ML20148P119

Notice of Receipt of Petition for Director'S Decision Under 10CFR2.206,requesting That NRC Determine If NSP Violated Requirements of 10CFR72.122(1) Using Matls License SNM-2506 for Isfsi.Director Denied Petitioner Request for Action

ML20148P119
Person / Time
Site:	Prairie Island
Issue date:	06/27/1997
From:	Miraglia F NRC (Affiliation Not Assigned)
To:
Shared Package
ML20148P115	List: ML20148P109 ML20148P119
References
2.206, NUDOCS 9707010240
Download: ML20148P119 (2)

Text

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7590-01-P l*'

a U.S. NUCLEAR REGULATORY COPMISSION 4 DOCKET NOS. 50-282, 50-306, and 72-10 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR PLANT, UNITS 1 AND 2 i PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION

_ RECEIPT OF PETITION FOR DIRECTOR'S DECISION UNDER 10 CFR 2.206 l

l: Notice is hereby given that by a Petition filed pursuant to 10 CFR 2.206 on Flay 28, 1997, Prairie Island Indian Community (Petitioner) requested that l the NRC (1) determine that Northern States Power (the licensee) violated the i

requirements of 10 CFR 72.122(1) by using its Materials License No. SNH-2506 l for an Independent Spent Fuel Storage Installation (ISFSA) prior to i

! establishing conditions for safely unloading the TN-40 dry storage containers; (2) suspend Materials License No. SNM-2506 for cause under 10 CFR 50.100 until j such time as all significant issues in the unloading process, as described in

{ the Petition, have been resolved, i.he unloading process has been demonstrated, and an independent third-party review of the TN-40 unloading procedure has been conducted; (3) provide Petitioners an opportunity to participate in the reviewing of the uq1oading procedure for the TN-40 cask, hold hearings, and l

allow Petitioners to participate fully in these and any other procedures 4

initiated in response to tne Petition; and (4) update the Technical Specifications for the Prairie Island ISFSI to incorporate mandatory unloading procedure requirements.

The Petition has been referred to the Director of the Office of Nuclear Reactor Regulation. By letter dated June 27, 1997, the Director denNd Petitioner's request for immediate action. As provided by 10 CFR 2.206, further action will be taken within a reasonable time.

9707010240 970627 PDR ADOCK 05000282 G PDR ,

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4 A copy of the Petition is available for inspection at the Comission's Public Document Room at 2120 L Street NW., Washington, DC, and at the local public document room located at the Minneapolis Public Library. Technology and Science Department. 300 Nicollet Mall, Minneapolis, Minnesota.

Dated at Rockville. Maryland, this 27th day of June 1997.

FOR THE NUCLEAR REGULATORY COMMISSION i Frank J ira 1 , Acting Director Office of Nuc ear Reactor Regulation i

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• Curtis C:mpbell Sr. D:relynn Lehts President Vice-President Byron White Alan W. Childs Sr.

Secretary Treasurer Michael J. Childs Sr. '

Assistant Secretary / Treasurer I L. Joseph Callan

! Executive Director of Operations -

US Nuclear Regulatory Commission Weshington, DC 20555 j

Dear Mr. Callan:

i

, Enclosed please find a petition, pursuant to Section 2.206, Title 10 of the Code of Federal i Regulations (CFR). The Prairie Island Indian Community is petitioning the Nuclear  ;

I Regulatory Commission (NRC) to review the procedure developed by Northern States l Power (NSP) to unload a Transnuclear dry cask (TN-40) in use at the Prairie Island l Nuclear Generating Plant. l

} We strongly believe the issues we are raising in this petition are worthy of immediate i

consideration by the Commission. l

Thank you for your attention to this inatter.

l Respectfully, b (L

/

YA441k]

l President' D lynnphto, Vice President A4&w x yrd te,' Secretary 'Afan Childs Sr., Treasurer i e i

Michael Childs Sr., Ass't Sec./freasurer 2

1

a 5636 Sturgeon Lake Road

• Welch, MN 55089

9d (612) 385-2554 + 1-800 554-5473

• Fax (612) 388-1576

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L. Joseph Callan RE: 2.206 Petition May 28,1997 j Page 2 s

Distribution List <

Senator Paul Wellstone Representative Gil Gutknecht s NRC Chainnan Shirley Ann Jackson NRC Region III Administrator Arthur B. Beach NRC Spent Fuel Project Office Director William I. Kane NRC Spent Fuel Project Office Deputy Director Charles J. Haughney BIA Minneapolis Area Office Director Larry Morrin e

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION IN THE MATTER OF:

Docket 72-10 s

THE PRAIRIEISLANDINDIAN COMMUNITV )

Petitioners )

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vs l,

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)

UNITED STATES NUdLEAR REGULATORY COMMISSION )

Respondent i Petition i l

Pursuant to 10 CFR Part 2.206 of the Commission's regulations, the Prairie Island Indian Community petitions the Nuclear Regulatory Commission (NRC) to:

1) Determine that Northern States Power (NSP) violated the requirements of 10 CFR 72.122(1) by using its Materials License No. SNM-2506 for an Independent Spent Fuel Storage Installation (ISFSI) prior to establishing conditions for safely unloading the TN-40 dry storage containers.

2) Suspend Materials License No. SNM-2506 for cause under 10 CFR 50.100 until such time as all significant issues in the unloading process, as described herein, have been resolved, the unloading process has.Acen demonstrated, and until an independent third party review of the TN-40 unloading procedure has been conducted.

3) Provide petitioners an opportunity to participate fully in the reviewing the unloading procedure for the TN 40 cask, hold hearings and allow petitioners to participate fully in these and any other procedures initiated in response to this petition.

4) Update the Technical Specifications (TS) for the Prairie Island ISFSI to incorporate mandatory unloading procedure requirements.

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! fagts i

1. Northern States Power (NSP) owns and operates the Prairie Island Nuclear Generating

- _ Plant on Prairie Island in Minnesota. - --

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2. The Prairie Island Nuclear Generating Plant is located next to the Prairie Island Indian

! Community.

3. The Prairie Island Indian Community is a federally recognized Indian tribe.

4. All government agencies andsxecutive departments, including the NRC, have a Trust

! Responsibility towards Indian tribes. The United States Government's "tmst

responsibility" toward American Indians, which is the unique fiduciary and legal duty of l the United States to assist Indians in the protection of their property and rights. This trust i responsibility arises out of treaties, statutes, executive orders, legal precedence, the United States Constitution and the course of dealings between the United States

Government and Indian tribes.

5. On April 29,1994, President Clinton issued an Executive Memorandum laying out the principles for every executive department and agency to follow in their interactions with federally recognized Indian tribes. At the core of these principles is the premise that the United States Government has a unique legal relationship with Indian tribes. Moreover,  :

the President's memorandum also stated that each executive department and agency shall l consult with tribal govemments prior to taking actions that affect tribal governments and I shall assess the impacts their actions have on tribal trust resources.

6. On August 31, 1990, NSP submitted a license application to the NRC for an  !

Independent Spent Fuel Storage Installation (ISFSI), pursuant to the requirements of Title l 10, Part 72 of the Code of Federal Regulations (10 CFR 72). The application was l assigned Docket #72-10. Included in the application were Technical Specifications and l Safety Analysis Report (TSSAR) and an Environmental Report (ER).

7. NSP's ISFSI application consisted of two components: the cask (a TN-40 cask, designed by Transnuclear, Inc.) and the actual storage area (the concrete pad). The application provided general information regarding the ISFSI, the type of cask to be used, conformity to design criteria (as required by 10 CFR 72, Subpart Pj.

8. In their application, NSP stated that they planned to operate the ISFSI for the licensed life of the plant. The design basis life of each cask is twenty-five years, although the NRC has stated that waste can safely be stored in dry casks for up to one hundred years.

9. NRC regulations require that storage systems (i.e., the casks) be designed to allow ready retrieval of spent fuel for further processing or disposal (10 CFR 72.122(1)). In their application, NSP addressed this requirement by ensuring that " fuel criticality is l

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prevented, cask integrity is maintained, and fuel is not damaged so as to preclude its 2

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4 I ultimate removal from the cask"(TSSAR, p. 3.2-1) and through a Decommissioning Plan ,

j described in the TSSAR (p. 4.6-1). Neither of these sections discusses the actual steps to

be taken and the likely problems to be encountered when removing fuel from a TN-40 dry .

i cask. With respect to decommissioning, there is just mention of removing waste from the

! TN-40 and shipping it in a licensed transportation cask. s

10. The Technical Specifications (TS) document is issued by the NRC and provides general guidance regarding the safe receipt, possession, and storage of irradiated nuclear

, fuel at the ISFSI (i.e., design criteria, cask operating limits, surveillance requirements,

! etc.). For each issue there is a definition (i.e., the limiting condition), its applicability,

action to be taken, surveillance: requirements, and the basis for the specification. For

example, two of the limiting conditions which would apply to unloading include the verification of the dissolved boron concentration of the spent fuel pool (e.g., greater than 3 or equal to 1800 parts per million (PPM)) to ensure that spent fuel is subcritical (TS p.

3/4-3) and ensuring that the outside cask surface temperature is not greater than 250

• F (121 *C), which ensures that fuel cladding will be protected against degradation (TS p.

. 3/4-5). He TS specifies that if the cask surface temperature is greater than 250 *F, it

! must be unloaded.

11. On July 10,1992, NSP provided the NRC with information regarding the unloading

of a TN-40 dry cask, as requested. He procedure described by NSP is as follows:

j " Assuming that the spent fuel in the TN-40 cask will be transferred to a licensed

, transportation cask using a normal 'in pool' fuel transfer, the sequence of operations j discussed in Section 5.1 of the ISFSI Safety Analysis Report (SAR) and in particular listed in Table 5.1-1, will be essentially performed in reverse." The letter lists the steps j taken in Table 5.1-1 of the SAR, only in reverse. (Attachment A).

l 12. On July 28,1992, the NRC issued a Finding of No Significant Impact (FONSI) based

! on its Environmental Assessment (EA) for the site. The EA referenced both the TS and I the SAR included in NSP's application. The NRC found that no significant impacts from ,

the constmetion of the ISFSI were to be expected. No impacts were expected from the I i operation of the ISFSI either.

With respect to radiological impacts, NRC staff expected that impacts from cask loading and preparation would be minimal. Table 5.3 in the EA (Attachment B) describes the

. steps that will be taken by NSP to receive, load, decontaminate, and store a TN-40 dry cask. There is no discussion of how a cask might be unloaded.

Table 6.1 in the EA summarizes the radiological occupational exposures expected to occur as a result of cask loading, decontaminating, and placement on the ISFSI (Attachment C). There is absolutely no mention or discussion of cxpected radiological epational exposure during cast unloading.

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j With respect to cask decommissioning (i.e., at the end of service), the EA only mentions that fuel could be removed from the storage cask and placed in a certified transportation

! cask for shipment to a repository.

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13. On October 19,1993, NRC issued Materials License No. SNM-2506 to NSP for the s

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ISFSI. Under the terms of the license, NSP is authorized to receive, possess, store, and j transfer Prairie Island Nuclear Generating Plant spent fuel at the ISFSI in up to 48 TN-40 '

casks for a period of twenty years. , l l

Along with the license, the NRC transmitted a Safety Evaluation Report (SER), which is

! an evaluation and revie.w of the:TSSAR submitted by the licensee. The review of the 1

TSSAR addresses the handling, transfer, and storage of spent fuel in a TN-40 dry storage

cask at the ISFSI. The SER discusses the design features of the TN-40 (e.g., 40

assemblies, enrichment factors, rninimum cooling, etc.), protection against environmental

conditions, natural phenomena, confinement barriers (i.e., protection of fuel cladding to l i ensure that degradation and gross rupture do not occur over the design life of the ISFSI),

i and criticality control (NRC regulations require that the spent fuel handling, transfer, and l storage system be designed to be maintained suberitical (10 CFR 72.124(a)).

s The maximum acceptable cladding temperature should not exceed 340 *C (644 'F),

according to the SAR, during normal storage conditions to prevent cladding degradation and subsequent gross ruptures. ,

The SER states that the procedure to unload a TN-40 cask will be a reverse of the loading sequence. No mention is made of the potential safety issues that may be associated with cask unloading (cask reflooding, thermal shock, flash steam, etc.) and potential exposure to workers.

14. Within the ISFSI license (SNMd506), a number of preoperational license conditions were specified:

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A training exercise (dry run) of all TN40 loading and handling activities must be ,

conducted and shall include the following (but is not limited to); l

a. Moving cask in and out of spent fuel pool area l

b. Loading fuel as sembly (using a dummy assembly) l

c. Cask drying, sealing, and cover gu h-M11ing operations

d. Moving cask to, and placing it on, the storage pad

e. Returning the cask. to the auxiliary building

f. Unloading the casl:

g. Decontaminating the cask

h. All dry run activities shall be done using written procedures L 'Ibe activities above shall be performed or modified and performed to show that each activity can be successfully executed before actual fuel loading.

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! As specified by the NRC, the above listed steps did not need to be performed in the order

listed.

, 15. On March 6,1995, NSP requested an exemption to the requirements of 10 CFR

72.82(e). 10 CFR 72.82(e) requires that "a report of the preoperational test acceptance s criteria and requirements must be submitted to the NRC....at least 30 days prior to the

, receipt of spent fuel or high level radioactive waste." The purpose of this requirement is j to allow the NRC 30 days to review and assess the licensee's ability to load (or unload) a j cask. NSP requested that they be exempted from the 30 day waiting period following the submission of their preoperational test results (i.e., loading and unloading) and instead be

) allowed to wait only three days before they loaded the first cask.

16. On April 20,1995, NSP transmitted to the NRC a report of their preoperational test

results, pursuant to the requirements of 10 CFR 72.82(e). Within the correspondence

were loading and unloading procedures and work orders requesting that certain aspects of

each procedure be tested.

With regard to preoperational testing, NSP staff submitted a work order to internally test j certain steps in the unloading procedure (D95.2) in February 1995 (Note: the work orders

< were included in the loading and unloading procedures). The request specified that steps

} 7.1 through 8.11 be tested (7.1 prepares the cask for transportation back to the Auxiliary l Building,8.11 tequires the removal of the lid).

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I In a later letter to NSP, however, the NRC noted that they did not require that the lid be l removed under water and the cask to be filled with water to prevent any damage to the cask (Attachment D).

17. NRC staff have stated to Prairii. Island Indian Community staff that they (the NRC) i have no formal mechanism to approve or disapprove loading or unicading procedures.

Any deficiencies to the procedures are identified through the NRC's inspection program.

A Notice of Violation (NOV) is transmitted to the utility if deficiencies are identified and the utility must correct the problem.

I8. On April 28,1995, the NRC conducted a public exit interview with NSP to present findings relative to dry cask storage activities at the Prairie Island plant. Among other things, the availability of space in the spent fuel pool to hold assemblies from a TN-40, in the case of an emergency off-loading, was questioned (i.e., would there be enough space in the spent fuel pool).

19. On May 3,1995, NSP provided to the NRC information regarding the unloadmg of a TN-40 cask. In their letter, NSP stated that they could unload a cask "if such action

-becomes wry"(Attachment E).

'Ihe unloading procedure transmitted in this letter is comprised of 9 steps, which are essentially the reverse of the loading procedure in the SAR. The information was 5

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] requested by the NRC because of questions raised during the April 28,1995 exit

! interview regarding the capacity of the spent fuel pool (after a planned outage) to store

waste after the first cask has been loaded (i.e., if a cask needed to be unloaded).

20. On May 5,1995[in response to NSP's submittal of information regarding the unloading of a TN-40 cask, the NRC found the plan for an unanticipated unloading prior

! to loading another cask would allow ready retrieval of the spent fuel for further

processing or disposal as required by 10 CFR 72.122(1).

21. On May 11,1995, the NRC gmnted NSP the 10 CFR 72.82(e) exemption. In the end, NSP waited 20 days before they loaded their first casks, instead of the requested 3.

j days.

i 22. On May 12,1995, the NRC approved the preoperational test report and authorized j' NSP to load the first cask. The first cask was loaded immediately.

23. On June 30,1995, the NRC issued an Inspection Report covering a variety of dry l cask storage activities between January 24 and May 11,1995 (Attachment D). Among-

! other things, the inspection assessed NSP's performance relative to dry cask storage

, activities. The Inspection Report noted that the licensee (NSP) "did not complete review

! and approval of the unloading procedure igttil the day followine the submincion of the

oreooerational test reoort (emphasis added). Submission of this report [to the NRC]

implied that the licensee was ready to load a cask with spent fuel and subsequently unload i l the cask, ifnecessary."

! In that same Inspection Report, the NRC issued NSP a Notice of Violation (NOV) for not l including certain technical specifications in their unloading procedure. The letter  !

i specifically cited NSP for omissionsin the unloading plan and "overall poor planning for

} dry cask storage activities." The violations with regard to the unloading plan were i i technical specification deviations: 1) verification of boron concentrations not adequately

specified (the concern is the potential for inadvertent criticality); and 2) verification of

! fuel integrity not adequately specified; no hold point was identified to ensure that work would not continue until results had been reviewed. NSP later corrected these omissions.

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24. On May 30,1996, at a briefing for the commissioners of the NRC, Andrew Kugler, l Iead Project Manager, Dry Cask Storage, NRC, stated while the NRC has found dry cask i loading procedures to be acceptable, the unloading procedures are far more complex i (Attachment F). In loading a cask, the fuel has been characterized (i.e., its integrity i

verified), a loading dry run has been performed, and that licensees can take advantage of

lessons teamed from other licensees (in loading their casks).

' l Mr. Kugler stated that the older SAR's [ Safety Analysis Reports] do not recognize this complexity and indicate that unloading would be the reverse of loading, which Kugler 1 stated was not true. The unloading procedure in NSP's most recent SAR, and the SER i for the ISFSI, is described as the reverse of unlaading.

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! i j Of particular concern with respect to unloading, stated Mr. Kugler, are the potential  !

condition of the fuel nnd issues associated with the reflooding of the cask with spent pool ,

water: cask pressurization due to steam generation as colder spent fuel pool water is I placed into the cask. Thermal shock to the fuel during reflooding and radiological s exposure to workers during the operation from the venting the caak (venting will either be 4

done directly to the pool or the ventilation system).

l j ;25. On November 8,1996, NSP submitted to the NRC a revised TN-40 unloading plan.'

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26. The SAR for the Prairie Island ISFSI states that the fuel cladding may be as hot as 1 1

-340 *C (644 'F) and spent fuel pool may be as cool as 110 'F. In the SAR, NSP stated that before a cask was returned to the spent fuel pool for unloading " cold water would be

! pumped into the cavity to reduce the temperature" and that steam might be produced i when the water hits the cavity surface. Although the SAR recommended pumping cold j water into the cask prior to immersion in the spent fuel pool, the TN-40 un6Hng i

procedure contains no such provision (see Attachment G).  !

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{ 27. The Prairie Island SAR also states that the fuel assemblies should be inspected for i any physical damage which could potentially cause problems during removal from cask.

l De actual unloading procedure, however, contains no such requirement. He procedure

only requires that the location of three of the forty assemblies be confirmed prior to off-l loading (recall that NSP received a NOV for not including the fuel verification requirement in their procedures in June 1995). As Mr. Kugler mentioned on May 30, ,

l 1996, the potential condition of the spent fuel, with respect to reflooding, is a concern. '

28. He SAR only discussed potential occupational exposure to radiation with respect to i d

receipt of cask, cask loading, deco 6tamination, removal to storage area, and periodic '

j maintenance (Attachment H). There is no mention of potential radiological exposure to

! workers during cask unloading. The issues raised by Mr. Kugler, with regard to potential

! increases in radiological occupational exposure during cask unloading are not addressed (especially flash steam due to the hotter temperature of the inside cavity (644 'F) being

exposed to the much cooler spent fuel pool water (110 'F).

29. Mr. Kugler stated on May 30,1996 that there was " essentially no cask unloading j experience" for licensees to look back on for lessons learned, like there is for cask i loading. In a later letter to Dr. Mary Sinclair of Don't Waste Michigan, Mr. Kugler stated j that there were, in fact, three instances where casks were unloaded during the loading

evolution, due to problems encountered (Attachment I). In all three instances, however, the loading evolution had not yet been completed and none of the casks had moved j ' beyond the decontamination area (i.e., none of the casks had been removed to the storage l area).

30. A dry cask, after it is has been loaded and in storage /use, has never been unloaded

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I j Petitioners Claim

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] 1. The procedure to unload a TN-40 dry cask at the Prairie Island Nuclear Generating

Plant has not been adequately evaluated or tested by either NSP or the NRC.

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} 10 CFR 72.122(1) provides that:

i Retrievability. Storage systems must be designed to allow ready retrieval of spent fuel or high-level radioactive waste for further disposal or storage.

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$ Our interpretation of 10 CFR 725122(1) is that the storage system (i.e., the cask ) must be

! designed to allow ready retrieval (i.e., unloading) of spent fuel, not whether the spent fuel pool can accommodate spent fuel from a cask needing to be unloaded. This requirement has not been met because the neither the NRC nor NSP has completely demonstrated whether a TN 40 dry cask can be (or has been) unloaded aner it has been sitting on the

storage pad for a number of years. Thus the NRC regtdrement that storage systems be j designed to allow ready retrieval of spent fuel has not been met because it has never been fully evaluated. No dry cask has ever been unloaded aner it has been in use.

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. The question of"retrievability" was discussed at a public exit interview convened by the j NRC in April 1995. The context in which the issue of retrievability was discussed, '

! however, was wrong. Participants at the public exit interview were asking whether NSP would have enough space in spent fuel pool for the 40 assemblies they planned to load

, into a TN-40 cask right before a planned outage if an emergency warranted offloading i the first TN-40 cask. The NRC may recall that NSP's spent fuel pool was already full i and there was a possibility that, aRet the planned outage, there would not be enough space in the spent fuel pool for the 40 assemblies (they would be short fifteen spaces). In l response to these concerns, NSP assured the NRC that they would indeed have enough l space in the spent fuel pool if they moved some non-fuel bearing components. In their

. letter, NSP calculated the number of spaces in the spent fuel pool they would have available, noting that they would be short by fiReen spaces, and offered two scenarios

, under which they might make space available. The NRC appeared satisfied that the requirements of 10 CFR 72.122 (1) had been met.

l As stated above, NSP submitted their preoperational test acceptance criteria and results

! (and loading and unloading procedures) to the NRC before they were done testing it.

l Also as stated above, the NRC cited NSP for omissions in their unloading plan and noted

[ that NSP "did not complete review and approval of the unloading procedure until the day following the submission of the ysperational test report. Submission of this report implied that the licensee was ready to load a cask with spent fuel and subsequently unload i the cask,if necessary."

! How could NSP have claime;l they were ready to load, and subsequently unload a TN-40

cask, if necessary, if they had not completed a review and testing of their own I

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4 i procedures? If these procedures were not fully tested,can the NRC be sure that 1) the licensee has the ability to unload a TN-40 cask and 2) these casks can be unloaded safely.

How can members of the Prairie Island Indian Community feel safe that the casks can and j .

will be unloaded should someAing go wrong with one of the casks? Many tribal i members are fully aware of the dry cask situation at the Palisades plant (i.e., the problem s

] with the VSC-24) and do not wish to see this occur at Prairie Island.

The TN-40 dry cask is in use only at the Prairie Island Nuclear Generating Plant. Here are no other plants currently employing a cask designed by Transnuclear Inc. He NRC has licensed the TN-24, but it is not in use anywhere in the country. In affidavit to the Mmnesota Court of Appeals, Mr. Jon Kapitz, NSP's Project Manager for Dry Cask l

Storage, stated that a TN-24P storage cask was successfully unloaded as part of a project sponsored by the Department of Energy (DOE) and the Electric Power Research Institute (EPRI). (Note: this Affidavit was submitted in response to legal action initiated by the i Prairie Island Indian Community). In 1987, the TN-24P was tested in a cooperative

! research program sponsored by the DOE, Virginia Power Company, and EPRI. He

' purpose of this research was to determine the thermal, shielding, and operational performance of the TN-24P storage cask (not whether it could be unloaded). The testing was conducted at the Idaho National Engineering Lab (INEL) using fuel that has been irradiated at the Surry plant in Virginia. The fuel was moved from Virginia in TN-8 transportation casks (which hold three PWR assemblies). Dry runs, to train personnel, were performed with nonitradia.ted fuel (i.e., dummy assemblies),

ne transfer of spent fuel from the TN 8L into the TN-24P was done in the INEL hot shop in the air, via remote operation. That is, the TN-8 was not placed back into a spent fuel pool for transfer to the TN-24P. With respect to the performance of the TN-24P, the report stated that "the test demonstrated that the cask could be satisfactorily handled and loaded dry" (i.e., not in the pool). Therefore, the issues raised by Kugler, with respect to reflooding a hot cask (flash steam, pressure build-up, fuel integrity, etc.) would not have been addressed in report. This experience dm aqt demonstrate that a fully loaded TN-40 dry cask can be safely unloaded after it has been out on the storage pad.

2) The NRC allowed NSP to load their first TN-40 cask, and subsequent casks, without a full evaluation of the unloading procedure. Both the SAR and SER for the ISFSI stated that unloading a cask is the reverse ofloading a cask, implying that it was quite easy to do. NRC staff, however, have stated that cask unloading is quite complex and not the reverse ofcask loading.

As stated above, NSP had not finished testing its own loading and unloading procedure 1 before submitting its preoperational test report to the NRC, thereby declaring that they were ready to load, and unload a cask if-my. According to the actual loading and unloading plan and preoperational test results, only part of the unloading procedure has been tested (refer to Fact No.16). How can the Prairie Island Indian Community be assured that a cask can be unloaded if the procedure has never been fully tested before andit has never been done?

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3-l l 3. As stated above, NSP submitted a revised unloading procedure to NRC in November j 1996. It is our belief that NSP's unloading procedure does not address the problems that likely would be encountered prior to and during the procedure.'NRC staff have identified -

potential unloading problems that have not been addressed in NSP's unloading plan. Our

analysis of the unloading plan identified several deficiencies

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! a) Failed Fuel Considerations Although the revised plan contains a step to sample the intema! gas for fission products (as an indication of fuel failure), there are no procedures

on what to do if radioactive air concentrations indicate fuel with gross cladding defects.

If the internal gas sampling indicates failed fuel, procedures abould be developed to recover the fuel from the cask without contaminating the spent fuel pool. As the j unloading procedure is currently written, the dry cask is placed in the pool for unloading,

regardless of the radioactive concentration in the intemal gas. This action may 2

contaminate the fuel pool with microscopic irradiated fuel particles (fuel fleas), thereby l creating a major radioactive hazard. (This has occurred at Southern California Edison's I

, San Onofre 3 reactor). ,

l There are no procedures in place to determine whether fuel in a TN-40 cask has gross cladding defects. During storage, the cladding is under pressure and may crack, due to j the high internal temperature of the cask (up to 644*F). Over time, the cracks can i become larger. This is a cumulative process; the longer fuel is in storage, the greater the l likelihood of cladding degradation. If fuel with cladding defects is utsaded, the spent i fuel pool may become contaminated and workers may be exposed to radiation. Plant

personnel will have to decide whether or how fuel with cladding defects will be removed.

1 i b) Venting of Radioactive Gases The unloading procedure does not indicate whether or I how radioactive gases might be vented from the cask, if concentrations are greater than i certain limits. Plant personnel will have to. decide how to vent the . cask. If the 1 radioactive air is to be vented, is a permit required?

I c) Radiation Monitors Before venting the cask, a stop-check must be instituted to verify that ventilation systems and radiation monitors are functioning. There is no such stop-

, check in NSP's unloading plan.

d) Steam Build-Un When water is pumped into the TN-40, it is very likely that steam j will be created and pressure will build within the cask if the fuel cladding is greater than

'; 212*F. Although NSP has included steps to cool down the cask and fuel cladding, it is

. not clear what the maximum cladding temperature will be prior to the addition of water into the cask. The steam overflow from the cask, which is directed into the spent fuel pool, is likely to be very hot. Warnings need to be posted regarding this hazard.

s The pressure build-up also places certain stresses on the hose couplings and piping-this should be evaluated. As the unloading procedure is currently written, water is pumped into the cask at a rate not to exceed 10 psig and the temperature is kept below 240 *F, to 10

0 ensure that pressure not build up too quickly. The thermometer measuring the temperature should have a range greater than 50 *F to 300 *F; the maximum temperature

-should be 900 *F. The pressure gge should also be changed to one with a maximum pressure of 50 psig.

s

4. He Prairie Island Indian Community respectfully requests that the NRC thoroughly review the unloading procedure written by NSP to ensure that the technical issues raised by Mr. Kugler have been met. The NRC, as a federal agency, has special obligation to protect the trust resources ofIndian tribes. Trust resources includes the health and safety of tribal members.

Conclusion The public relies on the NRC to make decisions that these containers are safe, based on a comolete evaluation. Neither the NRC nor NSP have fully evaluated the TN40 unloading procedure, as evidenced by the lack of documentation regarding cask unloading and potential consequences in either the SAR or the SER. Both the SAR and SER implied that unloading a cask was very straightforward and just the reverse of loading a cask. It now appears that the NRC has somehow reversed itself(with respect to cask safety) by stating that there are now some concerns with regard to unloading a cask.

NSP has not adequately demonstrated their ability to unload a TN40 dry cask.

A dry cask has never been unloaded, there is no certainty that it can be done. The NRC has an obligation ~to the Prairie Island Indian Community to review the TN40 unloading procedure to determine whether the issues raised by their own staff and within this petition have been fully evaluated and included in the unloading plan for a TN40 cask.

11

Anoendix of Attachments Petition of the Prairie Island Indian Community RE: Northern States Power procedure to unload a TN-40 dry cask s

Attachment A 4

Table 5.1-1 of the Safety Analy sis Report (SAR) and July 10,1992 letter (NSP to NRC)

Attachment B ,

Table 5.3 of Environmental assessment (EA) for the ISFSI

Attachment C i

Table 6.1 of the EA Attachment D June 30,1995 letter from Greenman (NRC) to Watzl (NSP) and Notice of Violation Relevant pages only Attachment E May 3,1995 letter to NRC from NSP regarding cask unloading Attachment F .-

Transcript from May 30,1996 Commissioners meeting Relevant pages only Attachment G Step 8.21 ofNSP's unloading plan Attachment H Table 5.1-2 of the SAR for the ISFSI Attachment 1 June 18,1996 letter to Dr. Mary Sinclair from Andrew Kugler 12

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• ISFSI SAR TABLE 5.1-1  !

j j SEQUENCE OF OPERATIONS l 1

i A. Receiving i

1. Unioad empty cask and separately packaged seals at plant site. l

\

4 2. . Inspect the following for shipping damage: exterior surfaces, sealing i surfaces, trunnions, seals, accessible interior surfages and basket j assembly, bolts, bolt holes and threads, neutron shield vents. l j 3. . Remove weather shield and: install plug in neutron shield vent hole. j i (threaded hole in the top of the steel shell surrounding the resin which contains a pressure relief valve during storage).

4 Remove lid bolts and lid.

5. Install protective plate over cask body sealing area.

4

! 6. Obtain lid and lid seal from storage.

i 7. Attach lid seal to lid by means of six retaining screws. j

8. Nove to spent fuel pool area. jI i '

i B. Snent Fuel Pool Area i  !

Iower cask into cask loading pool,

! 1.

i l

2. Imad preselected spent fuel assemblies into the 40 basket compartments.

l d

. 3. Verify identity of the fuel as',semblies loaded into the cask.

1

! 4. Remove protective plate from cask body flange. )

\ {

4

5. Iower lid and place on cask body flange over the two alignment pins.

j 6. Lift cask to surface of pool and install lid bolts.

7. Connect drain line to quick-disconnect coupling in the drain port.

) '

t

! 8. Bolt special adapter, with quick disconnect coupling, to vent port bolt i holes.

1 i

9. Connect plant compressed air line to special adapter quick-disconnect i coupling.

10. Pressurize cavity to force water from cavity through drain port to the

! spent fuel pool.

~

.)

i

, s .. ,.

,.. ~

l TAB 12 5.121 REV. 2 9/91

4

. ISFSI SAR TABLE 5.1-1 (Centinusd)

SEQUENCE OF OPERATIONS i

11.

Disconnect disconnect couplings.plant compressed air line and drain line from their quick -

12.

Move cask to the decontamination area.

s C.

Decontamination Area (Rail Ravi i 1.

Decontaminate cask until acceptable surface dose levels are obtained .

2.

1 Torque lid bolts using the prescribed procedure.

i 3.  :

' Remove plug from tieutron shield vent and install pressure relief valve.

i 4.

i Connect Vacuum Drying System (VDS) to vent port.

5.

I Evacuate cavity to remove remaining moisture using prescribed procedure .

6.

{ Break air intovacuum by closing vacuum valve and opening air valve to admit dry the cavity.

7.

Disconnect VDS at vent port and install vent port cover with seal and bolts.

8. .I Connect drain port.Vacuum-Backfill System (VBS) to quick-disconnect coupling in the 9.

Evacuate cavit:y to 10 millibar and backfill with dry helium gas.

10.

Pressurize cavity to about 2 ata with helium.

11.

Disconnect VBS at the drain port quick-disconnect coupling and install drain port cover with seal and bolts.

12.

Perform helium leak test of lid seals.

13. Remove overpressure port cover.

14. Install top neutron shield drum.

15.

Torque the bolts using prescribed procedure. I 16.

Pressurizeofoverpressure pressure about 5.5 sta.system, (seal interspaces), with Helium to a 17.

Perform leak test on overpressure system.

~

O

. MjE: ,-

1 TABLE.s5.1.'xA, REV. 2 9/91

. ISFSI SAR TABLE 5.1-1 (Csntinusd)

SEQUENCE OF OPERATIONS 4

4

18. Check external surface temperatures using an optical pyrometer.

i

19. Check surface radiation levels.

s

20. Install protective cover with seal and bolts. (could be performed at storage area)

21. Load cask on transport vehicle.

~

i 22. Move cask to Storage Ares.

l D. Storage Area i 1. Unioad cask from transport vehicle.

j 2. Position cask in preselected location on storage pad.

3. Check for surface defects.

4. Connect pressure instrumentation to cask and to monitoring panel. l

) S. Check that pressure instrumentation is functioning.

6. Check surface radiation levels.

G e

's

. TABLE 5.1-1 REV. 2 9/91

+.@4E

- ..-.. .- . - - - . - . . - - - . . . . =- -

Concrete Pad The storage casks will be stored in two parallel rows of 12 casks on each of two 216-foot long x 36-foot wide x 3-foot thick concrete pads. The two slabs will be positioned end to end with 40 feet in between. To improve foundation performancs and earthquake safety,3 $

feet of soil beneath each slab will be excavated and replaced with comp =M structural fill.

'Ihe pad elevation will be 693 feet 6 inches above mean sea level (m:1) to preclude immer-sion of the cask seals during the pmbable maximum flood. "Ihey will be surrounded by a 17-foot high eartian berm.  :

! 5.3 ISFSI OPERATIONS i .

l Fuel handling and cask loading operations in the Auxiliary BuiMing will be done in

! accordance with requirements of the Prairie Island Nuclear Generating Plant 10 CFR Part 50

Operating Tiram DPR-42 (Unit #1) and DPR-60 (Unit #2). Cask traaW and storage at j the ISFSI will be subject to requirements of the Prairie. Island ISFSI 10 CFR Part 72 l U- . 'Ihe major steps ==M=W with the placing of fuel in the Prairie Island ISFSI are ,

i j presented in Table 5.3. -

4 i

. TABLE 13 j ISES[ OPERATIONAL EDES i

i A. RECEIVING

~

i i 1. Unload empty cask and separately packaged seals at plant site.

i j 2. Inspect the following for =Maping damage: exterior surfaces, sealing surfaces, trun-nions, seals, accessibic inter.ior airfaces and basket mienhly, bolts, bolt holes and

! threads, neutron shield vents.

i i

22 i,

M i

I

9 4

3. Remove weather shield and install plug in neutron shield vent hole.

4. Remove lid bolts and lid.

j

5. Install protective plate over cask body sealing area. s

6. Obtain lid and lid seal from storage. -

i.

7. Attach lid seal to lid by means of six retaining screws.

8. Move to spent fuel pool area.

B. SPENT FUEL POOL AREA

1. Lower cask into cask loading pool.

2. ' bad preselected spent fuel assemblies into the 40 basket compartments.

3. Verify identity of the fuel assemblies loaded into the cask.

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4. Remove piMye plate from cask body flange. l 4

5. Lower lid and place on cask body flange over the two =11pmaat pins.

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6. Lift cask to surface of pool and install lid bolts.

7. Connect drain line to quick di+:-:-.a=-N coupling in the drain port.

8. Bolt special adapter, with quick d!+x+=-M coupling, to vent port bolt holes.

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_ ,__ .- _ ,_ _. _ __ .- . _ _ ___.___ _ _ ._...__. ..._ _ .._ .. _ _ . _ _.. _ _._~_._ _ _ _ _

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9. Connect plant compressed air line to special adapter quick-disconnect coupling.

i j 10. Pressurize cavity to force water from cavity through drain port to the spent fuel i pool.

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11. Disconnect plant compressed air line and drain line from their quick-disconnect couplings.

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12. Move cask to the decontarninntinn area.

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4 C. DECONTAMINATION AREA (RAIL BAY) 4 i 1. Decontaminate cask until acceptable surface contaminatW levels are obtained.

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2. Torque lid bolts using the pre M procedure.

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3. Remove plug from neutron shield vent and install pressure relief valve.

l 4. Connect Vacuum Drying System (VDS) to vent port.

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l 5. Evacuate cavity to remove r*mmining moisture using prescribed procedure.

6. Break vacuum by closing vacuum valve and opening air valve to admit dry air into the cavity. -

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, 7. Di={-,1=t VDS at vent port and install vent port cover with seal and bolts.

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j 8. Connect Vacuum-Backfill System (VBS) to quick dir{+nen coupling in the drain

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i 9. Evacuate cavity to 10 millibar and backfill with dry helium gas.

4 l 10. Pressurize cavity to about 2 ATM with helium.

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11. Disconnect VBS at the dain port quick-connect coupling and install drain pon cover s l with seal and bolts.

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12. Perform helium leak test oflid seals.

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i 13. Remove over pressure port cover,

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! 14. Install top neutron shield drum. i 1

l 15. Torque the bolts using preM procedure.

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16. Pressurize over pressure system with Helium to a pressure of about 5.5 ATM.

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17. Perform leak test on over pressure system.

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} 18. Check external surface ^speratures using an optical pyrometer.

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! 19. Check surface radiation levels",

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20. Install protective cover with seal and bolts.

i i 21. I. cad cask on transport vehicle.

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22. Move cask to Storage Area.

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. . D. STORAGE AREA

! 1. Unload cask from transport vehicle, l 2. Position cask in pr== Mad location on storage pad.

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3. Check for surface defects.

] 4. Connect pressure instrumentation to cask and to monitoring panel.

i j 5. Check that pressure inmusw. station is functioning.

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6. Check surface radiation levels.

The admiaientive procedures for the ISFSI will be the.same as those used for the Prairie Island Nuclear Generating Plant. Any changes to these procedures will be reviewed and approved by the Station Operations Committee and Safety Audit Committee. Iiefore startup and during the lifetime of tiie ISFSI, the cask monitoring instrumentation, the electrical system, the com=miadons system, and the storage casks will be tested to ensure their proper functioning. De existing training program at the plant will be used to provide and maintain a well qualified work force for safe and efficient operation of the ISFSI. All personnel working in the fuel storage ' area will receive radiation and safety training and those actually performing cask and fuel handling functions will be given additional tmining in specific areas as required by the PadinewtProtection program in effect at the Prairie Island Nuclear Generating Plant.

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TABLEfl DESIGN BASIS OCCUPATIONAL QHE IIbig EXPOSURES DURING CASE LOADING. TRANSPORT AEn EMPLACEMENT 8 s

Task Time No. of Dose Rate Dose Required (hr) persons (mrem /hr) (Person-rem)

Placement in pooF 2 5.0 3, 0.03

~

Imding pincess 5 5 5.0 0.125 Removal from pool 5 5 30.0 0.75 Transfer to decontaminatino area 1 3 30.0 0.09 Piccessing of cask 6.5 2 30.0 0.39 Helium leak test 2 2 30.0 0.12 Decontamiandon 2 3 30.0 0.18 Install neutron shield, .

pressurize, test - 3 2 30.0 0.18 Preparation for transport 1 3 30.0 0.09 Transfer of cask to ISFSI 1 -

3 20.0 0.06 Final cask empl-t 2 5 30.0 0.30 -

TOTAL , 2.315 8 Dose rates at 1 meter were utilized for all cases except cask transfer, when individuals will typically be at least 2 meters away from the cask.

8 Steps from Table 5.3.

34 96 e- = wee

UNITED CTATES p , %'s NUCLEAR RESULATORY COMMISSION j

j 2

REGION M 801 WARRENVlu.E ROAD

• • g v ,/ usLE. ituNOIS 80632-4351

• • • • • June 30, 1995 I

i Mr. E. Watzl, Vice President '

! Nuclear Generation

. Northern States Power Company i 1 414 Nicollet Mall ,-

Minneapolis, MN 55401 l .

Dear Mr. Watz1:

! This refers to the special NRC inspection from January 24 through - -

1 May 11, 1995, of dry cask storage activities at the Prairie Island site.

i This inspection was conducted by the resident inspectors, selected RIII based

• 1 inspectors, and technical staff from the Office of Nuclear Reactor Regulation l l and the Office of Nuclear Materials Safety and Safeguards. The purpose of l i this inspection was to evaluate the acceptability of the as-built TN-40 cask j and to assess your performance relative to dry cask storage including the i nreooerational testina activities.

. m

! We discussed the results of this inspection with you and other members of your 3

staff at a public exit meeting on April 28, 1995. At that meeting we j identified five items that required further resolution. You provided us with I additional information for each of these items and we completed our review of

, the subject items during the next two weeks. On May 11, the NRC issued a schedular exemption from the requirements of 10 CFR Part 72.82(e) allowing you

! to submit the results of your preoperational test less than 30 days before the

! receipt of fuel at your onsite Independent Spent Fuel Storage Installation.

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

The enclosed copy of our inspection

• report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective

! examination of procedures and representative records, observations, and 3 irtterviews with personnel.

Based on the results of this inspection, we concluded that you were ready to i 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 j subject' cask. However, one violation of NRC requirements was identified i during the course of this inspection, as specified in the enclosed Notice of -

! Violation (Notice). This violation pertained to cask handling, loading, and .-

i unloading activities that were not prescribed by procedures of a type s"

) 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 j inspection. Therefore, no response with respect to this violation is required. However, we are disappointed that NRC inspectors, rather than your i own staff, identified these procedural deficiencies.

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3

. E. Watzl We also identified several weaknesses with your overall performance relative to dry cask storage activities. These weaknesses included: 1) poor oversight of vendor activities until late in the dry cask storage project; 2) lack of s

effective engineering involvement in vendor fabrication activities; 3) the ineffectiveness of your quality assurance organization in assessing vendor performance during the cask fabrication process; 4) the absence of a comprehensive plan for,. inspecting, auditing, and monitoring dry cask storage activities onsite, particularly those activities associated with the 10 CFR Part 50 license; and 5) overall poor planning for dry cask storage activities.

Based on the above weaknesses aRd as discussed at the exit meeting on April 28, we request that you provide us with a formal performance improv'ement 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 request within 30 days of the date of this inspection report. We will continue to < ?aluate the effectiveness of your corrective actions to improve your pe'. tonce in dry cask activities during future NRC inspections. ,

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, the enclosure, and your response to this letter will be placed in the HRC 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 s Reduction Act of 1980, PL 96-511.

We will gladly discuss any questions ycu have concerning this inspection.

1 Sincirely,

~

Ok -

T Edward G. Greenman W Senior Oversight Manager Region III Dry Cask Activities

(

Docket No. 50-282 Docket No.53-306 Docket No. 7F-10

Enclosures:

E 1.

2.

Notice ef Violation Inspection Report No. 50-282/95002; E 50-306/95002; 72-10/95002(DRP)

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See Attached Distribution

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1

." Watz1  ;

Distribution

cc w/ enc 1: Site General Manager, PINGP s  !

John W. Ferman, Ph.D., l Nuclear Engineer, MPCA I

. State Liaison Officer, State. '

of Minnesota .

State Liaison Officer, State of Wisconsin

Tribal Council

Prairie Island Dalota Community t

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- . - - - = . - - - - - - . - . - - . - . - . . . .- ... .- -. __. _ - .

l , NOTICE OF VIOLATION l Northern States Power Company Dockets No. 50-282; 50-306; 72-10  ;

i Prairie Island Nuclear Plant Licenses No. DPR-42; DPR-60; SNM-2506 -

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{ During an NRC inspection conducted from January 24 through May 11, 1995, a violation of NRC requirements was identif.ied. In accordance with the " General l i Statement of Policy and Procedures for NRC Enforcement Actions," 10 CFR l Part 2, Appendix C, the violation is listed below:

]

! 10 CFR Part 72.142(b) requires 3 licensee to establish, mainta h , and execute  !

i a quality assurance (QA) prograi with regard to an Independent Spent Fuel j Storage Installation (ISFSI) that satisfies each of the applicable criteria of

Subpart G, " Quality Assurance." In meeting the Part 72.142(b) requirement, i 10 CFR Part 72.142(d) accepts a Commission-approved quality assurance program l -

which satisfies the applicable criteria of Appendix B to 10 CFR Part 50. As i such, the ISFSI Safety Analysis Re> ort ' states that the previous 1v annroved i Northern States Power _QA program witch satisfies applicable criteria of 10 CFR

Part 50, Appendix B,'will De applied to activities, structures, systems, and components of the ISFSI commensurate witn snetr importance to safety. -

1 Criterion V of Appendix B to 10 CFR Part 50 requires that activities affecting i quality be prescribed by documented instructions, procedures, or drawings, of

! a type ap.)ropriate to the circumstances and that these activities be I accomplis 1ed in accordance with the associated instructions, procedures, or ,

drawings. Cask handling, loading, and unloading are activities affecting 4 quality.

l Spatrary to the abovi, cask handlina loadino. and unlandino activities were i j got erac m nae ov aoorovec 3roceduree af = tvna fooropriate to the ~

j }Wcumstances; as evidencec )y therfollowing examples: .

1. Surveillance Procedure. SP 1077, "Special Lift Fixture for the TN-40 Dask," did not address dimensional checks of the special lifting device, as required. -

~

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

3. Procedure D95.1, "TN-40 Cask Loading Procedure," s prerequisites section that SP 1077 be performe@pecified ays prior toinloading the j a cask. However, the TS 4.19-requirement to perform a visual _inenetion  !

of the lifting device (lift beam and extension) an& verify operability) of the device 7 days prior to use, was not identif eo in un.a. m re  !

also was no procedure identifying actions required to verify operability 1 of the lifting device.

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• Notice of Viola' av.. t

4. Procedure D95.1, "TN-40 Cask loadina Prr>cedure," did not include a step

to perform radiation wrvave of the cas ( surrace nefore moving a cask to Ine 1stsi, as required by TS 4.6.1.

i 5. Procedure D95.2, "TM-40 Cask Unloading Procedure," did not adtauntely '

i address the TS-requirement to samnia rne spent ruei pooi or >oron

! concentrauon witn1n tour hours of flooding tne cask cavily for

! unloaatng tne ruei assemonies. , , -.

6. Procedure D95.2, "TN-40 Cask Unloading Procedura" did not contain a hold

! point to ensure worc woaic not continue until tie results of tne inner

l'ITE volume sampie isad l>een reviewed. Inis procecural hold point-is j impor6ans to ensure snat an unpianned and unmonitored release path is -

j not created while the cask is in the Auxiliary Building.

7. The li~censee did not have a procedure for conducting 10 CFR Part 72.48 i safety evaluations. '

i .

i This is a Severity Level IV Violation (Supplement I) (50-282/95302-01;

} 50-306/95002-01; 72-10/95002-01(DRP)).

i I With respect to this violation, the inspection showed that steps had been

! taken to correct the identified violation and to prevent recurrence.

l 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

• e e

4

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l o While the inspectors reccgnized 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)ects of the procedure, which should have been in ciace oerore tne dry tor u ampie, lecnnicai specicication requirements were not

e ectively incorporated into the loading and unloading procedures '

)

(paragraph 3.2). In addition, the licennen die not complete rev ew and I l

_ approval of the unloading proceaure unti ;he c av r o' lowing su)m'ssion j g l j or the creoperationai test report. Submission 0 " "h-is rennu imDiled 7 '

4 that the 1- censee was reativ to oad a cask with spent fuel and l l subsecuentiy unload the cask, if necessary. -

i e The licensee did not take a disciplined approach to inspecting the fuel i designated for cask storage as evidenced by weaknesses identified by the 1 1 inspectors during observation of fuel inspection activities (paragr,aph l 7.3).

l l e Some weaknesses were'noted with the licensee's documented basis for l safety evaluation conclusions (paragraph 8.2).  :

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oparational checks of vehicle brakes, lifting equipment, turntables, jacks,

, and cask links.-

1 3.1.5 Surveillance Procedure. EP 1075. "TN-40 Fuel Selection and Identification" l The inspectors reviewed SP 1075 and the cask loading procedure, D95.1, to ,

verify that selected Technical Specification (TS) requirements had been incorporated into procedures. Surveillance requirements for ensuring that fuel assemblies which satisfy the criteria of TS 3.1.1 would be loaded into

.the cask, are defined in TS 4.1. ,

TS 3.1.1(6) required that, " fuel assemblies known or suspected to have i structural defects or gross cladding failures (other than pinhole leaks) sufficiently severe to adversely affect fuel handling and transfer capability shall not be loaded into the cask for storage." The licensee originally * -

intended to visually inspect fuel assemblies designated for loading with binoculars to identify any " structural defects or The inspectors questioned the efficacy of this tec$ross cladding failures."nique to pro inspection of the fuel. After further discussion with Region III staff on  ;

fuel inspection techniques, the licensee elected to use video recording )

, equipment to perform the fuel inspection. The inspectors considered this a 1 preferable method for identifying fuel anomalies and ensuring compliance,with 1 TS 3.1.1. The inspectors observed portions of the actual fuel inspection and i

identified weaknesses with the licensee's approach to this activity as l discussed in paragraph 7.3.

During the review of SP 1075, the inspectors identified that the procedure did not incorporate the requirement of TS 4.1.2, which stated that "before ,

inserting a spent fuel assembly into a cask..., the identity of each fuel assembly shall be independently verified and documented." The inspectors

! discussed the independent verification requirements of TS 4.1.2 with the licensee. Subsequently, the licensee revised SP 1075 to address independent 1 i

verification of fuel assembly identification. Based on observations of the l j actual fuel inspection, the inspectors concluded that the licensee not all TS  :

i requirements for fuel identification. The failure to incorporate the )

requirements of TS 4.1.2 into SP 1075 is considered an example of a violation i i of Criterion V of Appendix B to 10 CFR Part 50 (50-282/95002-01;  !

50-306/95002-01; 72-10/95002-01(DRP)). -

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! 3.2 Loadino and Unloadina Procedures

! The inspectors reviewed the loading (D95.1) and unloading (D95.2) procedures l for technical adequacy and to determine if the lessons learned from the l 1 preoperational testing / dry run had been appropriately incorporated into the l l procedures.

3.2.1 D95.1. "TN-40 Cask Loadina Procedure" a  :

l The original D95.1 procedure specified in the prerequisites section that i SP 1077 be perfomed 30 days prior to loading a cask. However, the Technical

g Specification (TS) 4.19 requirement to perform a visual inspection of the ,

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lifting device (lift beam and extension) and verify operability of the device j 7 days prior to use, was not identified in D95.1. There was no procedure in 1 2

existence identifying actions required to verify operability of the lifting i

! device. This issue was identified by the inspectors. The inspectors verified l that D95.1 had been updated to include the preoperational testing requirements of TS 4.19. )

4 I In addition, the original procedure did not include a step to perform *

, radiation surveys of the cask surface before moving a cask to the ISFSI, as  :

required by TS 4.6.1 to ensure compliance with TS 3.6.1. The inspectors 1 discussed this issue with the licensee and verified that D95.1 was revised to i include specific steps for performing TS-required gamma and neutron dose rate l surveys.

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' The failure to incorporate the requirements of TS 4.19 into D95.1, to develop i a procedure identifying actions required to verify operability of the lifting ,

i device, and to include a step for performing radiation surveys of the cask

surface before moving a cask to the ISFSI as required by TS 4.6.1, are

, considered examples of a violation of Criterion V of Appendix B to 10 CFR Part j 50 (50-282/95002-01; 50-306/95002-01; 72-10/95002-01(DRP)). l 3.2.2 D95.2. "TN-40 Cask Unlordina Procedure" l

, The inspectors identified that the final revised and approved D95.2 unloading I

procedure did not adequately address the "S-reauirement to sample the spent

fuel pool for boron concentration. SpeciFically, TS 4.2.1.2 required

verification within four hours of flooding the cask cavity for unloading the i

' fuel assemblies, that.the dissolved borun concentration in the spent fuel pool- i j

• water introduced into the cask cavity was greater thari wr equal to 1800 ppm. i However, D95.2 required sampling four hours prior to lowering the cask in the l

! .. pool. The inspectors noted that there may be some time delay between -

! .. partially lowering the cask into the spent fuel poc1 and filling the cask. -

i */ /The subject TS requirement is important in that it increased the i.'

l,'. ' defense-in-depth for ensuring that..there was not the potential for an j ' .

inadvertent criticality. The inspectors verified that D95.2 was revised to j incorporate the TS requirement.

1

' The inspectors also verified that the D95.2 procedure contained specific steps for sampling the inner cask atmosphere to verify the integrity of the stored fuel. The inspectors noted that 095.2 did not contain a hold poir t +a ansure .

I work wniild nnt ennd nue until the sample resu ts nac oeen raviawec . The fY {

Tnspectors considered this procedural hold po' nt important to ensure that an A.d ,

unplanned and unmonitored release path would not be created while the cask was , 1 in the Auxiliary Building. The inspectors verified that D95.2 was revised to incorporate the subject hold point. The inspectors concluded that the final D95.2 procedure contained adequate guidance to ensure that the sampling and ot' tr. unloading evolutions were performed in a manner that would maintain l e .psures to workers as-low-as-reasonably-achievable. j l

The failure of D95.2 to adequately acdress the TS-requirement for samp11nn the i spent ruel pool ano to inciuae u noic point to ensure the results or the inner i cask volume sample naa Deen rev' ewed before allowing work to oroceed, are i

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, considered examples of a violation of Criterion V of Anendix B to 10 CFR Part 50 (50-z52/95002-01; 50-306/95002-01; 72-10/95002-01( m )).

1 3.3 Emeraency/Off-Normal Procedures

, The Part 72 license re 2d th: IPgee to develop an abnormal operating 4 .y procedure (AOP) fort uried cask event.yThe inspectors asked the licensee if

( any other emergency 7Ufi-nonnai piucednes were required in addition to alam I e response procedures and the buried cask A0P. 'The inspectors reviewed the cask '

handling procedures to determine if contingency actions for abnormal events had been addressed. .

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The licensee does not have any procedures, in addition to the buried cask, j which address off-normal events 2 The inspectors noted that step 5.0, of procedure D95.1, "TN-40 Cask Loading Procedure," stated that, "Should anything

, not look right during the performance of this procedure it is imperative that the issue be resolved prior to proceeding. All those involved in the i

perfomance of this procedure SHALL have their questions satisfactorily'

answered prior to having to perform their task." In addition, to this general

precaution, the inspectors noted that D95.1 contained specific " hold points" l at various steps in the procedure which required that the loading evolution be  ;

i stopped and any abnormal condition evaluated before proceeding. The . l

! inspectors did not have any further concerns with this issue.

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3.4 Conclusions j '

The licensee did not complete development of the loading and unloading procedures until the day following submission of Ine preoperationai ract i renort. suom1ssion of Inis report impliea Inat the licensee was ready to load a cask with soent fuel. While the inspectors recognized that finaitzing these j procedures was contingent upon completion of the preoperational testing i

evolution or " dry run" and the subsequent incorporation of any lessons l i learned, there were many aspects of the procedures which should have been in ,

place before the dry run. For example, Technical Specification requirements were not effectively incorporated into the loading and unloading procedures. -

i Assuming procedural adherence, the final procedures in place for cask handling l and loading were adequate to ensure that these evolutions would be conducted safely. -

i 4.0 Audit Reports. Source Insoections. and Vender Records

! 4.1. Agdit and Source Insoection Reoorts The inspectors reviewed a sample of the licensee's audit and source inspection reports to detemine if there were any issues that could affect the quality of the cask. This review included documentation pertaining to associated audit i findings. The inspectors also reviewed several fabrication records to verify I compliance 'with the design basis documents, including applicable industry 4 standards. The following documentation was reviewed:

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! . 6.6 Instrument Calibrations The inspectors reviewed the licensee's procedures for calibrating cask survey instruments and determined that the procedures were adequate to ensure proper i calibrations. The inspectors will observe instrument calibrations and survey j techniques during the actual cask loading evolution.

6,7 ISFSI Monitorina

• The inspectors walked-down the ISFSI facility and ensured that TS-required j thermoluminescent dosimeters were in place.

4 6.8 Radiation Protection (RP) Practices Durina Preonerational Testina l

, The inspectors observed RP practices during the loading dry run and noted that

! workers were kept informed of the radiological conditions and that RP - -

j personnel were prompt and thorough in perfoming dose rate surveys to monitor j changing radiological conditione. The inspectors also considered the

decontamination techniques used by the RP staff during the dry run adequate to i ensure Technical Specification limits for surface contamination of the cask

would not be exceeded.

i 6.9 Neutron Shield Performance The NRC issued a violation in NRC Inspection Report 72-0010/94-212(NMSS) for inadequate control of special processes pertaining to the neutron shield resin pour during cask fabrication. Specifically, the data record sheet associated with the resin pour procedure indicated that the temperature of the resin mix before adding the catalyst was 63 degrees Fahrenheit rather than between 68 and 70 degrees as required by the procedure. In response to this violation, ,

the licensee committed to perform a thorough survey of the cask following fuel l load to verify that the integrity of the neutron shield was not affected by the procedure deviation. The inspectors reviewed the licensee's plans for surveying the neutron shield and de.termined that the survey techniques were adequate to confirm that the neutron shiel'd was performing its design function.

6.10 Conclusions With the exception of the procedural content problems discussed in paragraph 3.2, the licensee developed and implemented an effective radiological controls program for monitoring cask loading and unloading activities and storage in the ISFSI. Cask handling procedures and associated RWPs ap addressed items such as dosimetry requirements for workers,propriately survey techniques and the use of calibrated instruments, required air sampling, protective clothing requirements, radiation ar.d contamination area postings, and procedural hold points and work stoppage criteria.

7.0 Pre-coerational Testino (Loadino and Unloadino Dry-Run)

'The NRC license for tho ISFSI required the licensee to conduct pre-operational testing to demonstrate cask handling capabilities before loading the first i

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4 cask with spent fuel. The inspectors observed and/or reviewed several pre-operational testing activities. These included: cask arrival and receipt inspection; transport vehicle pre-operational testing; cask transport to/from the ISFSI storage pad / Auxiliary Building; cask pressure monitoring system pre-operational testing; cask vacuum drying, helium backfill, and seal performance testing; fuel inspection; placement of'the cask in the spent fuel .

pool and simulated fuel loading; and cask removal from the spent fuel pool a#f ,

subsequent decontamination. [The removal of the cask lid under water and th6 filling of the cask with water were two evolutions that were not demonstra'te l oy tne iicensee auring dry run activities.' These exemptions were approve 4 l the HRC to prcyont any unnecessary damage to the lid seating surface duridst_ g.

the dry run and did not affect the licensee's ability to demonstrate unloading a was performed to demonstrate that the transporter and cask would not tip 3

over during cask transport should a seismic event occur. However, the subject l SE did not address the consequences of a tip-over accident in the Auxiliary Building rail bay. ,

1 l The inspectors discussed this issue with the licensee and with representatives

, from NMSS. Based on these discussions and the results of a previous analysis  :

i involving the loss of all cask confinement barriers during a spent fuel i shipping cask handling accident, the inspectors concluded that if a release of 1 i radioactivity occurred due to a tip-over event in the Auxiliary Building, the )

. releare would be substantially less than 10 CFR Pert 100 guidelines. Thus, the inspectors agreed with the licensee's "no" response to'the subject question. However, the documented basis was incomplete in that it did not address the consequences of a cask tip-over event within the Auxiliary Building. ,

i While the inspectors noted some weaknesses with the quality of SE No. 344, the .

inspectors determined that the licensee's conclusion that operation of an I i ISFSI would not create an unreviewed safety due to an adverse impact on l

! reactor plant operations, was valid. l

! 7.1 Seal performance Test '. I i

1 The ins)ectors reviewed the licensee's methodology for perfomance testing of I

. the cas( seals. The lid sealing system was designed with three sets of double  ;

j 0~ rings: one set on the circumference of the main lid and one set on the  !

fu nge covers for each of the vent and drain ports. The spaces between the 0-rings for the lid and each flange were interconnected via drilled channels

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4 to the overpressure (OP) port. The OP port was connected to the OP tank which j was designed to apply helium pressure to the volume of space between all of J the 0-rings. Should inner-seal leakage occur, helium would leak from the OP tank into the cask (because cask pressure was lower than OP tank pressure).

Should outer-seal leakage occur, helium would leak from the OP tank to the environment. OP tank aressure would be monitored on the storage pad and an alam generated if tan ( pressure was low. The pressure monitoring equipment was prepared and tested prior to cask transport. After cask placement, the pressure monitoring system would be installed on the cask and tested via a surveillance procedure. Completion of these activities was documented in D95.1.

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. ' 8.4 OA Overview of Dry Cask Storaae Activities

After receipt of the first TN-40 cask at the site, the inspectors determined 1 4

that the licensee did not have a comprehensive plan to inspect, audit, or i monitor dry cask activities onsite, in particular, those activities that i interface with the Part 50 license. The inspectors identified several issues j that should have been identified by the licensee. After discussion with the i i inspectors, the licensee developed an " Integrated Dry Cask QA Assessment s

, Plan," which provided direction for the Nuclear Quality Department in the

inspection, audit, and surveillance of dry cask storage activities. Once 1 established, the licensee'.s quality verification efforts were effective in 1

identifying issues with the dry cask storage project which required resolution f i by the line organization.

8.5 Retrievability -

j ,

! On May 3, 1995, the licensee submitted on the docket, correspondence that

, addressed the ability to unload the first TN-40 cask following completion of i the May 1995, Unit 2 refueling outage and prior to receipt of the second cask

onsite. The NRC's Office of Nuclear Material Safety and Safeguards responded i on May 5,1995 to the licensee and stated that the plans described in the j May 3 letter to address unanticipated unloading of a cask before another cask 1 i had been loaded, would allow ready retrieval of the spent fuel for further  ;

i processing or disposal as required by 10 CFR Part 72.122(1).  !

i ,

j 8.6 Exit Interview i i The inspectors met with the lic.ensee representatives denoted in paragraph 8.7 )

l during the inspection period and at the conclusion of the inspection on

{ April 28, 1995. The inspectors sumarized the scope and results of the i inspection, and discussed the likely content of this inspection report. The

licensee acknowledged the information and indicated that some of the j

information disclosed during the inspection could be considered proprietary in j nature.

! s 1 8.7 Fersons Contacted ]

l Northern States Power Comoany l  ;

l #E. Watz1, Vice President Nuclear Generation

fM. Wadley, Plant Manager i

! fK. Albrecht, General Superintendent, Engineering l G. Lenertz, General Superintendent, Maintenance

1. 0. Schuelke, General Superintendent, Radiation Protection i and Chemistry i J. Sorensen, General Superintendent, Plant Operations i J. Goldsmith, General Superintendent, Nuclear Generation Services Engineering j fT. Amundson, Director, Generation Quality Services i

1. P. Kaman, Generation Quality Services t fJ. Hill, Manager, Generation Quality Services l fJ. Bystrzycki, General Superintendent, Project Management l 29 .

l l

I

{ Atteclunant E Northem States Pswer Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, unnesota 55069 4

s May 3, 1995 10 CFR Part 72 RECEIVED i

MAY 0 91995 U S Nuclear Regulatory Commiss6n Attn: Document Control Desk .r ..s.At. CC  : :.'.

Washington, DC 20555 PRAIRIE IS1AND INDEPENDENT SPENT FUEL STORACE INSTALIATION Docket No. 72-10 Materials License No. SNM-2506 Information Related to Unloading of TN-40 Cask The attached information is provided in response to questions raised during the April 28, 1995 public meeting on the Prairie Island Independent Spent Fuel Storage Installation. The qu&stions were related to the ability of NSP to completely unload the first TN-40 cask following completion of the May 1995 Unit 2 refueling outage and prior to receipt of the second cask onsite. As shown in the attached assessment, NSP will have the capability to completely unload fuel from the first TN-40 cask back into the spent fuel pool in a timely manner, following the May 1995 Unit 2 refueling outage, if such action becomes necessary.  ;

We have made no new Nuclear Regulatory Commission commitments in this letter or the attachment. Please contact Gene Eckholt (612-388-1121) if you have any questions related to the information provided.

Me,J -

Roger O Anderson Director Licensing and Management Issues c: Director, Office of Nuclear Material Safety and Safeguards, NRC NMSS Project Manager, NRC Re51 onal Administrator - Region III, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silberg Prairie Island Independent Spent Fuel Storage Installation Service List

Attachment:

Assessment of Capability to Unioad TN-40 Cask Following May 1995 Unit 2 Refueling Outage p w

Page 1 of 2 i

ASSESSMENT OF CAPABILITY TO UNI 4AD TN-40 CASK j FOLLOVING MAY 1995 UNIT 2 REFUELING OUTAGE

,i j The Prairie Island spent fuel pool is designed and licensed to store 1386 fuel j assemblies. Due to inaccessible locations and other non fuel bearing s components, the practical storage capacity is normally considered 1362. Aft r

] the May 1995 Unit 2 refueling outage there will be a total of 1377 spent fusi

assemblies on site, 1337 in the pool and 40 in the first TN-40 cask. Using the practical storage capacity of 1362, this would leave 25 spaces available in the spent fuel pool that could be used for storage of spent fuel from a

cask. Thus, 15 additional pool locations would be required to completely

unload s TN-40 cask back into the spent fuel pool.

1 However, 15 of the non-fuel bearing components noted above can be temporarily j relocated as described below to provide the 40 pool locations required to 4

unload a TN-40 cask. These non fuel bearing components could be relocated by either of the following proc. esses; ,

1. Relocation to Temporary Pool Location:

1 Move non-fuel bearing components to a temporary location in the pool (most i likely the fuel transfer canal). A conceptual design of the ' hardware f

required for this temporary storage has been developed. We estimate the required hardware could be fabricated and the non-fu21 bearing components

relocated to their tempora,ry locations in approximicely 1 working week.

j This would adequately support any credible situation requiring cask unloading.

or, i

2. Relocation to TN-40 Cask: ,

i '.

! Even though the TN-40 cask being returned to the spent fuel pool may not be

qualified to hold spent fuel, it quite possibly could still safely hold i irradiated non fuel bearing components. If this is the case, as the TN-40 cask is being unloaded, the required non-fuel bearing components could be

, relocated, on a temporary basis, into the TN-40 cask. The cask would then 1

be removed from the pool until another cask is available to remove the

spent fuel. Following loading of the replacement cask, the non-fuel bearing components would be relocated back to the spent fuel pool.

l l Conceptually, the following basic operations would be required to perform these options:

i 4

Relocation to Temporary Pool location:

l 4

1. Bring cask back from the ISFSI into the plant Auxiliary Building.

2. Move cask into the spent fuel pool and remove lid.

I -

4 3

i.

J

@%MG Pags 2 of 2

3. Fcbricots ths h:rdwsro nicssacry to temporarily relocate non-fuel bearing components to the transfer canal.

i 3

4. Relocate non-fuel material to transfer canal.

5. Off-load fuel from the cask into spent fuel racks.

S j 6. Remove cask from the spent fuel pool.

7. Repair existing cask or pro' vide replacement cask.

l 8. Load the repaired or replacement cask with fuel and return it to the ISFSI.

9. Relocate the non-fuel bearfng components back into spent fuel racks.

l Relocation to TN-40 Cask:

1. Bring cask back from the ISFSI into the plant Auxiliary Building.

2. Hove cask into spent fuel pool and remove lid.

3. Remove fuel assemblies from the cask, place them back in the spent fuel racks, and relocate the required non-fuel bearing components into the cask.

4. Replace lid and remove cask from spent fuel pool.

5. Relocate fuel from the spe'nt fuel pool into a replacement cask.

6. Hove the loaded. replacement cask to the ISFSI.

7. Place the original cask back into the spent fuel pool.

8. Relocate the non-fuel bearing $omponents back into the spent fuel pool.

9. Remove the empty cask from the spent fuel pool and repair if possible. -

In summary, either of the options described above would allow NSP to -

completely unload fuel from the first TN-40 cask back into the spent fuel pool in a timely manner, following the May 1995 Unit 2 refueling outage, if such action becomes necessary.

i s .

MR. KUOLER: In torac of tho procedurco 'Ns lhctsolvo,s,theinopoctorshavofoundtheloadingproceduror i ,

1 o bs cccoptablo. Thoro aro a numbor of factoro that iicylify the preparation of loading procedures as compared io snicading procedures. During loading process you've iherectorized the-fuel; you know what condition it is in as 1

au put it-into the cask. Also you can take advantage of icoceno learned from other. licensees and from the dry runs 1 hot the licensee performs-on site.

For the unloading procedures, what we are finding . t lsthattheyaremorecomplexthantheloadingprocedures.

47

. ofertunately some of the older SARs fail to recognize this Iad tond to indicate that unloading is simply the reverse of

!ccding, which is not true. For one thing, licensees need io osnoider the potential condition of the fuel when they go io unload it. Depending on the situation, the fuel may have lCO2 in the cask for decades, and they need to evaluate the icadition of the fuel to the extent possible before they tart caloading it.

We do put an inert environment into these casks to irevent oxidation of the fuelt Assuming that that

cvironment has been maintained, the fuel should be in good jo2dition when they go to unload it, but they need to ivaluate.

! There are also issues associated with the lcflooding of the cask. During thn unloading process we 4cvo to refill the cask with water. There are some issues

coscisted with that such as cask pressurization duo'to

team generation as you put cold water onto the hot fuel.

'ico the consideration of any thermal shock to the fuel as es cre reflooding it, and also radiological protection for

ho corkers during that. phase, because you will be venting ho cack. Generally they are going to direct that venting

!ithertothepoolortoaventilationsystem,buttheyneed i o conoider that. -

1

{ In addition, there is essentially no cask l 48 iniseding experience for them to look back on for lessons carned. So they don't have that information available to

}hascocomparedtoloadingprocedures.

In addition to the working group activities, the taff has been putting increased emphasis on our inspection stivities in this area. The procedures for the recently oilt facilities have been inspected during the - '

recporational phase using the new inspection procedures het Bill Travers had mentioned. These inspections were a Gint effort between the regions, NRR and RMSS. We colecily pool our resources and our expertise'to perform beco inspections.

. We plan to cbatinue those inspections for all l cttre facilities. 1 We are also taking a look back at some of-the old i ceilities and looking at what. inspections have been  !

crformed there to determine whether we feel that we have i ccus:sted well enough that those procedures have been m p cted. If we determine that these older facilities were ' '

et well documented, we are going back and take a look at ,

.hans 40 well and do further inspections in those locations. '

That is all I planned to say on loading and

.ciccding. If there are no questions, I will turn it over e chov14 en talk ahont NRs etaff in4+4me4wa=

4

E LRLD1 \

NORTHERN STATES P3WER COMPANY MAINTENANCE PROCEDURES

$*hjf$$dk $ TITLE NUMBER:

. Ed s TN-40 D95.2

$@MP* h j' aa$ CASK UNLOADING PROCEDURE REY: 0 "dh ijjyf. f..Q, .,,, ."F g  ;... da Page 22 of 80 i

A;$'  ! ' , \ 7; While the Aux Building Crane is moving the cask into the i s: " ;"i ; r ;; / ' d' i Spent Fuel Pool, the crane switch will be in " CRITICAL" '

'"Q, , g',' $1T t '; position. In this condition, the crane will be unable to f i

• N'~: move more than 1 inch east or west once it passes the roof

.  %, , ,'i 7d: slot centerline and is within 6 feet either side of the "I' $[,l'hl J$','J //[; '; h enclosure.

• l C.17 Place the Aux Building Crane in the " CRITICAL" mode.

8.17.1 brn the key switch on the crane controls to the " CRITICAL"

, position.

i Rigger Date 8.17.2 Key switch on the crane controls verified in the " CRITICAL" position by a rigger different from the rigger who changed the key switch position.

Rigger Date 8.17.3 Place the key in the key cabinet in the Maintenance Supervisor's Office.

i 8.18 Within four (4) hours prior to }oading fuel, yerify SFP boron concentration is >2280 ppm. TwoTa~mples must be drawn and analyzed independently by two separate individuals. Record sample suits on the Cask Loading Report, Appendix A.

8.19 Operi the spent fuel pooi enclosures roof hatches.

8.20 Raise the cask to thn 755' level and position it over the pool.

Igure 1 8.21 Slowly lower the cask into the spent fuel pool whl!e spraying the cask and lift beam with demineralized water to provide a film of clean water on the cask surfaces.

j J

--l

^

2

. _ . . - - - .- -~ .-- -

~ ~

. _ . . . . . -.- - - - - . - - - ~ . - -

Attachmsnt H

• ISFSI SAR i .

i TABLE 5.1-2 -

ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR '

. CASK HANDLING OPERATIONS Ooeration No. of Time Avg. Distance '

Personnel Jain), (ft) from Cark Receiving

1. Unloading (A1) * * *

! 2. Inspection (A2 through A7) * *

• i 3. . Transfer to cask  ::

j loading pool (A8) * *

• Cask Imadine Pool

$ 4. Iower cask into pool (B1) * * *

5. Ioad fuel (B2 through B4) 5 * *

6. Place lid on cask (B5) 5 * *

7. - Lift cask to pool surface (B6) 5 30 5

8. Install lid bolts (B6) 5 120 3 j 9. Drain cavity (B7 through Bil) 5 90 6

10. Transfers to decontamination j area (B12) 3 60 10 .

i . .

{ Decontamination Area -

11. Decontaminate. cask (C1, C2) 3 120 3 ,

12. Remove vent plugs 2 30 5

{ 13. Drying, evacuating, l 1 backfilling (C3 through C13) 2 480 5 l

14. Install top neutron shield C14) 2 15 3

15. Install pressure . l l transducers (C15 through C17) 2 30 5 j 16. Pressurize interspace (C18) * * *

17. Check leakage (C19) 2 30 5 ,

t 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 O

TABLE 5.1-2 REV. 2 9/91

. l l

2

A

• ' ISFSI SAR l

/ ',

TABLE 5.1-2 (Continued) 1 ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR l CASK HANDLING OPERATIONS l Ooeration No. of Time Avg. Distance l Personnel IRID 1 (ft) from Cask 4 i

1 i

Storaae Area )

23. Unload from vehiclo position in location ~

(D1, D2, D3) 5 60 5

24. Check surface dose rate (D6) 5 30 3 j 25. Connect pressure

! instrumentation (D4, D5) 5 30 5 4

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 I

Maior Maintenance - l t (once in 20 years) l l .

j 1. Replace cask lid seals 3 1950** 8 i

i I

• No measurable dose associated with this activity. Therefore, the number l of personnel, tire and distance are not significant. I

! # 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.

1 4

i

}

?

TABLE 5.1-2 REV. 2 9/91 l

1 l

1

( JrEh* I N ' *

~

Attachment I 1

1 .;n.y. T '

n. siaciale ' . $

j ,,,,,,,,,. g, -

seUCLEAR IIEeulATolly CohandesS40N -)y

• ri 1 .

- = event .t point .eaa eccorred . na, n. ign. .ee m ilceasee st,m a

. '! spark ulth the antamated unidtag anchlee to hegle unidtag the shleid 114. a hydrogen here occurred that generated sofficient pressere la the task to L

. displace the tid semeuhat, leavlag it cocked at an angle. As le the case at Jose 14.1gg Eshisses, the cast had met been draised (eacept for a small area sear the top A of the cast to facilitate meldt ). After lavestigattag the situaties and I Br. Ilary Staclair detersielas a safe course of act es, the licensee straightened the lid, filled j

' gen't unste Akhlgan the air e at the top of the cast mIth mater, and saved 84 hack late the

• i C

P.S. See M82 spent 1. The licensee thee removed the shleid lid and usleaded the I i

[ Rearse. Illchlges 48161 fuel. The staff. lacludlag en Aegeanted Inspecales Isam maattered  !

Itcensee actlettles dorlog this event. The lavestigatlas of this eveet and "

I Sts laplicattens is centleslag. The RSC has issued confirmatory acties Seer Sr. Slac1str8 letters (Cats) to the llenesees for tee ledependent spent feel st{

lastallettens at Pelet teach. Palisades, and Arkansas lloclear One 1,essees le e letter to ested Decaster 38.1996 I stated that the statt ses met estag the vlC-24 cask). The Cats document the agreement of these Iceasees meerg of W. . Ret 1eer Rogalatory Cennissles (K) dry storage to refrale from leadlag er enloadles casks esitll after they have completed the .

casha that been unloaded ly M reacter llconsees. lace that letter, the actless discussed le the Cats and contacted the IIRC. l

! 59scamste Electrk Peuer Campasqf. the licensee for the Pelet Seach plaats.

enloaded e cost sitar e Bedrogen Sgnitten event that occurred thee the I trust t at tMs infernstlen ulli he helpful to yes. I apelegtze that the l

j licensee 5 lllated mold for He closure of the shleid lid, le addittee. I leformelles ce%tataed la er December 28 IgeS. letter was not completely e beso recenti learned of past cases le uhkh 1kessees have anleaded casks occurate. Is yee have any geestleas pItase contact se. i efter teentl lag problems the leadlag process. Is all three cases. t the Ikonsee met completed leedlag process and some of the effected ggggggy, cashs were moved beyond the '===*==leetles eree. The purpose of this letter la to provWe yee ulth leformattee related to these activltles. , .0RIGlag Stufe STs the first case occurred at ths Sorry Pomer 5t'tles a le Tleglela le movember Andrew J. Keeler. Project Rasager 1986. The Ikeesee hed Iceded the first Caster W/21 cask a destge that uses Pro t Directorate 118 3 lue helted lids ulth seal Flags. 1he laser lid had been Installed and the Olv sten of Iteacter Projects Ill/IT

• cask was draleed. het the seau rlag failed the hellem leak test reestred for Offica of Iloclear Beacter Regulatlee j t feel

this of closure. The Ikonses moved the cast heck te ne scs 7tr. glchard W. $ medley i peel. ended 94, and lesered it late the cent pit. lAhen the 14 una i removed, the tscensee deteralsed that the seal rt had shifted est of its Caesumers Feuer Company l groove des to hydrodynaalc forces as the Ild was aced este the cask. The

slconsee revised the leadlag precedure le leser lid more slemly and the See east parp I' q prehlen has set recurred.

OISTRIAGillM.

- 1he seceed case happened at the 11.3. Rohleten Reclear Plant le february 1999. Decket mes. 50-255, 72-F. 721007 Centesi files '

I The itcessee bed leaded the first Insupus-7P cash and anved it to the MetIC P033 8/T (IDO #665) j decentanlentles area, le preparetten for meldlag the laser (shield lid the ggamberaal EtennaugFlee j llconseo perfereed a servey of the redlettes esse rates above the 1 . Jees EAdensas licensee found dose rates that eere higher than these pedicted and decided to lesssalk Creamer.

asse the cast back to the spent feel pool and unleed it sa part of the LClark. OEC Kleu, let11 losestigettee of the dose rates. The cask had set yet been drataed asd se ptag. IBE55 VThape,18:5$ '

lssess essectsted ulth refleedlag a cast more act 1kshie. The licensee *'

detersleed that variatten le the dose rates uns et le the accuracy of the *S methods used for dose prodktles and mell ulthle the.Ilotta le the tachalcal e===m.. 90.C.tetu.li s se - RM he== C3DRT-CASK h e e- v - c. \$leCLAla

=a -s .it.P.R.E.V.Ious t.TE c v Concur.REI.IC.E m

specificatless. The Ilcessee une proceeded te lead de cast. e,, gg w ,o33 gg ,s ,on. gc oo:s,po. I or.C. Ic o: con ie GManes b as 0FesteMarswa aKeeler m_ 1 W amar i laterentles conceratag these tuo cases uns set utdely heeum ettkle the NBC, 04TE 184/l2/96 e6/12/M e6/13/96 06/le/M e6/g/WW

staff because the prehlems that led the liceasses to enleed De casts mere set OfHCIAL Bl;0RD COPV i safety-stgallkant. la heth cases the licensees feend the prehtens through

' appropriate testlag er maatterleg and took prompt, cesservattue corrective - 8

! actless. .

• The )lcansee for this facility has set yet Isaded any casks.

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