ML20140E472
| ML20140E472 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/28/1997 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML20140E450 | List: |
| References | |
| 2.206, DD-97-18, NUDOCS 9706120071 | |
| Download: ML20140E472 (29) | |
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i BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION i
IN THE MATTER OF:
Docket 72.
THE PRAIRIE ISLAND INDIAN COMMUNITY
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j Petitioners
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UNITED STATES NUCLEAR REGULATORY COMMISSION )
l Respandent Petition I
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4 Pursuant to 10 CFR Part 2.206 of the Commission's regulations, the Prairie Island Indian l
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 2
Storage Installation (ISFSI) prior to establishing conditions for safely unloading the TN-40 dry storage containers.
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- 2) Suspend Materials License No. SNM-2506 for cause under 10 CFR 50.100 until such i
time as all significant issues in the unloading process, as described herein, have been resolved, the unloading process has been demonstrated, and until an independent third j
party review of the TN-40 unloading procedure has been conducted.
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- 3) Provide petitioners an opportunity to participate fully in the reviewing the unloading i
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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9706120071 970028 PDR ADOCK 05000282 i
G PDR l
s Facts
- l. - Northern States Power (NSP) owns and operates the Prairie Island Nuclear Generating Plant on Prairie Island in Minnesota.
- 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 and: executive departments, including the NRC, have a Trust Responsibility towr.rds Indian tribes. The United States Government's " trust responsibility" toward American Indians, which is the unique fiduciary and legal duty of the United States to assist Indians in the protection of their property and rights. This trust 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, i
the President's memorandum also stated that each executive department and agency sht.ll consult with tribal governments prior to taking actions that affect tribal governments and 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 10, Part 72 of the Code of Federal Regulations (10 CFR 72). The application was assigned Docket #72-10. Included in the application were Technical Specifications and 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 F).
- 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.n 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 prevented, cask integrity is maintained, and fuel is not damaged so as to preclude its 2
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ialtimate removal from the cask" (TSSAR, p. 3.2-1) and through a Decommissioning Plan 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 cask. With respect to deconunissioning, there is just mention of removing waste from the TN-40 and shipping it in a licensed transportation cask.
- 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 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). The 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. The procedure described by NSP is as follows:
" 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 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 taken in Table 5.1-1 of the SAR, only in reverse. (Attachment A).
- 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 the SAR included in NSP's application. The NRC found that no significant impacts from the construction of the ISFSI were to be expected. No impacts were expected from the operation of the ISFSI either.
With respect to radiological impacts, NRC staff expected that impacts from cask loating and preparction 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 expected radiological occupational exposure during cask unloading.
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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.
- 13. On October 19,1993, NRC issued Materials License No. SNM-2506 to NSP fcr the ISFSI. Under the terms of the license, NSP is authorized to receive, possess, store, and transfer Prairie Island Nuclear Generating Plant spent fuel at the ISFSI in up to 48 TN-40 casks for a period of twenty years.
Along with the license, the NRC transmitted a Safety Evaluation Report (SER), which is an evaluation and review of the:TSSAR submitted by the licensee. The review of the 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, minimum cooling, etc.), protection against environmental conditions, natural phenomena, confinement barriers (i.e., protection of fuel cladding to ensure that degradation and gross rupture do not occur over the design life of the ISFSI),
and criticality control (NRC regulations require that the spent fuel handling, transfer, and storage system be designed to 1 e maintained subcritical (10 CFR 72.124(a)).
l 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, thennal shock, flash steam, etc.) and potential exposure to workers, j
- 14. Within the ISFSI license (SNM-2506), a number of preoperational license conditions were specified:
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A training exercise (dry run) of all TN-40 loading and handling activities must be conducted and shall include the following (but is not limited to);
- a. Moving cask in and out of spent fuel pool area
- b. Loading fuel assembly (using a dummy assembly)
- c. Cask drying, sealing, and cover gas backfilling operations
- d. Moving cask to, and placing it on, the storage pad
- e. Returning the cask to the auxiliary building
- f. Unloading the cask
- g. Decontaminating the cask
- h. All dry run activities shall be done using written procedures I. The 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 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 to allow the NRC 30 days to review and assess the licensee's ability to load (or unload) a 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 intemally test certain steps in the unloading procedure (D95.2) in February 1995 (Note: the work ordent 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 Building,8.11 requires the removal of the lid).
In a later letter to NSP, however, the NRC noted that they did not require that the lid be 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 Prairie Island Indian Community staff that they (the NRC) have no formal mechanism to approve or disapprove loading or unloading 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.
- 18. 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 unloading of a TN-40 cask. In their letter, NSP stated that they could unload a cask "if such action becomes necessary"(Attachment E).
The 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
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 granted NSP the 10 CFR 72.82(e) exemption. In the end, NSP waited 20 days befort they loaded their first casks, instead of the requested 3 days.
- 22. On May 12,1995, the NRC approved the preoperational test report and authorized 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 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 until the day following the submission of the preonerational test report (emphasis added). Submission of this report [to the NRC]
implied that the licensee w as ready to load a cask with spent fuel and subsequently unload the cask, if necessary."
In that same Inspection Report, the NRC issued NSP a Notice of Violation (NOV) for not including certain technical specifications in their unloading procedure.
The letter specifically cited NSP for omissions in the unloading plan and "overall poor planning for dry cask storage activities." The violations with regard to the unloading plan were 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.
- 24. On May 30,1996, at a briefing for the commissioners of the NRC, Andrew Kugler, q
Lead Project Manager, Dry Cask Storage, NRC, stated while the NRC has found dry cask loading procedures to be acceptable, the unloading procedures are far more complex (Attachment F). In loading a cask, the fuel has been characterized (i.e., its integrity verified), a loading dry run has been performed, and that licensees can take advantage of lessons leamed from other licensees (in loading their casks).
Mr. Kugler stated that the older SAR's [ Safety Analysis Reports] do not recognize this complexity and indicate that unloading would be the reverse ofloading, which Kugler stated was not true. The unloading procedure in NSP's most recent SAR, and the SER for the ISFSI, is described as the reverse of unloading.
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Of particular concern with respect to unloading, stated Mr. Kugler, are the potential condition of the fuel and issues i.ssociated with the reflooding of the cask with spent pool water: cask pressurization due to steam generation as colder spent fuel pool water is placed into the cask. Thennal shock to the fuel during reflooding and radiological exposure to workers during the operation from the venting the cask (venting will either be done directly to the pool or the ventilation system).
- 25. On November 8,1996, NSP submitted to the NRC a revised TN-40 unloading plan.
- 26. The SAR for the Prairie Island ISFSI states that the fuel cladding may be as hot as 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 when the water hits the cavity surface. Although the SAR recommended pumping cold water into the cask prior to immersion in the spent fuel pool, the TN-40 unloading I
procedure contains no such provision (see Attachment G).
- 27. The Prairie Island SAR also states that the fuel assemblies should be inspected for any physical damage which could potentially cause problems during removal from cask.
The actual unloading procedure, however, contains no such requirement. The procedure only requires that the location of three of the forty assemblies be confirmed prior to off-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, 1996, the potential condition of the spent fuel, with respect to reflooding, is a concern.
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- 28. The SAR only discussed potential occupational exposure to radiation with respect to receipt of cask, cask loading, decontamination, removal to storage area, and periodic 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) bemg
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exposed to the much cooler spent fuel pool water (110 'F).
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- 29. Mr. Kugler stated on May 30,1996 that there was " essentially no cask unloading experience" for licensees to look back on for lessons leamed, like there is for cask j
loading. In a later letter to Dr. Mary Sinclair of Don't Waste Michigan, Mr. Kugler stated l
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 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 i
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Petitioners Claim l
- 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.
10 CFR 72.122(1) provides that:
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 72122(1) is that the storage system (i.e., the cask ) must be designed to allow ready retrieval (i.e., unloadind of spent fuel, not whether the spent fuel pool can acconunodate 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 after it has been sitting on the storage pad for a number of years. Thus the NRC requirement that storage systems be 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 after it has been in use.
The question of"retrievability" was discussed at a public exit interview convened by the 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 off-loading the first TN-40 cask. The NRC may recall that NSP's spent fuel pool was already full and there was a possibility that, after 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 response to these concerns, NSP assured the NRC that they would indeed have enough 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 fifteen 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.
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.
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 preoperational test report. Submission of this report implied that the licensee was ready to load a cask with spent fuel and subsequently unload the cask, if necessary."
How could NSP have claimed 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 8
' 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 will be unloaded should something go wrong with one of the casks? Many tribal members are fully aware of the dry cask situation at the Palisades plant (i.e., the problem 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. There are no other plants currently employing a cask designed by Transnuclear Inc. The NRC has licensed the TN-24, but it is not in use anywhere in the country. In affidavit to the Minnesota Court of Appeals, Mr. Jon Kapitz, NSP's Project Manager for Dry Cask Storage, stated that a TN-24P storage cask was successfully unloaded as part of a project l
sponsored by the Department of Energy (DOE) and the Electric Power Research Institute i
(EPRI). (Note: this Affidavit was submitted in response to legal action initiated by the l
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. The i
purpose of this research was to determine the thermal, shielding, and operational t
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 i
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 nonirradiated fuel (i.e., dummy assemblies).
The 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 does gLot 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 fu'l evaluation of the unloading procedure. Both the SAR and SER for the ISFSI stated that unloading a cask is the reverse of loading 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 of cask loading.
t As stated above, NSP had not finished testing its own loading and unloading procedure before submitting its preoperational test report to the NRC, thereby declaring that they were ready to load, and unload a cask if necessary. According to the actual loading and unloading plan and preoperational test results, only part of the unloading procedure has l
been tested (refer to Fact No.16). How can the Prairie Island hdian Community be assured that a cask can be unloaded if the procedure has never been fully tested before and it has never been done?
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- 3. As stated above, NSP submitted a revised unloading procedure to NRC in November i
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 i
potential unloading problems that have not been addressed in NSP's unloading plan. Our l
analysis of the unloading plan identified several deficiencies:
a) Failed Fuel Considerations Although the revised plan contains a step to sample the i
internal 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.
i If the internal gas sampling indicates failed fuel, procedures should be developed to recover the fuel from the cask without contaminating the spent fuel pool. As the i
unloading procedure is currently written, the dry cesk is placed in the pool for unloading, l
regardless of the radioactive concentration in the internal gas.
This action may l
contaminate the fuel pool with microscopic irradiated fuel particles (fuel fleas), thereby creating a major radioactive hazard. (This has occurred at Southern California Edison's l
San Onofre 3 reactor).
There are no procedures in place to determine whether fuel in a TN-40 cask has gross
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cladding defects. During storage, the cladding is under pressure and may crack, due to l
the high internal temperature of the cask (up to 644 F). Over time, the cracks can become larger. This is a cumulative process; the longer fuel is in storage, the greater the likelihood of cladding degradation. If fuel with cladding defects is unloaded, the spent fuel pool may become contaminated and workers may be exposed to radiation. Plant j
personnel will have to decide whether or how fuel with cladding defects will be removed.
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b) Ventine of Radioactive Gases The unloading procedure does not indicate whether or how radioactive gases might be vented from the cask, if concentrations are greater than
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certain limits. Plant personnel will have to decide how to vent the cask. If the l
radioactive air is to be vented, is a permit required?
c) Radiation Monitors Before venting the cask, a stop-check must be instituted to verify j
that ventilation systems and radiation monitors are functioning. There is no such stop-t check in NSP's unloading plan.
I d) Steam Bui!d-Up When water is pumped into the TN-40, it is very likely that steam will be created and pressure will build within the cask if the fuel cladding is greater than f
212'F. Although NSP has included steps to cool down the cask and fuel cladding, it is i
not clear what the maximum cladding temperature will be prior to the addition of water j
into the cask. The steam overflow from the cask, which is directed into the spmt fuel i
pool, i:, likely to be very hot. Wamings need to be posted regarding this hazard.
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The pressure build-up also places certain stresses on the hose couplings and piping-this i
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 4
10
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'nsure that pressure not build up too quickly.
The thermometer measuring the e
temperatere should have a range greater than 50 F to 300 F; the maximum temperature should be 900 'F.
The pressure gage should also be changed to one with a maximum l
pressure of 50 psig.
- 4. The 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.
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L Conclusion The public relies on the NRC to make decisions that these containers are safe, based on a complete evaluation.
Neither the NRC nor NSP have fully evaluated the TN-40 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
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SER implied that unloading a cask was very straightforward and just the reverse of l
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 concems with regard to unloading a cask.
NSP has not adequately demonstrated their ability te unload a TN-40 dry cask.
j 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 TN-40 unloading i
procedure to determine whether the issues raised by their own stafT and within this i
petition have been fully evaluated and included in the unloading plan for a TN-40 cask.
I i1
Appendix of Attachments Petition of the Prairie Island Indian Community RE: Northem States Power procedure to unload a TN-40 dry cask Attachment A Table 5.1-1 of the Safety Analysis Report (SAR) and July 10,1992 letter (NSP to NRC)
I Attachment B l
l Table 5.3 of Environmental assessment (EA) for the ISFSI i
Attachment C l
Table 6.1 of the EA
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Attachment D l
June 30,1995 letter from Greenman (NRC) to Watzl (NSP) and Notice of Violation Relevant pages only t
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Attachment E May 3,1995 letter to NRC from NSP regarding cask unloading l
Attachment F Transcript from May 30,1996 Commissioners meeting i
Relevant pages only l
Attachment G Step 8.21 of NSP's unloading plan i
AttachmentJi Table 5.1-2 of the SAR for the ISFSI i
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- Attachment I i
June 18,1996 letter to Dr. Mary Sinclair from Andrew Kugler 4
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e ;.1.,q; " v=...',8'Y.i 3
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flease tentact us if you have say questlena with respect to the attached
'a
. responses.
j a
a.5 ar.
a Parker Nanager Eksclear $wpport Servlees si 9tre..or. Offlea of Nuclear Material Safety and safeguards. ERC kNS$ Freject Manager. NRC Eaglemal Admintatrator Etegion lit. ERC Senter testdent leapector. ERC -
Lawrence Lleeraere Battenal laboratory
?
~
J E Silberg Pretrie Island independent spent Fuel Storage lastallatten Service 8.lst
Attachment:
' Response to HRC Questions.ie Ts.si Cask Thermal Analysts and on Precedure for i
Dateading Caak
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plF+T*<panet(toestten.
the leertta' leads free the seeldent end drop.
- g uost=tt of 1s. ore pet (0.a2 sus bend on che temperature.f no r at as war.t u -
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r kdpeettien the feet asseebilee.' ' V QP'*.*.
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es eatentate the basket penet stressee due se dif(eventsat thernet mapeneten. "
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V from that thernet anstyete were mood directly to the AlttTS etsiertural.medste e,
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$p N Desertbe the procesbare for eask. unloading prier to decommisstening.
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esp meten of the 30t statalese. the aluntons and beret fsee Seetten AS'of the -
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to estentate 'the mentanum sheer stresses at the-1/2* pits selds. It is 15F51 Safety Amstyete Deport for det'atted A5lSTS madet desertpttene).<.,tn order
- genserete the 1.3 te. diameter hele to the aluminum (and betal) plates are not.
" '.D
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- Amauming that the opent fpel tag the TW.40 emek will be trameterree to a alvely meewmed that the 1.38 Ja. diame' ve statalese plege that,
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- lleensed transport ensk thteg a mermal' *te poet
- fuel transfer. the seg=ence e
3
. ef operettena dieemssed to Seettee Lt of the 15F31 lafety Analyets Reperr.
i eentered, the ptogs are
- to be in dentest taittatty tot 70*F) with the a
D
. end in partiestre listed in Table 1.11. will be essentially performed in l
7 eppeetsg eldes of the two holes to the atentama fthe 'stees teuerd the center.
d
. reverse..
/
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4 5
'of the penet) se that the mentam interfetenee et a}untsume and steel witt.
5
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u eeeer when the passt is beated. In this wetst plug stealtament seee with the.
f,[ '
- The eask will be owed fres,The weather sever will be unbetted and reeeved.the ISFil back t ht$eet temperature (530*F).,seen by any portten of the beaket, the weld. sheer -
' s Bay using the ' transporter.
etrees eente reach a santeus of 23.434 pet as shown en Figure 43,.4 6 of the
- The *NPressere eyeten will thee be reeeved and the cavity goe asepted 1
[
iSF51 Safety Anatysta Report.
throug!L the vent port..
t
, & futt length eespartment well (160 ted toeg) with a span length of R:0$ in.
.'
- 8etressortsed and the task levered tate the spent fuel poet. With the cesL.
After eeving the cask late the fuel peel area. 'the envity will be te evaluated for sheer stresses.at 1/2* plus welde due to a Soc end drop; e
. tid at the poet eerface, fitt ans drale Ilmes wilt be eennected to the 18d '
Itze of mold. O P dia.
!i
- ' drein 'and went porte.. Derated weter will be slowly added to fall the east and to gradually eset the fuel in the eask. When the eask to fell, the ffl1 and'
/, drain Itnee will be removed. ' The east will then.be towered to the poet bettee Sunber of wtds 2 s 2 m 1H = to (each 8* speetag hae 'two rows of plege.
'e, 8
each ples has two welds ene en each olde) where the 18d would be reeeved making the feet seeessible for transfer.
. j *t.1s's to = 13.11 trart.eet Shear area
'rt. of 'aluutnm,. 2 m t.02 a 0.2S a 160 s.0.50'S - 47.42 lbs.
- f,
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- t. of berat - t a 7.30 a 0.075 a 140 a 0.0903 = 4.11 lbs.
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e ISFSI SAR TABLE 5.1-1 SEQUENCE OF OPERATIONS A.
ReceiviDE 1.
Unioad empty cask and separately packaged seals at plant site.
2.
Inspect the following for shipping damage:
exterior surfaces, sealing surfaces, trunnions, seals, accessible interior surfages and basket assembly, bolts, bolt holes and threads, neutron shield vents.
3.
Remove weather shield and install plug in neutron shield vent hole.
l (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.
6.
Obtain lid and lid seal from storage.
7.
Attach lid seal to lid by means of six retaining screws.
8.
Nove to spent fuel pool area.
B.
Soent Fuel Fool Area i
1.
Lower cask into cask loading pool.
2.
load preselected spent fuel assemblies into the 40 basket compartments.
3.
Verify identity of the fuel assemblies loaded into the cask.
4.
Remove protective plate from cask body flange 5.
lower lid and place on cask body flange over the two alignment pins.
6.
Lift cask to surface of pool and install lid bolts.
7.
Connect drain line to quick-disconnect coupling in the drain port.
. Bolt special adapter, with quick disconnect coupling, to vent port bolt 8.
holes.
9.
Connect plant compressed air line to special adapter quick-disconnect coupling.
10.
Pressurize cavity to force water from cavity through drain port to the spent fuel pool.
.f
~
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REV. 2 9/91 TABLE 5.1'1
..,v.-.
._._=_._.._..___._.._.___.__.._.m._.
s ISFSI SAR TABLE 5.1-1 (Continued)
SEQUENCE OF OPERATIONS 11.
Disconnect plant compressed air line and drain line from their quick-disconnect couplings, 1
l 12.
Move cask to the decontamination area.
C.
Decontamination Area (Rail Bav) 1.
Decontaminate cask until acceptable surface dose levels are obtained.
I 2.
Torque lid bolts using the prescribed procedure.
3.
Remove plug from tieutron shield vent and install pressure relief valve.
4.
Connect Vacuum Drying System (VDS) to vent port.
5.
Evacuate cavity to remove remaining moisture using prescribed procedure.
6.
Break vacuum by closing vacuum valve and opening air valve to admit dry air into the cavity.
1 7.
Disconnect VDS at vent port and install vent port cover with seal and bolts.
3 8.
Connect Vacuum-Backfill System (VBS) to quick-disconnect coupling in the 1
i drain port.
9.
Evacuate cavity to 10 millibar and backfill with dry helium gas.
)
10.
Pressurize cavity to about 2 ata with helium.
s 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.
I
- 13. Remove overpressure port cover.
14.
Install top neutron shield drum.
Torque the colts using prescribed procedure.
I 15.
I 16.. Pressurize overpressure system, (seal interspaces), with Helium to a pressure of about 5.5 sta.
17.
Perform leak test on overpressure system.
O M.g ; ; ' j;g TAB 12 5.1 1,.
REV.'2 9/91 g
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ISFSZ SAR TABLE 5.1-1 (Continued)
SEQUENCE OF OPERATIONS 18.
Check external surface temperatures using an optical pyrometer.
19.
Check surface radiation levels.
20.
Install protective cover with seal and bolts.
(cculd be performed at storage area) 21.
Load cask on transport vehicle.
- 22. Move cask to Storage Are's.
D.
Storare Area 1.
Unload cask from transport vehicle.
2.
Position cask in preselected location en storage pad.
3.
Check for surface delects.
4.
Connect pressure instrumentation to cask and to monitoring panel.
l
[
5.
Check that pressure instrumentation is functioning.
6.
Check surface radiation levels.
e TABLE 5.1-1 REV. 2 9/91
- -. -. ~ - - ~. -. - - - - - -..
Attachment B j
Concrete Pad i
The storage casks will be stored in two parallel rows of 12 casks on each of two 216-foot i
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 performance and earthquake safety, 3 feet of soil beneath each slab will be excavated and replaced with compacted ~ structural fill.
The pad elevation will be 693 feet 6 inches above mesn sea level (msl) to preclude immer-j sion of the cask seals during the probable maximum flood. They will be surrounded by a 17-foot high earthen berm.
i 5.3 ISESI OPERATIONS i
l Fuel handling and cask loading operations in the Auxiliary Building will be done in
{
accordance with requirements of the Prairie Island Nuclear Generatmg Pent 10 CFR Part 50 Operating Licenses DPR-42 (Unit #1) and DPR-60 (Unit #2). Cask transport and storage at i
the ISFSI will be subject to requirements of the Prairie Island ISFSI 10 CFR Part 72 4
i License. The major steps associated with the placing of fuel in the Prairie Island ISFSI are t
presented in Table 5.3.
s TABLESJ ISESI OPERATIONAL STEPS E
i A. RECEIVING s
1.
Unload empty cask and separaly packaged seals at plant site.
i i
2.
Inspect the following for shipping damage: exterior surfaces, sealing surfaces, trun-nions, seals, accessible interior surfaces and basket assembly, bolts, bolt holes and i
threads, neutron shield vents.
i i
22 i
i k
E
. = - -
- 3. - Remove weather shield and install plug in neutron shield vent hole.
i l
4.
Remove lid bolts and lid.
)'
5.
Install protective plate over cask body sealing area.
6.
Obtain lid and lid seal from storage.
'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.
IAwer cask into cask loading pool.
i 2.
Ioad preselected spent fuel assemblies into the 40 basket compartments.
i 3.
Verify identity of the fuel assemblies loaded into the cask. -
4.
Remove protective plate from cask body flange.
5.
Iower lid and place on cask body flange over the two alignment pins.
6.
Ilft cask to surface of pool and install lid bolts.
7.
Connect drain line to quick-disconnect coupling in the drain port.
8.
Bolt special adapter, with quick di== rect coupling, to vent port bolt holes.
23
9.
Connect plant compressed air line to special adapter quick-disconnect coupling.
- 10. Pressurize cavity to force water from cavity through drain port to the spent fuel pool.
- 11. Disconnect plant compressed air line and drain line from their quick-disconnect couplings.
- 12. Move cask to the decontamination area.
C. DECONTAMINATION AREA (RAIL BAY) 1.
Decontaminate cask until acceptable surface contamination levels are obtained.
2.
Torque lid bolts using the prescribed procedure.
3.
Remove plug from neutron shield vent and install pressure relief valve.
4.
Connect Vacuum Drying System (VDS) to vent port.
5.
Evacuate cavity to remove remaining moisture using prescribed procedure.
6.
Break vacuum by closing vacuum valve and opening air valve to admit dry air into the cavity.
7.
Disconnect VDS at vent port and install vent port cover with seal and bolts.
8.
Connect Vacuum-Backfill System (VBS) to quick-disconnect coupling in the drain Port.
24
9.
Evacuate cavity to 10 millibar and backfill with dry helium gas.
- 10. Pressurize cavity to about 2 ATM with helium.
- 11. Disconnect VBS at the drain port quick-connect coupling and install drain port cover with seal and bolts.
- 12. Perform helium leak test of lid seals.
- 13. Remove over pressure port cover.
- 14. Install top neutron shield drum.
- 15. Torque the bolti using prescribed procedure.
- 16. Pressurize over pressure system with Helium to a pressure of about 5.5 ATM.
- 17. Perform leak test on over pressure system.
- 18. Check external surface temperatures using an optical pyrometer.
- 19. Check surface radiation levels, i
- 20. Install protective cover with seal and bolts.
- 21. Load cask on transport vehicle.
- 22. Move cask to Storage Area.
25
.p-
=
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D. STORAGE AREA i
i j
1.
Unload cask from transport vehicle.
l.
2.
Position cask in preselected location on storage pad.
3.
Check for surface defects.
4.
Connect pressure instrumentation to cask and to monitoring panel.
5.
Check that pressure instrumentation is functioning.
6.
Check surface radiation levels.
The administrative 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. Before startup and during the lifetime of tim ISFSI, the cask monitoring instrumentation, the electrical system, the communications system, and the storage casks will be tested to ensure their proper functioning. The 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 training in specific areas as required by the Radiation. Protection program in effect at the Prairie Island Nuclear Generating Plant.
26 e gem o 34
. ~.. _. _.. _ _.. _. _ _ _. _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _. _. _ _.
____.___..m...
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Attachment C I
I TABLE fl a
l i
DESIGN BASIS OCCUPATIONAL QHE IIME EXPOSURES DURING CASK LOADING. TRANSPORT ANIl EMPLACEMENT 4
P Task Time No. of Dose Rate Dose Required (hr) persons (mrem /hr) (Person-rem)
?,
Placement in pool 2 2
3 5.0 0.03 1
l Loading process 5
5 5.0 0.125 i
Removal from pool 5
5 30.0 0.75 i
Transfer to decontamination area 1
3 30.0 0.09 Processing of cask 6.5 2
30.0 0.39 l
t j
Helium leak test 2
2 30.0 0.12 i
Decontamination 2
3 30.0 0.18 i
l Install neutron shield, j
pressurize, test -
3 2
30.0 0.18 l
Preparation for transport 1
3
-30.0 0.09 Transfer of cask to ISFSI 1
3 20.0 0.06 l
Final cask emplacement 2
5 30.0 0.30 TOTAL 2.315 i
4 i
i 1
1 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.
' Steps from Table 5.3.
4 i
34 1
1.
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- %g UNITED STATES j
g g
NUCLEAR REGULATORY COMMISSION e
o REGION til E
801 WARRENVILLE ROAD l
k, USLE, ILLINOIS 60532-4351 l
June 30, 1995 i
j Mr. E. Watzl, Vice President Nuclear Generation Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401
Dear Mr. Watz1:
)
This refers to the special NRC inspection from January 24 through 4
May 11, 1995, of dry cask storage activities at the Prairie Island site.
This inspection was conducted by the resident inspectors, selected RIII based j
inspectors, and technical staff from the Office of Nuclear Reactor Regulation and the Office of Nuclear Materials Saf~ ty and Safeguards. The purpose of e
this inspection was to evaluate the acceptability of the as-built TN-40 cask l
and to assess your performance relative to dry cask storage including the i
nrecoerational testina activities.
1 l
We discussed the results of his inspection with you and other members of your staff at a public exit meeting on April 28, 1995. At that meeting we identified five items that required further resolution.
You provided us with 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 i
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.
j On May 12 you loaded the first cask with spent fuel.
The enclosed copy of our inspection report identifies areas examined during j
the inspection. Within these areas, the inspection consisted of a selective i
examination of procedures and representative records, observations, and interviews with personnel.
l l
Based on the results of this inspection, we concluded that you were ready to j
safely load spent fuel into the TN-40 dry storage cask and transport this cask i
to the onsite ISFSI. We also did not identify any safety concerns'with the i
subject cask. However, one violation of NRC requirements was identified during the course of this inspection, as specified in the enclosed Notice of Violation (Notice). This violation pertained to cask handling, loading, and unloading activities that were not prescribed by procedures of a type s
4' j
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 5
}
this violation had been corrected and those actions were reviewed during this inspection. Therefore, no response with respect to this violation is required. However, we are disappointed that NRC inspectors, rather than your own staff, identified these procedural deficiencies.
5 e
_ _ _ _. _. _ _... _ _ _... _ _ _. _.. _ _ _. _ _ _. _ _ ~. _. _ _ _ _.. _.
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 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. 7 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 improvement plan documenting the specific corrective actions you have already taken and those you plan to implement to address the above weaknesses in dry cask activities.
Please respond to this request within 30 days of the date of this inspection report. We will continue to evaluate the effectiveness of your corrective actions to improve your performance in dry cask activities during future NRC inspections.
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 NRC Public Document Room.
The response requested by this letter is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
We will gladly discuss any questions you have concerning this inspection.
Sincerely,
{A. %)
Edward G. Greenman Senior Oversight Manager Region III Dry Cask Activities Docket No. 50-282 Docket No. 50-306 Docket No. 72-10
Enclosures:
1.
Inspection Report No. 50-282/95002; 50-306/95002; 72-10/95002(DRP)
See Attached Distribution i
E. Watzl '
Distribution:
cc w/ encl:
Site General Manager, PINGP John W. Ferman, Ph.D.,
Nuclear Engineer, MPCA State Liaison Officer, State.
]
of Minnesota State Liaison Officer, State of Wisconsin Tribal Council Prairie Island Dikota Community 4
J
~
4 e
i Y
a i
NOTICE OF VIOLATION Northern States Power Company Dockets No. 50-282; 50-306; 72-10 Prairie Island Nuclear Plant Licenses No. DPR-42; DPR-60; SNM-2506 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 Statement of Policy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below:
10 CFR Part 72.142(b) requires 3 licensee to establish, maintain, and execute i
a quality assurance (QA) prograrh with regard to an Independent Spent Fuel Storage Installation (ISFSI) that satisfies each of the applicable criteria of Subpart G, " Quality Assurance."
In meeting the Part 72.142(b) requirement, 10 CFR Part 72.142(d) accepts a Commission-approved quality assurance program which satisfies the applicable criteria of Appendix B to 10 CFR Part 50. As such, the ISFSI Safety Analysis Report ' states that the previously annroved Northern States Power QA program which satisfies applicable criteria of 10 CFR Part 50, Appendix B,'will be applied to activities, structures, systems, and components of the ISFSI commensurate with tnetr 1mpT>rtance to safety.
Criterion V of Appendix B to 10 CFR Part 50 requires that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and that these activities be accomplished in accordance with the associated instructions, procedures, or drawings. Cask handling, loading, and unloading are activities affecting quality.
pc trary to the above, cask handlina loadino. and uninaM na activities were 1
thot mroscrinne nv annrovec
)rocedurec nf a tvna inoropriate to the
~
W cumstances; as evidencec )y the(following examples: y 1
1.
. Surveillance Procedure. SP 1077, "Special Lift Fixture for the TN-40 Cask," did not address dimensional checks of the special lifting device, q
as required.
2.
Surveillance Procedure, SP 1075, "TN-40 Fuel Selection and Identification," did not incorporate the requirement of Technical Specification (TS) 4.1.2, which states that "before inserting a spent fuel assembly into a cask..., the identity of each fuel assembly shall be independently verified and occumented."
3.
Procedure D95.1, "TN-40 Cask Loading Procedure," specified in the prerequisites section that SP 1077 be performe@ays prior to loading a cask. However, the TS 4.19-requirement to perTorm a visual inenaction of the lifting device (lift beam and extension) ardrverify operability) of the device 7 days prior to use, was not identiftea in un.i.
mere also was no procedure identifying actions required to verify operability of the lifting device.
4 9
4
4.
Procedure D95.1, "TN-40 Cask Loadina Procedure," did not include a step to perform radiation curvaye of the cask surtace before moving a cask to tne 1stdl, as required by TS 4.6.1.
5.
Procedure 095.2, "TN-40 Cask Unloading Procedure," did not adeauately address the TS-requirement to samnie tne spent tuei pool or boron concentration within tour hours of flooding the cask cavi;y for unloading the ruei assemoiles.
6.
Procedure 095.2, "TN-40 Cask Unloading Procedure" did not contain a hold point to ensure wor ( wouto not continue until the results of the ir.ner cask volume sample lad been reviewed.
Ints procecural hold point ~is important tu ensure Inat an unplanned and unmonitored release path is -
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 safety evaluations.
This is a Severity Level IV Violation (Supplement I) (50-282/95002-O'1; 50-306/95002-01; 72-10/95002-01(DRP)).
j With respect to this violation, the inspection showed that steps had been j
taken to correct the identified violation and to prevent recurrence.
1 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
i 77
1 While the inspectors recognized that finalizing the loading and e
unloading procedures was contingent upon completion of the dry run and the subsequent incorporation of any lessons learned, there were many aspects _of the procedures which should have been in place Detore tne dry 3
tor exampie, technicai specirication requirements were not effectively incorporated into the loading and unloading procedures I
(paragraph 3.2).
In addition, the licensee dic not como' ete rev' ew and f jr approval of the unloading proceoure until the c an followung su)m'ssion of the oreoperational test report.
Submission 0" this rennr1 1mDiled
[
oad a cask with spent fuel and l
that the icensee was readv
- n subsequently unloac the cask, i:
necessary.
~
The licensee did not take a disciplined approach to inspecting the fuel e
designated for cask storage as evidenced by weaknesses identified by the inspectors during observation of fuel inspection activities (paragraph
~
7.3).
e Some weaknesses were noted with the licensee's documented basis for safety evaluation conclusions (paragraph 8.2).
3
operational checks of vehicle brakes, lifting equipment, turntables, jacks,
,and cask links.
3.1.5 Surveillance Procedure. SP 1075. "TN-40 Fuel Selection and Identification" The inspectors reviewed SP 1075 and the cask loading procedure, D95.1, to l
verify that selected Technical Specification (TS) requirements had been incorporated into procedures. Surveillance requirements for ensuring that fuel assemblies which satisfy the criteria of TS 3.1.1 would be loaded into
.the cask, are defined in TS 4.1.
TS 3.1.1(6) required that, " fuel assemblies known or suspected to have structural 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 gross cladding failures."
The inspectors questioned the efficacy of this technique to provide a thorough inspection of the fuel. After further discussion with Region III staff on fuel inspection techniques, the licensee elected to use video recording equipment to perform the fuel inspection. The inspectors considered this a preferable method for identifying fuel anomalies and ensuring compliance:with TS 3.1.1.
The inspectors observed portions of the actual fuel inspection and identified weaknesses with the licensee's approach to this activity as discussed in paragraph 7.3.
During the review of SP 1075, the inspectors identified that the procedure did not incorporate the requirement of TS 4.1.2, which stated that "before inserting a spent fuel assembly into a cask..., the identity of each fuel assembly shall be independently verified and documented." The inspectors discussed the independent verification requirements of TS 4.1.2 with the licensee. Subsequently, the licensee revised SP 1075 to address independent verification of fuel essembly identification. Based on observations of the actual fuel inspection, the inspectors concluded that the licensee met all TS requirements for fuel identification. The failure to incorporate the requirements of TS 4.1.2 into SP 1075 is considered an example of a violation of Criterion V of Appendix B to 10 CFR Part 50 (50-282/95002-01; 50-306/95002-01; 72-10/95002-01(DRP)).
3.2 Loadina and Unloadina Procedures The inspectors reviewed the loading (D95.1) and unloading (D95.2) procedures for technical adequacy and to determine if the lessons learned from the preoperational testing / dry run had been appropriately incorporated into the procedures.
3.2.1 D95.1. "TN-40 Cask Loadina Procedure" The original D95.1 procedure specified in the prerequisites section that SP 1077 be performed 30 days prior to loading a cask.
However, the Technical
- Specification (TS) 4.19 requirement to perform a visual inspection of the 10
t lifting device (lift beam and extension) and verify operability of the device 7 days prior to use, was not identified in D95.1. There was no procedure in existence identifying actions required to verify operability of the lifting i
device. This issue was identified by the inspectors. The inspectors verified that 095.1 had been updated to include the preoperational testing requirements l
of TS 4.19.
l 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 4
required by TS 4.6.1 to ensure compliance with TS 3.6.1.
The inspectors discussed this issue with the licensee and verified that D95.1 was revised to 1
include specific steps for performing TS-required gamma and neutron dose rate l surveys.
]
' The failure to incorporate the requirements of TS 4.19 into D95.1, to develop a procedure identifying actions required to verify operability of the lifting device, and to include a step for performing radiation surveys of the cask surface before moving a cask to the ISFSI as required by TS 4.6.1, are considered examples of a violation of Criterion V of Appendix B to 10 CFR Part 50 (50-282/95002-01; 50-306/95002-01; 72-10/95002-01(DRP)).
l 3.2.2 D95.2. "TN-40 Cask Unloadina Procedure" a
i The inspectors identified that the final revised and approved D95.2 unloading j
procedure did iot adequately address the "S-reautrement to samole the soent fuel pool for )oaan concentrat' on. Spect"1cally, TS 4.2.1.2 required 4
verification witiin fcur hours of flooding the cask cavity for unloading the fuel assemblies, that.the dissolved boron concentration in the spent fuel pool-water introduced into the cask cavity was greater than or equal to 1800 ppm.
However, D95.2 required sampling four hours prior to lowering the cask in the
!,o pool. The inspectors noted that there may be some time dalay between partially lowering the cask into the spent fuel pool and filling the cask.
l s
'/ /The subject TS requirement is important in that it increased the A'
~
i
' defense-in-depth for ensuring that there was not the potontial for an inadvertent criticality. The inspectors verified that D95.2 was revised to incorporate the TS requirement.
f 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 D95.2 did not contain a hold poiri ta ansure cand nua unti' the sanole resul ts nag Deen reV1awec. The f '?
work wanu nnt 4
unplanned and unmonitored release path would not be created while Tiispectors considered th s procec ura liold pount important to ensure that an s.
in the Auxiliary Building. The inspectors verified that 095.2 was revised to incorporate the subject hold point. The inspectors concluded that the final j
D95.2 procedure contained adequate guidance to ensure that the sampling and other unloading evolutions were performed in a manner that would maintcin 4
exposures to workers as-low-as-reasonably-achievable.
The failure o_f D95.2 to adequately acdress the TS-requirement for samplina the spent ruei pool ana to inciuae n io c notnt to ensure tna resuits or the inner 1
cask volume sample naa Deen rev ewed before allowing work to proceed, are 1
1 11 i
i
1
\\
_ considered examoles of a violation of Criterion V of AD)endix B to 10 CFR Part q
50 (50-282/95002-01; 50-306/95002-01; 72-10/95002-01(DR?)).
3.3 Emeraancy/Off-Normal Procedures The Part 72 license regujr.ed -the-14c4a. tee to develop an abnormal operating 7
procedure (A0P) fora 1uried cask event.;The inspectors asked the licensee if
(
any other emergency 70ff=normat prcceduie's were required in addition to alarm response procedures and the buried cask A0P. The inspectors reviewed the cask f
handling procedures to determine if contingency actions for abnormal events had been addressed.
i The licensee does not have any procedures, in addition to the buried cask, which address off-normal events, The inspectors noted that step 5.0, of procedure D95.1, "TN-40 Cask Loading Procedure," stated that, "Should anything not look right during the performance of this procedure it is imperative that the issue be resolved prior to proceeding. All those involved in the performance of this procedure SHALL have their questions 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" at various steps in the procedure which required that the loading evolution be j
stopped and any abnormal condition evaluated before proceeding. The j
inspectors did not have any further concerns with this issue.
3.4 Conclusions t
The licensee did not complete development of the loading and unloading
-procedures until the day following submission or the preope_rALlonaL teit recort.
5uomtssion of this report 1mpitea that the licensee was ready to load a cask wit'- siient fuel. ~ While the inspectors recognized that t1nal1 zing these h
procedures was contingent upon completion of the preoperational testing I
evolution or " dry run" and the subsequent incorporation of any lessons 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.
Assuming procedural adherence, the final procedures in place for cask handling and loading were adequate to ensure that these evolutions would be conducted i
safely.
4 4.0 Audit Reports. Source Insoections. and Vender Records J
4.1.
Audit and Source Inspection Reports The inspectors reviewed a sample of the licensee's audit and source inspection reports to determine if there were any issues that could affect the quality of the cask. This review included documentation pertaining to associated audit findings. The inspectors also reviewed several fabrication records to verify 1
compliance 'with the design basis documents, including applicable industry standards. The following documentation was reviewed:
12
i i
6.6 Instrument Calibrations The inspectors reviewed the licensee's procedures for calibrating cask survey instruments and determined that the procedures were adequate to ensure proper calibrations. The inspectors will observe instrument calibrations and survey techniques during the actual cask loading evolution.
6.7 ISFSI Monitorina The inspectors walked-down the ISFSI facility and ensured that TS-required 4
theFmoluminescent dosimeters Were in place.
6.8 Radiation Protection (RP) Practices Durina Preoperational Testina The inspectors observed RP practices during the loading dry run and noted that workers were kept informed of the radiological conditions and that RP personnel were prompt and thorough in performing dose rate surveys to monitor changing radiological conditions. The inspectors also considered the decontamination techniques used by the RP staff during the dry run adequate to ensure Technical Specification limits for surface contamination of the cask would not be exceeded.
j 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 load to verify that the integrity of the neutron shield was not affected by the procedure deviation. The inspectors reviewed the licensee's plans for surveying the neutron shield and determined that the survey techniques were adequate to confirm that the neutron shield 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 appropriately addressed items such as dosimetry requirements for workers, survey techniques and the use of calibrated instruments, required air sampling, protective clothing requirements, radiation and contamination area postings, and procedural hold points and work stoppage criteria.
7.0 Pre-operational Testina (Loadina and Unloadina Dry-Run)
The NRC license for the ISFSI required the licensee to conduct pre-operational testing to demonstrate cask handling capabilities before loading the first 22.
-f w..
.~,., - _ _ _
- =
1
l 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 i
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 l
pool and simulated fuel loading; and cask removal from the spent fuel pool an6 l
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'ted i
by tne licensee ouring dry run activities.~
These exemptions were approve \\.
the NRC to prevent any unnecessary damage to the lid seating surface duri l
the dry run and did not affect the licensee's ability to demonstrate unloadins i
a was performed to demonstrate that the transporter and cask would not tip over during cask transport should a seismic event occur. However, the subject SE did not address the consequences of a tip-over accident in the Auxiliary l
l Building rail bay.
The inspectors discussed this issue with the licensee and with representatives from NMSS.
Based on these discussions and the results of a previous analysis involving the loss of all cask confinement barriers during a spent fuel shipping cask handling accident, the inspectors concluded that if a release of radioactivity occurred due to a tip-over event in the Auxiliary Building, the release would be substantially less than 10 CFR Ptrt 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.
While the inspectors noted some weaknesses with the quality of SE No. 344, the inspectors determined that the licensee's conclusion that operation of an ISFSI would not create an unreviewed safety due to an adverse impact on reactor plant operations, was valid.
7.1 Seal Performance Test The inspectors reviewed the licensee's methodology for performance testing of the cask seals. The lid sealing system was designed with three sets of double 0-rings: one set on the circumference of the main lid and one set on the flange covers for each of the vent and drain ports. The spaces between the 0-rings for the lid and each flange were interconnected via drilled channels to the overpressure (0P) port. The OP port was connected to the OP tank which was designed to apply helium pressure to the volume of space between all of the 0-rings.
Should inner-seal leakage occur, helium would leak from the 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 pressure would be monitored on the storage pad and an alam generated if tank pressure was low. The pressure monitoring equipment was prepared and tested prior to cask transport. After cask placement, the pressure monitoring system would be installed on the cask and tested via a surveillance procedure. Completion of these activities was documented in D95.1.
t l
23
8.4 OA Overview of Dry Cask Storace Activities After receipt of the first TN-40 cask at the site, the inspectors determined that the licensee did not have a comprehensive plan to inspect, audit, or monitor dry cask activities onsite, in particular, those activities that interface with the Part 50 license. The inspectors identified several issues that should have been identified by the licensee. After discussion with the inspectors, the licensee developed an " Integrated Dry Cask QA Assessment Plan," which provided direction for the Nuclear Quality Department in the inspection, audit, and surveillance of dry cask storage activities. Once established, the licensee'.s quality verification efforts were effective in identifying issues with the dry cask storage project which required resolution by the line organization.
8.5 Retrievability On May 3, 1995, the licensee submitted on the docket, correspondence that addressed the ability to unload the first TN-40 cask following completion of the May 1995, Unit 2 refueling outage and prior to receipt of the second cask onsite. The NRC's Office of Nuclear Material Safety and Safeguards responded on May 5,1995 to the licensee and stated that the plans described in the May 3 letter to address unanticipated unloading of a cask before another cask had been loaded, would allow ready retrieval of the spent fuel for further processing or disposal as required by 10 CFR Part 72.122(1).
1 8.6 Exit Interview 1
The inspectors met with the licensee representatives denoted in paragraph 8.7 during the inspection period and at the conclusion of the inspection on 1
April 28, 1995. The inspectors summarized the scope and results of the inspection, and discussed the likely content of this inspection report. The licensee acknowledged the information and indicated that some of the information disclosed during the inspection could be considered proprietary in nature.
8.7 Persons Contacted Northern States Power Company
- E. Watzl, Vice President Nuclear Generation fM. Wadley, Plant Manager fK. Albrecht, General Superintendent, Engineering G. Lenertz, General Superintendent, Maintenance
- 90. Schuelke, General Superintendent, Radiation Protection and Chemistry J. Sorensen, General Superintendent, Plant Operations J. Goldsmith, General Superintendent, Nuclear Generation Services Engineering
- T. Amundson, Director, Generation Quality Services
- P. Kamman, Generation Quality Services
- J. Hill, Manager, Generation Quality Services fJ. Bystrzycki, General Superintendent, Project Management 29 c
-r
-n
, - - - + - - - -
Attachment E Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 10 CFR Part 72 May 3, 1995 RECEIVED MAY 0 91995 U S Nuclear Regulatory Commissibn Document Control Desk Attn:
Washin$ ton, DC 20555 PRAIRIE ISIAND INDEPENDENT SPENT FUEL STORACE INSTALLA Docket No. 72-10
(
Materials License No. SNM-2506 Information Related to Unloadint of TN-40 Cask The attached information is provided in response to questions raised during Fuel public meeting on the Prairie Island Independent Spent the April 28, 1995 The questions were related to the ability of NSF to 5 Storage Installation. completely unload the first TN-40 cask following completion of As Unit 2 refueling outage and prior to receipt of the second cask onsite.NSP w shown in the attached assessment, unload fuel from the first TN-40 cask back into the spent fuel pool in a i
timely manner, following the May 1995 Unit 2 refueling outage, if such act on becomes necessary.
We have made no new Nuclear Regulatory Commission commitments in this letter Please contact Gene Eckholt (612-388-1121) if you have any or the attachment.
questions related to the information provided.
f~
b
\\
Roger 0 Anderson Director Licensing and Management Issues Director, Office of Nuclear Material Safety and Safeguards, NRC NMSS Proj ect Manager, NRC Regional Administrator - Region III, NRC Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silber 5 Prairie Island Independent Spent Fuel Storage Installation Service List Assessment of Capability to Unload TN-40 Cask Following May 1995
Attachment:
Unit 2 Refueling Outage M-4-&u d-a P3 p p.
Attachment USNRC Page 1 of 2 '.
e.
ASSESSMENT OF CAPABILITY TO UNLOAD TN-40 CASK FOLLOWING MAY 1995 UNIT 2 REFUELING OUTAGE
~
The Prairie Island spent fuel pool is designed and licensed to store 1386 fuel assemblies. Due to inaccessible locations and other non-fuel bearing
. components, the practical storage capacity is normally considered 1362. After the May 1995 Unit 2 refueling outage there will be a total of 1377 spent fuel 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 poo,1 locations would be required to completely unload a TN-40 cask back-into the spent fuel pool.
However,15 of the non-fuel bearing components noted above can be temporarily relocated as described below to provide the 40 pool locations required to unload a TN-40 cask. These non-fuel bearing components could be relocated by either of the following processes;
- 1. Relocation to Temporary Pool Location:
Move non-fuel bearing components to a temporary location in the pool (most likely the fuel transfer canal). A conceptual design of the' hardware required for this temporary storage he oeen developed. We estimate the required hardware could be fabricateu and the non-fuel bearing components relocated to their temporary locations in approximately 1 working week.
This would adequately support any credible situation requiring cask unloading.
or,
- 2. Relocation to TN-40 Cask:
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 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 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.
Conceptually, the following basic operations would be required to perform these options:
Relocation to Temporary Pool Location:
- 1. Bring cask back from the ISFSI into the plant Auxiliary Building.
- 2. Move cask into the spent fuel pool and remove lid.
-w
+.
--m
-m
USNRC Page 2 of 2
- 3. Fabricate the hardware necessary to temporarily relocate non-fuel bearing components to the transfer canal.
- 4. Relocate non-fuel material to transfer canal.
I
- 5. Off-load fuel from the cask into spent fuel racks.
1 4
- 6. Remove cask from the spent fuel pool.
4
~
- 7. Repair existing cask or provide replacement cask.
)
e t'
- 8. Load the repaired or replacement cask with fuel and return it to the ISFSI.
4
- 9. Relocate the non-fuel bearing components back into spent fuel racks.
j Relocation to TN-40 Cask:
j i
- l. Bring cask back from the ISFSI into the plant Auxiliary Buildir
.y f
- 2. Hove cask into spent fuel pcol'and remove lid.
1
- 3. Remove fuel assemblies from the cask, place them back in the spent fuel j
rocks, and relocate the required non-fuel beat ing components into the cask.
- 4. Replace lid anc remove cask from spent fuel pool.
l 4
- 5. Relocate fuel from the spent fuel pool into a replacement cask.
- 6. Move the loaded-replacement cask to the ISFSI.
- 7. Place the. original task back into the spent fuel pool.
3 j
- 8. Relocate the non-fuel bearing components back into the spent fuel pool.
j
- 9. Remove the empty cask from the spent fuel pool and repair if possible.
j 1-
)
In summary, either of the options described above would allow NSP to j
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 i
4 j
action becomes necessary.
l, i
L e
m
, m
-(Slido.]
lhcas'divos, the iLopoctors havo found the loading procedurco N
.' MR. KUGI.ER :
In tormo of the proceduros j
jo ho acceptable.
There are a number of factors that jicplify the preparation of loading procedures as compared to unic.ading Procedures.
During loading process you've ihcracterized the fuel; you know what condition it is in as ieu put it into the cask.
Also you can take advantage of
!ccoono learned from other licensees and from the dry runs
- hat the licensee performs-on site.
For the unloading procedures, what we are fi-4ing
!o that they are more complex than the loading procedures.
i 47 lnfortunately some of the older SARs fail to recognize this
!ad tend to indicate that unloading is simply the reverse of jocding, which is not true.
For one thing, licensees need io concider the potential condition of the fuel when they go jo calcad it.
Depending on the situation, the fuel say have loca in the cask for decades, and they need to evaluate the icedition of the fuel to tho' extent possible before they l tort unloading it.
j We do put an inert environment into these casks to Frevont oxidation of the fuel.
Assuming that that invircament has been maintained, the fuel should be in good
!4 edition when they go to unload it, but they need to
!valuste.
There are also issues associated with the lcticeding of the cask.
During the unloading process we Ecva to refill the cask with water.
There are some issues Lcocciated with that such as cask pressurization due to btoca generation as you put cold water onto the hot fuel.
1 3 00 the consideration of any thermal shock to the fuel as peu ero reflooding it, and also rrA4clogical protection for i.ho workers during that. phase, boggtse you will be venting l.ho cack.
Generally they are going to direct that venting
- .ither to the pool or to a ventilation system, but they need 4.0 cenoider that.
In addition, there is essentially no cast 48 niecding experience for them to look back on for lessons corned.
So they don't have that information available to
'hrn so compared to loading procedures.
In addition to the working group activities, the staff has been putting increased emphasis on our inspection
.ctivities in this area.
The procedures for the recently oilt facilities have been inspected during the
. resp 0 rational phase using the new inspection procedures
.hnt Sill Travers had mentioned.
These inspections were a ciat offort between the regions, NRR and NMSS.
We coically pool our resources and our expertise to perform
.hoco inspections.
We plan to chatinue those inspections for all
'cture fucilities.
We are also taking a look back at some of the old
'coilities and looking at what inspections have been crformed there to determine whether we feel that we have
.ecamented well enough that those procedures have basa C:pOcted.
If we determine that these: older facilities were
.ct well documented, we are going back.and take a look at Juna cc well'and do further inspections in those locations.
That is all I planned to say on loading cad
.cicading.
If there are no questions, I will turn it over a cher11. en talk ahnne Na9.tarr inie4ae4ua.
.n
,n
I8RAIRIE ISLAND NUCLEAR GENERATING PLANT Attachment G
,, NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES d
TITLE NUMBER:
s,
W4
,, h TN-40 D95.2
'W a,-
CASK UNLOADING PROCEDURE REv:
0 1,
Page 22 of 80 s
' hile the Aux Building Crane is moving the cask into the W
')y^ Spent Fuel Pool, the crane switch will be in " CRITICAL" q 1M,5is Sh. d..T..-
l p siti n. In this condition, the crane wlil be unable to v
gn move more than 1 inch east or west once it passes the roof slot centerline and is within 6 feet either side of the
~
- 4 enclosure.
8.17 Place the Aux Building Crane in the " CRITICAL" mode.
8.17.1 Turn the key switch on the crane controls to the " CRITICAL" position.
Rigger Date j
8.17.2 Key switch on the crane controls verified in the " CRITICAL" j
position by a rigger different from the rigger who changed the j
key switch position.
4 Rigger Date 8.17.3 Place the key in the key cabinet in the Maintenance Supervisor's Office.
8.18 Within four (4) hours prior to loadinn fuel, verify SFP boron l
(
concentration is >2280 ppm."Tio samples must be drawn and
\\
analyzed independently by two separate individuals. Record sample suits on the Cask Loading Report, Appendix A.
8.19 Open the spent fuel pooi enMosures roof hatches.
8.20 Raise the cask to the255' level and position it over the pool.
@)
8.21 Slowly lower the cask into the spent fuel pool while spraying the cask and lift beam with demineralized water to provide a film of clean water on the cask surfaces.
J
)
il o f o 3 'oNr 4 3pp
~~ Attachment H l
c i
?
e 8
f 9'
ISFSI SAR TABLE 5.1-2 l-ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS l
Operation No. of Time Avg. Distance Personnel
.(EiD1 (ft) from Cask Receivina l
1.
Unloading (A1) 2.
Inspection (A2 through A7) 3.
Transfer to cask loading pool.(A8) l Cask Loading Pool l
4.
Lower cask into pool (B1) 5.
Load fuel (B2 through B4) 5 6.
Place lid on cask (BS) 5 7.
Lift cask to pool surface (86)
.5 30 5
l 8.
Install lid bolts (B6) 5 120 3
9.
Drain cavity (B7 through Bil) 5 90 6
10.
Transfers to decontamination area (B12) 3 60 10
~}
Decontamination Area 11.
Decontaminate. cask (C1, C2) 3 120 3
- 12. Remove vent plugs 2
30 5
13.
Drying, evacuating, backfilling (C3 through C13) 2 480 5
14.
Install top neutron shield C14) 2 15 3
15.
Install pressure transducers (C15 through C17) 2 30 5
16.
Pressurize interspace (C18) 17.
Check leak =ee (C19) 2 30 5
18.
Check surr>4 temperature (C20) 2 30 5
19.
Check sur w e 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 i
i I
l l
TABLE'5.1-2 REV. 2 9 h
i
ISFSI SAR TABLE 5.1-2 (Continued)
ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLING OPERATIONS Operation No. of Time Avg. Distance Personnel (min)
(f*1 from Cask Storare Area 23.
Unioad from vehicle position in location (D1, D2, D3) 5 60 5
24.
Check surface dose rate (D6) 5 30 3
25.
Connect pressure instrumentation (D4, D5) 5 30 5
Periodic Maintenance 1.
Visual surveillance (NA) 2 15 5
2.
Repair surface defects (NA) 2 60 3
3.
Instrument testing and calibration (NA) 2 180 5
4.
Instrument repair (NA) 2 60 3
Maior Maintenance (once in 20 years) 1.
Replace cask ' lid seals 3
1950**
8 No measurable dose associated with this activity.
Therefore, the number of personnel, time and distance are not significant.
Parenthetical information corresponds to Table 5.1-1 activity numbers.
l Total time to transfer cask to spent fuel pool, replace lid seals, and return cask to ISFSI pad.
i TABLE 5.1-2 REV. 2 9/91
fs/r/ /M At tachinent I
[
f]
- n. Sinclate
/
NUCLEAR REOULATORY COMMfSSION 77.rJ 7 s e. pmas' The event at Potat Seach occurred en Ray 28. 1996. When the Itcenste struck a spark with the automated melding sackfee to begin weldtag the shfeld 114. a hydesgen burn occurred that generated suff tctent pressure in the cask to L
e.see displace the 114 somewhat leawing it cocked at an angle. As in the case at June is* 1996 Roblasen, the cast had met been drained (except for a small area near the top of the cask to fact 11 tate melding). After investigating the situatten and W. Itacy Sinclatr deterstatng a safe course of actlen, the Itcensee straightened the Ild, filled Don't Waste Richigan the air space at the top of the cast with water, and seved it back late the p.O. Den 1002 spent fees peel. The Itcensee then removed the shleid 114 and unloaded the i
Monroe. Richigan atl61 fuel. The NAC staff. Including an Aegm6nted Inspectlen Team monttered licensee activities during this event. The investigatten of tais event and tts tupiteattens is continetag. The NRC has issued confirmatory actten l
Dear Dr. $Vlat**
1etters (CALs) to the Itcensees for the independent spent feel stgrage I
At m
la a letter to yee dated Decoster 28.1995 I stated that the staff was not h
, '*h e
at e t e I ensees autre of end 5. Ihgelear Regelatory Camissten (MRC) dry storage te refrate from leadtag er unloading casts until after they have coupleted the I
cesks thet peu been unloaded by MC reacter licensees. tace that letter, the actlens discussed in the CAtt and contacted the RRC*
visconste tiectric Power Campany, the 11teesee for the Point teach plants.
so?eeded e cast after a hydrogen Ignittee event that occurred when the I trust inat this information will be helpful to you. I spelegtze that the l'
'see tattlated we14 tag for the cle ere of the shield lit. In additten. I taformatten costained in ey December 29. 1995 letter was not casoletely be,e recently learned of two past cases in uhtch itcensees have onloaded casks accurate. Is you have any questtens. please contact me.
efter identiffre prehtees during the lendtng process. la all three cases.
the Ilconsee a 7 at completed the leading process and none of the affected
$ bcuely*
casts were asets beyond the decentaminatten aree. The perpase of this letter ts to preetde yee with teformetles related to these activities.
ORIGig $1G%to 81:
The first case occurred at the Sorry Power Statten in Virgista in hoveedper Andrew J. Kugler Project Manager 1996. The licensee had leaded the first Caster V/21 cast a design that uses Project Directorate Itt-3 two telted llds with seal rings. The inner tid had been Installed and the Otetsten of Reactor Projects III/IV cesk was drained. Det the seal ring failed the hellan leak test regatred for Office of Nuclear Reacter Regulatten this type of closure. The Itteesee moved the cast back to the spent feel poe1. refleeded it, and levered it into the cask pit. tihen the 114 was cc: Mr. Richard ti. Smedley rensved. f9ie Itcensee deterutned that the seal et had shifted est of its tentumers Power Ceepany e'eeve due to hydrodynamic fortes as the 114 was oced ente the cast. The Itcensee revis;d the leading procedure to lower i e Ild more slee's and the See mest page probise has not escered.
DISTRIOUTION The second case happened at the N.S. Rottasen Nvelerr Plant 14 february 198g.
Doctet nos. h25% 72-7* 7bl007 Cuted t h The Itcensee had leaded the first IRMyt!-7P test and seved it to the pygtyg P033 R/F (EDO e665) decentaminatten area. la preparatten for welding the inner (shteld) 116. the fEasberent EConna hten Itcensee performed a servey of the redtatten dose rates above the ild. The Jaea w
g Itcensee teend dose rates that were higher than these predicted and decided to 394asnik Creamer. OCC move the cast back to the spent feel peel and unload it as past of the LCluk. OGC Ktes. m lowesttgatten of the done rates. The cast had met yet been drstned and se Ptag.18455 YTharpe. fots5 issors assectated with refleeding a cast were not appitcable. The 19tensee
- w.\\.DRT.-CASK \\5INCLA.IR LT.R.*S.tt PRE.VIOUS CONCulu.ttrC.E.,,
deterstood that eartatten in the dose rates was withle t!e accuracy of the m n...DO.C.LsEN.T lt4ft:
C: o.w e - c c..
.v setn.eds used for done predtttten and well within the Itetts to the techntcal spectf tcattens. The Itcensee then proceeded to lead the cast.
Of flCE LA:PC33*
iC PM:PD33*
iC W:5fPC*
l OGC*
IC 0:P033 ic mmME OFester-Corseen AKvgler CMaughney Creamer GMarcus th lefermetten conceretng these two cases was not widely known within the Npt.
unit 06/ R/e6 06/12/96 06/I3/96 06/te/96 OM/f/96 @
staff because the prealees that led the Itcensees to unload the casts were not safety-stgnificant. In both cases the Itcensees fevnd the problems threvgh OHICIAL Ahuss COP:
appropetate testing er senttoring and took prompt, conservative corrective I
actlees.
gm UlE CENiiM The utensee fe, this facmt, ha, not,et isaded an, ca,1,.
,d f (-
et Post tils:lUED vierfic stAtts p
{
NUCI. EAR REGUL.ATO'.tT COMMISSION
=aowneten. e s sena.
RI!55CED
- ?.
3 e
Me 70.19%
_ - - _ _ _ _ _ _ _ _ _ _ _ _