent fuel cask Is filled with water prior to unloading the fuel (per Step 8.27 of 095.2, "TN-40 Cask Unloading Procedure*), discuss the likelihood that this will result in cracking of the spent fuel rods due to the Interaction of the cool spent fuel pool water with the hot fuel elements. If any fuel cracking la predicted, list the expected radlonuclides and quantities that will be released into the cask and into the fuel building when the cask i 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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MEMORANDUM TO: FROM: UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D,C_ Z0:1&5-4004 April 16, 1997 Cynthia D. Pederson, Director Oivision of Nuclear Materials Safety Region I II Jack w. Roe O rector~ Division o( eactor Projects 111/IV Office of Nuclear Reactor Regulation Exhibit E JJ

SUBJECT:

TASK INTERFACE AGREEMENT 96-0440; DEFINING DRY CASK STOPA~~ TERMS (TAC NOS. M97346 AND M97347) In response to your request dated November 26, 1996, NRR/DRPW and NMSS/SFPC have discussed the questicns raised and offer: the following clarifications regarding the ~erms ready r~t~1eva1 and structural defects. The two basic reasons to return a c~~k to the spent fuel pool and unload the spent fuel assemblies are either to (1) retr1eve the fuel assemblies for further processing or di!no~al, or {2) respond to an event or condttion that has potent1ally degraded the design requirements established for the cask. The staff has not identified otectfv~, mea to be taken Wl ---........,;~ site con u'i,1css O an ace ri a1 frc,m. as age cask. In regard to the ~ sk des1 ns must s..eent fue 1 for further pi.. ~ces sing or d 1 sposa e consistently taken a po~l'~!(in that licensees can satisfy this requirement without maintaining the capability to retrieve the spent fuel rrom a cask wiffiin asp ,ea perio~ or t,me, and maY, if necessary, develop altefnate d'pt1uns fbr fuel retrieval if a cask u oa n ed d; t~ ~ shortage of soa~, as en e ns ere aep a le because 11cen~~es have a grea eal of flexibility in their ability to schedule and plan for th~ transfer of spent fuel from a storage cask to another cask for storage !:r ~hipment. . Several of the actions requ'fr9d by ISFSI technical spec1f1cat1ons or cask certificates of compliance sp~cify that, 1n the case of certain events or condition.s, a cask. may need to be* unloaded, or otherwise returned to a safe storage condition. The N~C staff has stated that the potential need to unload a cask in response to an evant or condition 1n the technical specifications or certificates of compliance does not require licensees to maintain a cont1nu.ous ability to unload a cask w1th1n a specified time. This pos1tion 1s based on

• lthe absence of an identified event or condition involving the storage casks l)that would result in an iamediate threat to public health and safety. The position is reflected in past NRC decisions such as the acceptability of (1) licensees not having to maintain space in spent fuel pools to accornnodate CONTACT:

W1111am Reckley, NRR (301) 415-1314 APR 2 l fSl.

I unloading of a cask, and (2) several licensees sharing a single cask transport vehicle between different reactor sites. In the specific case of the Prairie Island ISFSI, the NRC staff, in its Safety Evaluation Report dated July 1993, stated that its review of the accident analyses determined that, "Dose equivalent consequences, from a single cask, to any individual. from direct or -indirect radiation.and gaseous activity release after postulated accident events, are less than 'the 50 msv (5 rem) limit established in 10 CFR 72.l06(b)." Additionally, in tts Environmental Ass*ssment, dated July 28, 1992, the staff assessed the accident dose at the site boundary as, "... a small fraction **. of the criteria specified ***. " and found that. "The doses are also much less than the Protective Action Guides established by the Environmental Protection Agency (EPA) for 1nd1v1dua1s exposed to rad1at1on as a result of accidents." Because 1t has been shown that the dose equivalent to any individual from postulated accidents involving a single cask 1s below levels required for taking protective actions to protect public health, the NRC staff considers that a t;me-urgent unloading of a TN-40 cask is a h; gh l y un 11 ke 1 y event. However, fo 11 owing certain events or conditions, the licensee is required to take corrective actions to ensure safe storage conditions and to perform inspections to ensure a cask continues to meet applicable design requirements. This may include returning a*cask to the Auxiliary Building and/or the spent fuel pool. However, once the cask is in the spent fuel pool, it* does !lilt have to be unloaded immediately to maintQ,in safe storage conditions.* The licensee would have time to consider available options, required precautions, and other special considerations that may be involved in the required unloading of a cask. The storage methods for spent fuel must protect against degradation of fuel assemblies or casks that would create operational safety problems during unloading. Operational safety problems are those that involve gross rupture of the fuel cladding such that significant quantities of fuel material and fission products are released to the storage environments. The design requirement to maintain fuel cladding integrity dur1ng storage leads to restrictions on the fuel assemblies that can be 1n1tially loaded into the casks. Acceptance criteria for fuel assemblies to be stored pertain to heat. generation rates, initial enrichments, assembly geometry, and other characteristics that establish boundary conditions for the analysis of fuel assembly performance during normal storage and potential off-normal conditions. The wording of Prairie Island ISFSI Technical Specification 3.1.1.(6) and the safety analysis report should be interpreted in light of the regulatory background set forth 1n thfs paragraph. 1 In addition, a 1TS 3.i.1.(6)- Fuel assemblies known or suspected to have *structural defects or gross cladding failures (other than pinhole leaks) sufficiently severe to adversely affect fuel handling and transfer capability shall not be loaded 1nto 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.

definition for "gross cladding defect" has been incorporated into NUREG-1536, "Standard Review Plan for Ory Cask Storage Systems," which was recently issued in final form. In the specff1c case of Prairie Island, neither 10 CFR 72.122(1) or specific ISFSI technical specifications introduce additional requirements for the fuel handling equipment used to actually load or unload the fuel assemblies into the cask since such matters are addressed under existing 10 CFR Part SO regulations and licenses. The structural requirements deffned by the ISFSI SAR and technical spec1f1cat1on *are satisfied even 1f 1t 1s necessary to use a special handling tool to overcome problems in lifting selected fuel assemblies, provided that these* assemblies do not have gross cladding failures and will otherwise maintain fuel assembly geometries assumed in the design-basis analyses parfonaed for the cask. The adequacy of the licensee's actions should be judged tn the context of the regulat,ons in 10 CFR Part 50 and the issoc1ated reactor facility operating license. If the 11.censee*s actions are easonable for the handling of fuel w1th*1n the spent fuel pool, those same ctians can be credited in the determination of whether the licensee satisfies the structural integrity requirements of the ISFSI technical spec1f1cation and fuel retr1evab11ity requirement of 10 CFR 72.122(1). If, on the other hand, the licensee's corrective actions are deemed inadequate or the spec1al fuel handling procedure increases the probab111ty of a fuel handling accident within the reactor facility, actions or 1nqu1r1es from the NRC staff should be presented 1n the context of regulations such as Appendix B to 10 CFR.50 or 10 CFR 50.59. The NRC Office of the General Counsel has reviewed this response and has no legal object1ons. Please contact William Reckley of rny staff at {301) 415-1314 if you have,any additional questions or concerns regarding this matter. (w/1ncom1ng): c. Hehl, ~I. B. Mallett, RII R. Scarano, RIV 2 Prairie Island ISFSI Technical Specification 1.3.2, "Fuel and Cask Handling Activities," states: Fuel and cask movement and handling activ1t1es which are to be performed in the Pra1r1e 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 DPR-60 and associated tRr.hnir.al 5p@c1f1cations..

. * ~RC Inspection Report - Sierra Nuclear Corp. Selected Reports Index I News and Infonnation I NRC Home Page I E-mail Mr. Art J. McSherry President Sierra Nuclear Corporation One Victor Square Scotts Valley, CA 95066 April IS, 1997

SUBJECT:

NRC INSPECTION REPORT NO. 72-1007/97-204 AND NOTICE OF NON CONFORMANCE

Dear Mr. McSherry:

s letter refers to the inspection conducted March 17-21, 1997, at your facility in Scotts Valley, California, and at two of your fabrication contractors' facilities: March Metalfab, Inc., in Hayward, California~ and Nor-Cal Metal Fabricators, in Oakland, California. The team examined infonnation about~seal weld failures on dry spent fuel storage casks at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. Additionally, the team assessed the adequacy of your corrective actions taken for the findings identified in. Inspection Reports 72-1007(96-204 and 96-208, regarding the Model VSC-24 dry spent fuel storage system manufactured under Certificate of Compliance No. 72-1007. The enclosed report (Enclos.:U-e 1) presents the results of our inspection. The team held an exit meeting with you in the Sierra Nuclear Corporation offices on March 21, 1997. During the inspection, the team found that you failed to meet certain Nuclear Regulatory Commission requirements. The team identified four nonconfonnances regarding failures to perfonn work in accordance your Quality Assurance Program. The nonconformances were failures to (I) examine the potential ric aspects of the shield-lid weld failures at ANO and Palisades, (2) submit a change to the Certificate of mpliance to correct a-nonconservative requirement for the drain-down time limit for a loaded cask, (3) submit a Safety Analysis Report change to correct the 1986 American Society of Mechanical Engineers Code omissiqn of nondestructive examination requirements for temporary attachments, and ( 4) control measuring ~t equipment. I Two of the nonconformances raise safety concerns. First, the shield-lid weld failures affect the integrity of a cask confinement boundary. The root-cause of the shield-lid failures and the potential for delayed cracking on loaded casks must be understood. Although the failure of both the cask's 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~ of the design requirements of the cask is the long-tenn protection of the fuel cladding [10 CFR 122(h)], such degradation would be unacceptable. Second, the nonconservative Technical Specification for cask drain-down time affects the margin to criticality.

t'a el Sierra Nuclear Corporation's lack of timely and comprehensive action, in dealing with these important safety issues, is a significant regulatory concern. As the certificate holder, Sierra Nuclear Corporation is responsible for the adequacy of the design of its fuel storage casks. We expect Sierra Nuclear to take a central role in resolving each technical problem associated with your cask design. We have arranged a meeting with you on May 6, 1997, to discuss this matter further. This meeting is open for public observation. At the meeting you should be prepared to discuss your short term and longer term corrective actions to address the issues and concerns raised by our inspection. Please provide us, within 30 days from the date of this letter, a written statement in accordance with the instructions specified in the attached Notice ofNonconformance (Enclosure 2). We will consider extending the response time if you can show good cause for us to do so. In accordance with IO CFR 2. 790 of the NRC's "Rules of Practice,". a copy of this letter, its enclosures, and your response will be placed in the NRC Public Document Room (PDR). Sincerely, /signed/ Frant Shankman, Chief ransportation Safety and Inspection Branch Spent Fuel Project Office, NMSS

Enclosures:

1. Inspection Report 72-1007/97-204

2. Notice ofNonconformance Docket No. 72-1007

• Confirmatory Action Letter - Arkansas Nuclear Selected Rel)orts Index I News and Information I NRC Home Page I E-mail CAL No. 97-7-002 Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operations, Inc.

1448 S. R. 333 Russellville, AR 7280 I May 16, 1997

SUBJECT:

CONFIRMATORY ACTION LETTER Mr. Hutchinson: During the week of March 17, 1997, U.S. Nuclear Regulatory Commission staff inspected Sierra Nuclear Corporation (SNC) and two of its fabrication contractor facilities. SNC holds Certificate of Compliance No. I 007 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 joming the cask shield lid to the multi-~sembly stQrage basket (MSB). The Palisades welding difficulty occurred in March 1995 and the first instance at ANO in December 1996. After the recent SNC inspection, problems arose while welding another ANO cask on March 26, 1997. NRC is concerned about the difficulties . ~nt~ed with the welds joining the shield lid to the MSB, since this weld __ i_s part of the confinement boundary ofttie VSC-'.'2-~fFurthemicire:*the*weid between-the MSB and the structuiaflid inay be susceptible to the same r~re mechanisms* as the shield lid wdd~ It is possi?le that_ t&StJi?~icular w~I~ problems may not develo until after cask welds have under one non-de rue 1ve ex.,,.....,,....,,., on. Although such weld failures cl not pose an off-site threat to pu lie health and safety, such an occurrence would cause the loss of the / __ ~ ~ __ a~~sphere inside the MSB. This_c_gndition coul_g_resuli~!!t fue!.fd~dding __ (J~~~tion an~ future fuel .1 handling and retrievabilit£problenis. - v ----=-------"c...J--- The March 1997 inspection revealed that neither SNC nor the user licensees had performed a comprehensive root-cause analysis of the first two weld problems. An understanding of the root cause is essential to preventing recurrence when welding future casks, and to assessing the possibility of additional weld problems, perhaps undetected or delayed, in loaded casks. On May 6, 1997, NRC held a public meeting with SNC representatives to discuss SNC's implemented and planned actions in response to the weld problems and inspection findings. Representatives of your staff attended this meeting. As stated at this meeting, the staff remains concerned that the root cause(s) of the weld problems have not been conclusively determined. Pursuant to a May 14, 1997, telephone conversation between Randy Edington and Charles Haughney, Deputy Director of the Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, it is our understanding that you will take the following actions before loading additional VSC-24 casks with spent nuclear fuel:

(1) Detenninethat your welding and*lnspection practices provide reasonable assu.:ance that cracking. includin ossiblc unde1e:.;ted or,dela-,;ed cracktp~ ;*Nil!.nut,.':iJOt,iJ! ID (he*.welds.~,hng the shield lid and structural lid to* e B. _If: necessary,,. mndify younwe.ldurg.,u1r1cesses ttJ i.nhib.it

• recurrence of these welding problems.

(2) On completion,of this act..iMi. *anil m Jw--:,t;.H ~y:r-.'~J?,-i:rlt 4:-~fu~ *,\\~Tfi'At~e* *;;:1:3c~2.g. -cask with spent fuel. you will submit to t~1e Ojrectot,,Qffit.-re,*,if't~1;:~ ;~"t-;a'f.i.uii !Sia~)' >::md *~e._w,tds,, "l!. vmtten description of any procedural or design.modificaui:m.,s:i!nade*~iib"~,1,w-J :~.'.9, *i.'t.1;,,m,; J_ T~ :wi-w.,~ :f1.lould include the technical justification ~r each 'ti10dimat~11:,1. *"~.,;tJ;~i?{!ii ::i.h~ *5:~farci.:~~ * ~t1f,~ :t:el'it *to William F. Kane, Director, Spent Fuel Proje.ct Ot:'fu-~t\\.aro:Ho J(!)OO -j;~~nril Arur..inrm:1r-&-L -\\~c*11,s ~-.include in this response the information required by,irern 2 :bfww,,r.orwrn.rit(!J~ :,af.l-~ ;a~;:;t'i;,i~Y; *rtqo1n1:'i 1'flY ~teen 1 above.) Pursuant to Section 182 of the.Atomic Energy.Act, 42 U:S. 'C '.?2~2, ;tlr~ mv~ required to: (I) Notify me immediately if your understanding differs from tmt ~- f~--ih above; (2) Notify me in writing when you have completed the actions addres.std in this Confirmatory Action Letter. Issuance of this Confirmatory Action Letter does nqt preclud~ issu.m-ce of an order formalizing the above commitments or requiring other licensee actions; nor does it preclude NRC from taking enforcement action "olations ofNRC requiremems that may have prompted the issuance of this letter. In addition, failure to 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, if you find it necessary to include such information, you should. clearly indicate the specific infonnation that youdesire not to be placed in the PDR, and provide the legal basis to support. your request for withholding the information from the public. Sincerely, Malcolm R Knapp, Deputy Director of Nuclear Material Safety afeguards Dockets 72-1007, 72-13, 50-313, 50-368

ISFSI SAR TABLE 5.1-2 ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS Operation No. of Time Avg. Distance Personnel (min) (ft) from Cask Receiving

1.

Unloading (Al)

2.

Inspection (A2 through A7)

3.

Transfer to cask loading pool (AS) Cask Loading Pool

4.

Lower cask int:o pool (Bl) Load fuel (B2 through B4) 5 Place lid on cask (BS) 5 Lift: cask to pool *urface (B6) 5 30 5

8.

Install lid bolts (B6) 5 120 3

9.

Drain cavity (B7 through Bll) 5 90 6

10. Transfers to decontamination area (Bl2) 3 60 10 Decontamination Area

11.

Decontaminate cask (Cl, C2) 3 120 3

12.

Remove vent: plugs 2 30 5

13.

Drying, evacuating, backfilling (C3 through Cl3) 2 480 5

14.

Install top neutron shield Cl4) 2

15.

3 15

• Install pressure

6.

transducers (Cl5 through Cl7).. 2 30 5 Pressurize int:erspace (ClS)

17.

Check leakage (Cl9) 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.

Load on transport vehicle (C23) 3 60 5

22.

Transfer to storage area C24) 3 60 10 TABLE 5.1-2 REV. 2 9/91

Ql!eration ISFSI SAR TABLE S:1-2 (Continued) ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS No. of Time Avg. Distance Personnel (min) (ft) from Cask Storage Area

23.

Unload from vehicle position in location (Dl, D2, D3)

24.

Check surface dose rate (D6)

25.

Connect pressure

• instrumentation (D4, DS)

Periodic Maintenance -3.

4.

Visual surveillance (NA) Repair surface defects (NA) Instrument testing and calibration (NA) Instrument repair (NA) ___.......;----Maintenance

• L !ii&!W 5

5 5 2 2 2 2 3 60 30 30 180 60 1950'1\\-k 5 3 5 5 3 5 3 8 No measurable dose associated with this activity. Therefore, the number of personnel, time and distance are not significant. '1\\-k Parenthetical information corresponds to Table 5.1-1 activity numbers. Total time to transfer cask t return cask to ISFSI pad. TABLE 5.1-2 replace lid seals, and REV. 2 9/91

I Operation ISFSI SAR TABLE 5.1-2 (Continued) ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS No. of Time Avg. Distance Personnel (min) (ft) from Cask Storage Area

23.

Unload from vehicle position in location (Dl, D2, D3) 5 60 5

24.

Check surface dose rate (D6) 5 30 3

25.

Connect pressure instrumentation (D4, DS) 5 30 s Periodic Maintenance ~

1.

Visual surveillance (NA) 2 15 5

2.

Repair surface defects (NA) 2 60 3

3.

Instrument testing and calibration (NA) 2 180 5

4.

Instrument repair (NA) 2 60 3 __.,.,... H1j or Ma;intenmce \\ (once in 20 years) l' Replace cask lid sea.ls 3 1950-A-it 8 No measurable dose associated with this activity. Therefore, the number of personnel, time and distance are not significant. Parenthetical information corresponds to Table 5.1-1 activity numbers. Total time to transfer cask to spent fuel pool, replace lid seals, and return cask to ISFSI pad. TABLE 5.1-2 REV. 2 9/91

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION Ill Mr. E. Watzl, Vice President Nuclear Generation Northern States Power Company 414 Nicollet Hall Minneapolis, MN 55401

Dear Mr. Watzl:

801 WARRENVILLE ROAD LISLE, IWNOIS 60632-4351 June 30, 1995 This refers to the special NRC inspection from January 24 through May 11, 1995, of dry cask storage activities at the Prairie Island site. This inspection was conducted by the resident inspectors, selected RIII based inspectors, and technical staff from the Office of Nuclear Reactor Regulation and. the Office of Nucledr Materials Safety and Safeguards. The purpose of his 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 nreooerational testina activities. We discussed the results of this inspection with you and other members of your staff at a public exit meeting on April 28, 1995. At that meeting we identified five items that required further resolution. You provided us with additional information for each of these items and we completed our review of

• the subject items during the next two weeks.

On May 11, the NRC issued a

• schedular exemption from the requirements of 10 CFR Part 72.82(e) allowing you to submit the results of your preoperational test less than 30 days before the receipt of fuel at your onsite Independent Spent Fuel Storage Installation.

On May *12 you loaded the first cask with spent fuel

• The enclosed copy of our inspection report identifies areas examined during he inspection. Within these areas, the inspection consisted of a selective xamination of procedures and representative records, observations, and
• interviews with personne 1
• 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 of Vfolation (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 inspection. Therefore, no response with respect to this violation is required. However, we are disappointed that NRC inspectors, rather than your own staff, identified these procedural deficiencies.

E. Watzl We also identified several 'da'.kne~s~li ;wkt'tt :ynur rJiJtir.:~}f} :~v.,,rumce relative to dry cask storage acthMt:i-e:i~

• Tbe-:;e :,-am,kMJnr.in c<1in:*1'-tidtI;i-

f) poor oversight of vendor activities until 1fil.te *tn t.~ drJ* -t:~'?tk st.or.age !pl~~t;' 2) lack of effective engineering i nvolveme.,-t *lJt "Ver~ *'.fahrt~st~m,,'M-'!1v1 ti es; 3) the i nef feet i veness of your qualt ty :1:s.1.GJ"'aK.s rt>r,'1J,ilr.b::a~rmt,~ :liis~;.'iS i ng vendor performance during the cask 'fabnc1rUft 1~ess,:; -4) 'the WeM.-1! of a comprehensive plan fot* i-nspectingv a.ooittr~,. ~

m.,t-un1r,g dry cask storage activities* onsite, particularly those :acti'dUes associated *ilith the 10 CFR Part 50 1 icense; and 5) overall poor planning for dry as.k 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 this req~est within 30 days of the date of this inspection report. We will continue to evaluate the effectiveness of your rrective actions to improve your performance in dry cask activities during ure NRC inspections. In accordance with 10 CFR 2.790 of the NRC's *Rules of Pra*ctice,* a copy of this letter, the enclosure, and your response to this*letter will be -placed in the NRC Public Document Room. / The response requested by this letter is not subject to the clearance, procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. We will gladly discuss any questions you have concerning this inspection

• Docket No. 50-282 Docket No. 50-306 Docket No. 72-10

Enclosures:

1. Notice of Violation

. Sincerely, c~AJ-~- Edward G. Greenman Senior Oversight Manager Region III Dry Cask Activities

2. Inspection Report No. 50-282/95002; 50-306/95002; 72-10/95002(DRP)

See Attached Distribution

NOTICE OF 'HOLATlOM Northern States Power Company Prairie Island Nuclear Plant Dockets No. '50--.?.a-2; 50-306; 72-10 licenses No. OPR-42; OPR-60; SNM-2506 During an NRC inspection conducted from January 24 through May 11, 1995, a. violation of NRC requirements was identified. In accordance with the *General Statement of Pol icy and Procedures for NR*c Enforcement Actions,* 10 CFR Part 2, Appendix C, the violation is listed below: 10 CFR Part 72.142(b) requires a licensee to establish, maintain, and execute a quality assurance (QA) program with regard to an Independent Spent Fuel Storage Installation (ISFSI) that satisfies each of the applicable criteria-of Subpart G, *Quality Assurance.* In meeting the Part 72.142(b) requirement, 10 CFR Part 72.142(d) accepts a Comission-approved quality assurance program

• which satisfies the applicable criteria of Appendix B to 10 CFR Part 50.

As such, the ISFSI Safety Analysis Report states that the (!reviously approved Northern States Power gA program which satisfies applicable criteria of 10 CFR Part 50, Appendix B, will be applied to activities, structures, systems, and components of the ISFSI commensurate wlth the1r 1mportance_ to safety. Criterion V of Appendix B to 10 CFR Part 50 requires that activities *affecting quality be prescribed by documented instructions, procedures, or draw1ngs, of a type appropriate to the circumstances and that these activities be* accomplished in accordance with the associated instructions, procedures, or drawings. Cask handling, loading, and unloading are activities affecting quality. ~ 2 *.

• es were Surveillance Procedure, SP 1077, *special lift Fixture for the TN-40 Cask, 1 did not address dimensional checks of the special lifting device, as required.

Surveillance Procedure, SP 1075, *TN-40 Fuel Selection and Identification,* did not incorporate the requirement of Technical Specification (TS) 4.1.2, which states that.*before inserting a spent fuel assembly into a cask ***, the identitf of each fuel assembly shall be independently verified and documented.

3.

Procedure 095.1, *TN-40 Cask Loading Procedure,* specified in the prerequisites section that SP 1077 be performeOlt)iays prior to loading a cask. However, the TS 4.19-requirement to pe~rm a vis

• on of the lifting device *(lift beam and extension) a r1 y operability of the device 7 days prior to use, was not identifie

/ also was no procedure identifying actions required to verify operability of the lifting device.

. Notice of Violation 4. Procedure 095. 1, Procedul'e.~ 11 to erform radi at cas sur a , as req

5.

Procedure 095. 2 uately ron

6.

Procedure 095.2, *TN-40 Cask Unloading Procedure* did not contain a hold point to ensure work would not continue until the results of the inner cask volume sample had been reviewed. lhis procedural hold po1nt is impoftanc co ensure that an Unplanned and unmonitored release path ts* not created while the cask is in the Auxiliary Building.

7.

The licensee did not have a procedure for conducting 10 CFR Part 72.48 safety evaluations. I; is is a Severity Level IV Violation (Supplement I) (50-282/95002-0l; 306/95002-01; 72-10/95002-0l(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

While the inspectors recognized that finalizing the loading and unloading procedures was contingent upon completion of the dry run and the subsequent incorporation of any lessons learned, there were many as have been in the dry r 10n requ1reme graph 3.2). In a and effectively incorporated into the loading and unloading procedures I l~~~t;tt~ittittt~tt~~~~ttt:1~\\\\~ltit~tlli~~~~\\t~'it~ 1 ~n

• The licensee did not take a disciplined approach to inspecting the fuel designated for cask storage as evidenced by weaknesses identified by the inspectors during observation of fuel inspection activities (paragraph 7.3).

Some weaknesses were noted with the licensee's docwnented basis for safety evaluation conclusions (paragraph 8.2). 3

operational checks of vehicle brakes., lifttng equipment, t'-lrnt,aiiles, jacks, and cask links. 3.1.5 Surveillance Procedure, SP 1075. *rN-40 Fuel_S,tlectton...Jn.4 Identification* I The inspectors reviewed SP 1075 and the cas*k loading iprl)tedun'.., D95.1, to verify that selected Technical Specification (TS) requirements had been incorporated into procedures. Surveillance requirements for 12nsuring that fuel assemblies which satisfy the criteria of TS 3.1.l 'Would be loaded into ~he cask, are defined in TS 4ol. TS 3.1.1(6) required that, *fuel assemblies known or suspected to have structural defects or gross cladding failures (other than pinhole leaks) suff;ciently severe to adversely affect fuel handling and transfer capability 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 uipment to perform the fuel inspection. The inspectors considered t~~s a. eferable method for identifying fuel anomalies and ensuring compliance-*. with 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 pro~edure 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 quirements of TS 4.1.2 into SP 1075 is considered an example of a violation Criterion V of Appendix B to 10 CFR Part 50 (50-282/95002-01; ~~306/95002-01; 72-10/95002-0l{DRP)). 3.2 Loading and YoJoadin; Procedures The inspectors reviewed the loading (D95.1) and unloading (D95.2) procedures for technical adequacy and to detennine if the lessons learned from the preoperational testing/dry run had been appropriately incorporated into the procedures. 3.2.1 D95.I, *TN-40 Cask Loadjnq Procedure* l The original D95.1 procedure specified in the prerequisites section that SP 1077 be performed 30 days prior to loading a cask. However, the Technical Specification (TS) 4.19 requirement to perform~* visual inspection of the

Emile: UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 22, 1997 Please log in this petition for rulemaking from the Prairie Island Coalition. This petition*was docketed as PRM-72-4 on October 16, 1997 -- - ~-----------

NOTE TO: FROM:

SUBJECT:

Emile Julian Chief, Docketing and Services Branch Carol Gallagher ADM,DAS June 12, 1998 DOCKETING OF COMMENT ON PRM-72-4 Attached for docketing is a comment letter related to the PRM-72-4. This comment was received via the interactive rulemaking website on June 11, 1998. The submitter's name is Carol A. Overland, 402 Washington Ave. So., Northfield, MN 55057. Please send a copy of the docketed comment to David Meyer (mail stop T6-D-59) and Stan Turel (mail stop T9-F-31) for their records.

Attachment:

As stated cc w/o attachment: S. Turel

From: To: Date:

Subject:

John Hoyle WND1.WNP2(ELJ), CJS 5/28/98 8:50am SAPL COMMENT ON SEABROOK AMENDMENT -Forwarded -Forwarded -Reply OK, Jeanne. Pls send us a copy of the letter in the next acknowledgement batch. > > > Jeanne Shoemaker 05/28/98 08:34am > > > FYI -- David Meyer also left a voice message for me saying that they do not acknowledge comments they receive. He thinks it would be appropriate for you to send an acknowledgment saying that you received the letter and it has been forwarded to appropriate NRC staff for consideration (David would appreciate receiving a copy of your acknowledgment to keep with the comments). When they receive the comment letter from SAPL, they will log it into their comment tracking system and forward it to the staff for review Gust like RAS does with mlemaking comments). CC: TWFN_DO.twf3_po(DFM),

From: To: Date:

Subject:

Jeanne Shoemaker JCH,ELJ 5/28/98 8:34am SAPL COMMENT ON SEABROOK AMENDMENT -Forwarded -Forwarded FYI -- David Meyer also left a voice message for me saying that they do not acknowledge comments they receive. He thinks it would be appropriate for you to send an acknowledgment saying that you received the letter and it has been forwarded to appropriate NRC staff for consideration (David would appi-eciate receiving a copy of your acknowledgment to keep with the comments). When they receive the comment letter from SAPL, they will log it into their comment tracking system and forward it to the staff for review Gust like RAS does with rulemaking comments). CC: DFM _J

From: To: Date:

Subject:

Mike Lesar: David Meyer MTL 5/28/98 8:17am SAPL COMMENT ON SEABROOK AMENDMENT -Forwarded I called Jeanne S. and left a message requesting SECY send a letter stating they have received the comment and have forwarded the comment to the appropriate staff for review. SECY will include us on distribution for their response. David Meyer CC: WND1.WNP2.CJS

From: To: Date:

Subject:

Jeanne Shoemaker TWD2.TWP6.DLM1, TWD2.TWP6.MTL 5/27/98 5:20pm SAPL COMMENT ON SEABROOK AMENDMENT You will be receiving (through the EDO's office) a May 22, 1998, letter from Steve Haberman of the Seacoast Antipollution League commenting on the Seabrook amendment (63FR19972-74, April 22, 1998), which was addressed to Mr. Hoyle by mistake. The last line of the letter says "... we look forward to a response to our concerns." Since the letter is addressed to Mr. Hoyle, he would like to make sure that they at least get an acknowledgment. Will an acknowledgment go out from your office? If not, he may send one saying that the letter has been forwarded to your office for consideration by the staff. Thanks, Jeanne CC: JCH}}