DD-97-18, Director'S Decision DD-97-18 Re 970528 Petition Filed by Prairie Island Indian Community Requesting That USNRC Take Action to Determine That NSP Violated Requirements of 10CFR72.122(i).Request Denied Due to Listed Reasons
| ML20216G400 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/29/1997 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20216G398 | List: |
| References | |
| 2.206, DD-97-18, NUDOCS 9709150179 | |
| Download: ML20216G400 (19) | |
Text
___ _ _ _ __ ._
i, DD 97 18 UNITED STATES OF AMERICA i
- NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Samuel J. Collins, Director in the Matter of )
)
NORTHERN STATES POWER COMPANY ) Docket Nos. 50 282,
) 50 306, and 7210 (Prairie Islend Nuclear Plant and )
Prairie Island independent Spent ) (10 CFR 2.206)
Fuel Storage Installation) )
)
DIRECTOR'S DECISION UNDER 10 CFR 2.206
- 1. INTRODUCTION On May 28,1997, the Prairie Island Indian Community filed a Petition pursuant to Section 2.206 of Title 10 of the Code of Federal Reaulations (10 CFR 2.206) requesting that the U.S. Nuclear Regulatory Commission (NRC) take action to accomplish the following:
- 1. Determine that Northem States Power (NSP) violated the requirements of 10 CFR 72.122(l) by using its Materials License No. SNM 2506 for an independent Spent Fuel Storage Installation (ISFSI) prior to establishing conditions for safely unloading the TN 40 dry storage containers;
- 2. Susper,o Materials License No. SNM 2506 for cause under 10 CFR 50.100 until such time as all significant issues in the unloading process, as described herein [the Petition), have been resolved, the unloading process has been demonstrated, and until an independent third party review of the TN-40 unloading procedure has been conducted;
- 3. Provide Petitioners an opportunity to participate fully in the reviewing of the unloading procedure for the TN 40 cask, hold hearings and allow Petitioners to participate fully in these and any other procedures initiated in response to this Petition; and 4, Update the Technical Specifications (TS) for the Prairie Island ISFSI to incorporate mandatory unloading procedure requirements.
The Petition has been referred to me pursuant to 10 CFR 2.206. The NRC letter dated June 27,1997, to Byron White, on behalf of the Petitioners, acknowledged receipt of bbDO O
O 202 PDR
i, 2
the Petition and provided the NRC staffs determination that the Petition did not require immediate action by the NRC. A notice of receipt was published in the FEDERAL REGISTER on July 3,1997 (62 FR 36085).
On the basis of the NRC staffs evaluation of the issues and for the reasons given below, the Petitioners' requests are denied, ll. SACKGROUND On October 19,1993, the NRC issued Materials License No. SNM 2506 to NSP (the l
l licensee) to allow storage of spent nuclear fuelin TN 40 dry storage casks, designed by l
Transnuclear incorporated, at the ISFSI located at the Prairie Island Nuclear Plant. No spent nuclear fuel was allowed to be loaded into a storage cask at Prairie Island until several i preoperationd license conditions were satisfied. Among the preoperational license conditions were a required training exercise (dry run) of the loading, handling, and unloading activities for 15a TN-40 casks and the implementation of written procedures describing the actions to be taken during operation, off-normal, and emergency conditions associated with the Prairie Island ISFSI. The NRC issued TS defining operating limits, surveillance requirements, design features, and administrative controls as Appendix A to Materiais License No. SNM 2506.
A report dated April 20,1995, submitted by the licensee to the NRC pursuant to 10 CCR 72.82(e), provided the results of the preoperational tests that were required to be performed by the licensee before loading of spent fuel into a TN-40 cask. On May 11,1995, the NRC granted a schedular exemption to the provision of 10 CFR 72.82(e) that requires licensees to submit the preoperational test results at least 30 days before receipt of spent fuel into the ISFSI. The basis for the exemption was the fact it.at the NRC staff had reviewed cask fabrication records, observed portions of the preoperational test activities as they occurred,
3 and completed its review of the report submitted on April 20,1995. On May 12,1995, the licensee began loading spent fuel assemblies into a TN 40 cask. The licensee subsequently placed the cask, and casks luaded since that time, onto the storage pad within the Prairie Island ISFSI.
NRC regulations include a requirement that an ISFSI be designed to provide for the ready retrieval of spent fuel or high level radioactive waste for further processing or disposal. This regulation,10 CFR 72.122(l), provides as follows:
Retrievability. Storage systems must be designed to allow ready retrieval of spent fuel or high level radioactive waste for further processing or disposal.
Certain events or conditions could warrant removing a TN-40 cask from the Prairie Island ISFSI and retuming it to the spent fuel pool and unloading the stored fuel assemblies.
In addition to the regulatory requirements in 10 CFR 72.122(l) pertaining to retrieval of the fuel assemblies for further processing or disposal, the TS for the Prairie Island ISFSI requires the licenses to take corrective actions in response to those design-basis events or conditions that may challenge the integrity of the storage cask or the cladding of the spent Gel assemblies. For example, Section 2.3," Maximum Cask Ufting Height," Section 3/4.3,
" Maximum Helium Leak Rate," and Section 3/4.5," Maximum Cask Surface Temperature," of the TS include provisions for unloading of a TN-40 storage cask in response to the specified events or conditions.
NRC regulations in 10 CFR Part 72 require that the design of the storage system anu ine procedures implemented by sp3cific licensees suppon the unloading activity, whether it is being performed to allow further processing or disposal of the spent fuel of it is required as part of the response to an unplanned event or condition, while preventing gross ruptures of
e 4
the fuel cladding in order to prevent operational safety problems. Unloading procedures should also include contingencies in case fuel cladding has degraded during storage such that additional measures are necessary to address increased radiological hazards during the unloading process.
NRC regulations, facility licenses, and NRC approved quality assurance programs require j licensees to establish and maintain a formal process for the preparation and issuance of procedures and changes thereto. NRC assessments of licensee procedures are generally conducted as part of the NRC's inspection program. In this instance, the major procedures pertaining to dry cask storage activities at Prairie island, including the procedure for unloading a cask, were reviewed .y the NRC staff during a specialinspection conducted from January 24 through May 11,1995. In addition to the review of the licensee's facility and procedures, the NRC inspe: tors observed preoperational testing that the licensee was required to oerform before loading casks with spent fuel assemblies. The inspection findings are deumented in NRC Inspection Report 50-282/95002; 50-306/95002; 7210/95002(DRP),
detsd June 30,1995.
The NRC inspectors identified several instances in which the procedures for dry cask storage activities that the licensee had in place at the beginning of the inspection, including the procedures for loading and unloading of the TN 40 casks, did not ensure compliance with the requirements of the license. Although the inspectors were able to verify that the licensee corrected the identified procedural deficiencies during the course of the inspection, a Notice of Violation was issued to the licensee for falling to satisfy Criterion V of Appendix B to 10 CFR Part 50, which for activities affecting quality, requires the preparation and adherence to procedures appropriate to the circumstances, in addition, the inspectors identified
5 weaknesses in the licensee's initial performance in overseeing the activities of the cask vendor and in overall planning for dry cask storage activities. However, on the basis of the licensing reviews and inspection findings, documented in inspection Report 50 282/96002; 50 306/95002; 7210/95002(DRP), the NRC stan concluded that as of May 1995, the licensee I
had corrected the identified deficiencies and was ready to safely load and, if necessary, unload spent nuclear fuel in TN 40 casks.
l In July 1995, the NRC staff issued an action plan for dry cask storage to manage the resolution of a variety of technical and process issues that were identified during the licensing reviews and inspections completed for the first several ISFSI facilities. An item related to the loading and unloading of dry storage casks was added to the action plan, in part to ensure i
that the importance of the unloading procedures was emphasized to licensees and technical issues related to unloading problems were resolved. Addition of an item pertaining to unloading was deemed prudent because the staff observed that some unloading procedures implemented by licensees neglected to consider contingencies and assumptions related to possible fuel degradation, gas sampling techniques, cask design issues, radiation protection requirements, and the thermal-hydraulic behavior of a cask during the process of cooling and filling it with water from the spent fuel pool.
To implement the action plan, the NRC staff formed a working group to identify issues associated with loading and unloading processes for dry storage casks and to propose means of informing the industry and the NRC staff of those issues. The working group considered industry experiences, concems identified during reviews and inspections, and other issues related to loading and unloading procedures, The working group completed its reviews in April 1996. The concems related to unloading procedures reviewed by the
6-I working group were found to involve either (1) isolated occunences that had been adequately resolved by site specific corrective actions or (2) generic issues that were addressed by incorporating remedial measures into ongoing staff activities, such as the preparation of revised inspection procedures or other guidance documents.
To fulfill some of the goals included in its dry cask storage action plan, the NRC staff has emphasized the importance of unloading procedures and shared observations with licensees using or considering dry cask storage during opportunities such as the Spent Fuel Storage and Transportation Workshop held in May 1996 and meetings with Individuallicensees. The staff revised inspection procedures to specifically instruct NRC inspectors to review unloading procedures developed by licensees and to identify those issues that warrant particular attention. Guidance hcluded in NRC inspection Procedure 60855, " Operation of an ISFSI,"
issued February 1,1996, states:
For unloadine activities, attention should be paid to how the licensee has prepared to deel with the potential hazards associated with that task.
Some potentialissues may include: the radiation exposure associated with drawing and analyzing a sample of the canister's potentially radioactive atmosphere; steam fiashing and pressure control as water is added to the hot canister; and filtering or scrubbing the hot steam / gas mixture vented from the canister, as it is filled with water.
Similar guidance was included in NUREG 1536, " Standard Review Plan for Dry Cask Storage Systems," issued in January 1997. Application of the revised guidance ensures that recent and future reviews will address the adequacy of unloading procedures developed by licensees. The staff also issued NRC Information Notice 97-51," Problems Experienced with Loading and Unloading Spent Nuclear Fuel Storage and Transportation Casks," dated July 11,1997, to inform licensees of operating experiences and problems encountered with the loading and unloading of storage and transportation casks for spent nuclear fuel.
e 7
To address those ISFSis that began operation before the improvements in thu NRC's review and inspection guidance, the staff performed audits or inspections of those ri;ensee programs for which the inspection record did not document whether the unloading procedures adequately addressed the major issues included in the action plan. Regarding Prairie Island, ,
the staff reviewed the available information and determined that additional reviews or inspections were not necessary because the assessment of the unloading procedure l performed as part of the inspection documented in NRC Inspection Report 50 282/95002; 50 306/95002; 7210/95002(DRP) adequately addressed the concems included in the NRC action plan.
Ill. DISCUSSION The Petition requests four act:ons by the NRC on the basis of the contantion that the unloading procedure implemented by the licensee was inadequate and, therefore, the licensee violated the NRC regulation requiring it to have the ability to readily retrieve spent fuel or high-level radioactive waste for further processing or disposal.
Item 1: Determine That the Licensee Violated 10 CFR 72.122m in support of the Petition's contention that the licensae violated NRC requirements, the Petitioners claim that the procedure to unload a TN-40 cask at Prairie Island has not been adequately evaluated or tested because neither the NRC nor NSP has completely demonstrated that a TN-40 dry cask can be unloaded after it has remained on the storage pad for a number of years. The Petitioners state that their request is supported by the fact that the preoperational test results for the Prairie Island ISFSI were submitted to the NRC on the day before the unloading procedure was approved by the licensee's Operations Committee. The Petitioners also express concem that nnly portions of the licensee's
L.
8 unloading procedure were tested during the required preoperational tests and, therefore, the tests did not provide assurance that an unloading can be done safely, in addition, the Petitioners state that procedures for unloading a cask should address specific conoems L regarding failed fuel recovery and possible contamination of the spent fuel pool, venting of radioactive gases, functional checks of radiation monitoring and ventilation systems, and the build-up of steam when water is pumped into the cask as part of the unloading process.
As previously mentioned, cask designs and associated procedures are required to l support the unloading of the spent fuel assemblies either to support further processing or disposal or in response to an unplanned event or condition that may challenge the integrity of the storage cask or the cladding of the spent fuel assemblies. Although the NRC staM agrees with the Petitioners' promise that actually unloading a storage cask would likely result in licensees loaming lessons that could result in additional enhancements to unloading
. procedures, the sta# does not agree that an actual demonstration of the unloading procedure at Prairie Island is warranted. in addition to the staffs review of the procedure for unloading a TN-40 cask at Prairie Island, reasonable assurance that the TN 40 casks can be safely unloaded is provided by a variety of experiences related to the use and storage of radioactive meterials. These experiences include the dry-run exercises that were performed to verify key aspects of unloading procedures for the TN 40 cask; related research sponsored by the commercial nuclear industry, the U.S. Department of Energy, and the NRC; actiJa! loading and unloading of transportation casks; loading of storage casks; handling of spent fuel assemblies under various conditions; and performing relevant maintenance and engineering activities associated with reactor facilities.
Regard!ng the Petitioners' concems pertaining to the dates'of the submittal of
.g.
l I
preoperational tests and the approval of the licensee's unloading procedure, the NRC staff identified this discrepancy in Inspection Report 50 282/95002; 50 306/95002; 7210/95002(DRP). The administrative controls included in the TS for the Prairie Island ISFSI require that the Operations Committee review and approve procedures and changes thereto. The approval of the Operations Committee is usually the last step in the process for preparing or revising a procedure. The fact that the Operations Committee approved the procedure shortly after submittal of the preoperational test results and before fuelloading I
! satisfied the preoperational license condition to implement written procedures before loading t
spent nuclear fuelinto a TN 40 cask. This matter does not, therefore, represent a violation of NRC requirements or introduce concems pertaining to the technical adequacy of the unloading procedure.
The Petitioners identified several concems pertaining to the lack of specific guidance in the unloading procedure to address a scenario in which significant fuel degradation occurs during storage. The NRC staff agrees with the Petitioners that such a scenario would complicate the unloading process by requiring additional measures and precautions to limit the release of radioactive materials from the cask into parts of the reactor facility and nearby environs. The licensee's unloading procedure includes a step to sample the atmosphere within the cask cavity to test for Jr.Jioactive and flammable gases before venting the cask cavity and loosening the bolts securing the cask lid. Following the analysis of the gas
! sample, the licensee's unloading procedure includes a hold point to allow personnel to determine whether additional steps or precautions are warranted. While acknowledging many of the Petitionera' legitimate concems regarding the potential difficulties in retrieving failed fuel from dry storage casks, the NRC staff has concluded that licensees need not be
10 required to incorporate specific guidance into the normal unloading procedure to address this l
)
unlikely situation. The staff's conclusion is based, in part, on the fact that the required compensatory actions and precautions needed to address such situations may vary significantly, depending on the actual results from the analysis of the gas sample. Requiring the licensee to include contingencies or steps in the unloading procedure to address the unlikely event of failed fuel may unnecessarily complicate and delay the unloading of fuel assemblies that have remained intact during storage. On the basis of licensees' experiences in developing and implementing plans to address the problem of fuel assemblies damaged during reactor operations, in handling radioactive wastes of various forms, and in resolving other comparable problems, the NRC staff has confidence that licensees could, if necessary, develop a plan to retrieve damaged fuel from a storage cask while minimizing the radiological consequences to plant workers and the general public, in addition to the general confidence of the NRC staff that the technical problems associated with retrieving failed fuel could be overcome, requirements for planning and executing such an activity are contained in the licenses issued for the Prairie Island ISFSl and the Prairie Island Nuclear Generating Plant, and NRC regulations in 10 CFR Parts 20,50, and 72. The NRC staff has, therefore, accepted gas sampling and defined hold or decision points before breaching the cask confinement boundary as an adequate means to address concems pertaining to the unlikely degradation of fuel assemblies during storage.
The specific issues raised by the Petitioners to support their claim that the licensee's unloading procedure is deficient are addressed below.
fa) Failed Fuel Considerations As previously discussed, the NRC staff has accepted that procedures developed by
9 licensees to support unloading of dry storage casks do not need to address the retrieval of failed fuel provided that measures to detect possible fuel degradation and a defined hold point for determination of possible compensatory actions are appropriately placed within the subject procedures. As documented in NRC Inspection Report 50 282/95002; 50 306/95002; 7210/95002(DRP), the licensee had odginally failed to incorporate a step in the unloading procedure for taking a gaseous sample from the cask in order to ensure that fuel degradation had not occurred during storage. However, in response to the findings of the NRC I inspectors, the licensee incorporated sampling of the cask atmosphere and a hold point for deliberation into the unloading procedure and the revised procedure was in place before spent nuclear fuel was loaded into a TN-40 cask. The NRC staff has found that this action, in combination with the requirement that spent fuel assemblies loaded into TN 40 casks be free of gross clado,;ig defects, provides reasonable assurance that the licensee will not unknowingly breach the confinement boundary of a cask containing failed fuel. In the unlikely event that the gaseous sample indicates that spent fuel assemblies have degraded during storage, the unloading procedure it.structs the licensee's Operations Committee to add steps or precautions to the procedure in order to minimize the radiological consequences of retrieving the failed fuel. The NRC staff has found this approach to be acceptable and does not require the licensee's normal unloading procedure to include contingency actions to address the possible release of radion:the materials to parts of the reactor facility, including the spent fuel pool, that may ;~ cur if fuel assemblies degrade during storage. The NRC staff
. believes, however, that the Petitioners have identified valid concems regarding the potential recovery of fuel assemblies that have unexpectedly degraded during storage. As previously mentioned, the staff believes that the regulations and licenses issued by the NRC require the 4
licensee to address these and other problems that may occur in the unlikely event that fuel assernblies that have degraded during storage need to be unloaded from dry storage casks.
(b) Ventina of Radioactive Gases The possible need to vent radioactive gases from a cask is among the issues that the l
licensee would need to address if the required sampling of the atmosphere within a cask indicates that the spent fuel assemblies have experienced unanticipated degradation during i
storage. As with the concem regarding the contamination of the spent fuel pool, the need to vent the cask while minimizing the radiological consequences of unloading a cask containing l
l failed fuelis an issue that the licensee would need to address before revising the procedure and proceeding with the unloading process. In addition to ensuring that the unloading activity results in occupational doses and doses to members of the public that are as low as is reasonably achievable (see 10 CFR 20.1101), the licensee would need to perform the venting of a cask containing failed fuel in accordance with the Prairie Island Nuclear Generating Plant Facility Operating Licenses, associated TS, and applicable regulations.
(c) Radiation Monitors The Petitioners contend that the unloading procedure must include a "stop-check" to verify that ventilation systems and radiation monitors are functioning before the venting of a cask is performed Although agreeing with the Petitioners' general premise that prerequisites to preforming procedures should include establishing confidence in the tools and equipment balng used, the NRC staff notes that during the anticipated unloading of spent nuclear fuel that has not degraded during storage, special ventilation or radiation monitoring equipment beyond that specified in the licensee's unloading procedure and radiation protection program is not required. The unloading procedure requires the involvement of radiation protection s
personnel and the activity must be controlled in accordance with the licensee's radiation protection program, which includes provisions for the maintenance and calibration of radiation detectors. Although the venting process is not expected to need ventilation systems equipped with filters and radiation monitors, the spent fuel pool special ventilation system could be used if necessary. The spent fuel pool special ventilation system is required to be operable during subsequent steps in the procedure if spent fuel assemblies are being moved 1
and the system must be tested and maintained in accordance with the TS for the Prairie Island Nuclear Generating Plant. In the unlikely event that the licensee needs to unload a cask containing degraded fuel assemblies, confirming the operability of those ventilation j systems and additional radiation monitoring equipment being used to minim!ze the release of 1 I
radioactive materials is an activity that the licensee would need to address before revising the procedure and proceeding with the unloading process.
(d) Steam Build-uo The Petitioners expressed concerns regarding the reaction of the cask and stored fuel assemblies to the introduction of spent fuel pool water during the execution of the unloading procedure. The unloading procedure includes the partial immersion of the TN-40 cask into the spent fuel pool, connection of hoses to the vent and drain connections, and the slow introducti>n of spent fuel pool water to the car.k cavity and stored fuel assemblies. The procedure instructs personnel to continuously monitor the temperature and pressure i
instrumentation installed on the vent connection and to stop pumping water if the pressure exceeds 10 psig or the temperature exceeds 240 'F. In the staffs judgment, the cooling process imposed by these limitations on temperatures and pressures at the vent port of the cask will adequately ensure that the cooling of the cask and spent fuelis gradual and,
l .
- 14
- thereby, prevent safety problems that could hypothetically result from damage to the cask or 4
the fuel assemblies because of stresses induced by a poorly controlled addition of cooling water from the spent fuel pool.
The Petitioners expressed conoams pertaining to the range of the instrumentation used during the venting of a TN 40 cask and stated that higher ranges for temperature and pressure are necessary. The instrumentation ranges specified in the unloading procedure's drawing of the cask vent port adapter are 50
- 300*F for temperature and 0
- 30 psig for pressure. While not judging if these are the optimum ranges for the instrumentation, the NRC staff finds that the ranges are adequate to support the administrative limits of 240*F and 10 psig established in the procedure and the related response action of stopping the addition of water to the cask if these administrative limits are exceeded. Regarding the Petitioners' concem regarding the need to post hazard wamings during the refilling of a cask, the unloading procedure does include several notes and precautions to remind personnel that the fluid exiting the vent port may present radiological and thermal hazards.
In summary, many of the Petitioners' concems pertain to potential problems with unloading spent fuel from a TN 40 cask if the fuel cladding has degraded during storage.
While acknowledging that such concems regarding the potential difficulties in retrieving failed fuel from dry storage casks are legitimate, the NRC staff has concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to address this unlikely situation. On the basis of its review of the information provided by the Petitioners and its reviews of the licensee's procedure for unloading TN-40 casks at Prairie Island, the.NRC staff has not identified violations of 10 CFR 72.122(l) or other regulatory requirements pertaining to the content or quality of the licensee's unloading procedure.
o.
15-ltem 2: Suspend Materials License No. SNM 2506 On the basis of the contention that the licensee's unloading procedure was inadequate, the Petitioners requested that Materials License No. SNM 2506 be suspended until such time as the significant issues 1/1 the unloading process have been resolved, the unloading process has been demonstrated, and an independent third-party review of the TN-40 unloading procedure has been conducted.'
As previously stated, the NRC staff has performed a review of the procedure for unloading a TN 40 cask at Prairie Island. The review, including verification that the licensee's unloading procedure was revised to address deficiencies identified by the NRC inspectors, is documented in NRC Inspection Report 50 282/95002; 50-306/95002; 7210/95002(DRP). The review performed during the NRC inspection, subsequent evaluations performed by the NRC staff as part of the activities associated with the dry cask storage action plan and the review of this Petition, and the required control of the procedure in accordance with licensee programs developed in accordance with NRC regulations, facility licenses, and NRC-approved quality assurance programs provide reasonable confidence that the licensee could, if necessary, safely unload a TN 40 cask.
Regarding a third-party review, the NRC staffs concem about the quality of licensees' unloading procedures led it to include the issue in the dry cask storage action plan. The action plan provided a framework for the identification and resolution of various technical and
' The Petitioners request that Materials License No. SNM-2506 be suspended for cause in
- accordance with 10 CFR 50.100. Provisions for the modification, revocation, or suspension of
- the licenses for ISFSI facilities are contained in 10 CFR 72.60. The possible reasons for suspending licenses for ISFSis in accordance with 10 CFR 72.60. are similar to the corresponding reasons for suspending licenses for production and utilization facilities in accordance with 10 CFR 50.100.
16 -
administrative issues related to the use of dry storage casks. The previously mentioned actions taken by the NRC staff and licensees adequately resolved the identified issues pertaining to cask unloading procedures. In the specific case of the unloading procedure at Prairie Island, the licensee revised the procedure to address the problems identified by the staff during its inspection. On the basis of the actions it has already taken, the NRC staff does not believe that the situation warrants additional review of the licensee's unloading procedure by an independent third party, item 3: Allow Petitioners to Review Procedure. and for NRC to Hold Hearinas and Allow Petitioners to Participate in the Proceedinas The licensee has provided the NRC with the unloading procedure, including Revision 2, i dated November 8,1996, for placement into the public record, and the Petitioners have been supplied with or have obtained copies of the procedure from the NRC's document control system. Accordingly, Petitioners have had the opportunity to review a recent revision of the unloading procedure. For the reasons previously discussed in this decision, the NRC staff sees no reason to undertake additional reviews of the procedure or to initiate a formal proceeding in which the Petitioners could participate. Although the NRC has decided not to initiate a hearing in response to inis Petition, the Petitioners are encouraged to continue their interactions s'th the NRC staff regarding concems or questions about the operation of the Prairie Island Nuclear Generating Plant or the Prairie Island ISFSt.
Item 4: UDdate the Technical Soecifications for the Prairie Island ISFSI to incomorate Mandatorv Unioadino Procedure Reauirements The TS for ISFSIs are required, by 10 CFR 72.44, to include requirements in the following categories:
4e 17 -
(1) Functional and operating limits and monitoring instruments and limiting control settings; J (2) Limiting conditions; (3) Surveillance requirements; l (4) Design features; and l (S) Administrative controls.
Although the TS for the Prairie Island ISFSI requires that TN 40 casks be unloaded if '
certain events or conditions defined in the TS are satisfied, the TS do not include specific requirements for the unloading process. The content of the TS for the Prairie Island ISFSI is typicalin this respect since neither 10 CFR 72.44 nor the associated regulatory guidance documents specify that technical specifications should include special requirements for the l unloading procedure.' Instead, the functional and operating limits, limiting conditions, l
l administrative controls, and other requirements included in the TS for the Prairie Island ISFS; are intended to maintain the cask and stored spent fuel assemblies within the limits established for safe operation during storage within the ISFSI and activities such as loading and unloading of the casks. For example TS 2.3 limits the allowable lifting heights during movement of the cask from the ISFSI and TS 3/4.2 requires a measurement of the boron concentration of the water in the spent fuel pool before water is introduced to the cask during the unloading process.
The absence of specific requirements in the TS to control the unloading process does not diminish the importance that the NRC staff places on this activity or the validity of the 8
Recent NRC staff guidance pertaining to the appropriate content of technical specifications is provided in NUREG-1536, " Standard Review Plan for Dry Cask Storage Systems," published in January 1997 Similar guidance is provided by NRC Regulatory Guide 3.61," Standard Forma' and Content for a Topleal Safety Analysis Report for a Spent Fuel Dry Storage Cask," issued in February 1989, and NRC Regulatory Guide 3.48," Standard Format and Content for the Safety Analysis Report for an independent Spent Fuel Storage Installation (Dry Storage)," issued in October 1981.
l l
Petitioners' concems. The NRC staff believes that other regulatory requirements provide an equivalent level of protection to the Petitioners' request to include specific requirements in the TS to control the unloading of a TN 40 cask. The administrative controls in the TS for the Prairie Island ISFSI require that the associated procedures, including the unloading procedure, be prepared, reviewed, and maintained in accordance with the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses and associated TS.
In addition. under existing NRC requirements, the licensee must adequately implement procedures to control loading, maintaining, and unloading of dry storage carks (see 10 CFR l 72.122,10 CFR 72.150, and 10 CFR 72.152). For example, the NRC inspection documented l In inspection Report 50-282/95002; 50 306/95002; 7210/95002(DRP) resulted in a Notice of Violation issued to the licensee because the licensee failed to satisfy the NRC's requirements in Criterion V of Appendix B to 10 CFR Part 50 by not having incorporated appropriate steps and precautions into the original procedure deve!oped to control unloading of a TN 40 cask.
As demonstrated by the example, no changes to the TS or the Safety Analysis Report (SAR) are needed to ensure that enforceable operating controls and limits are in place to address the unloading of a cask, in regard to another concern raised by the Petitioners, the Prairie Island ISFSI SAR and other docketed correspondence do state that unloading a TN-40 cask would be performed using a procedure that is basically the reverse of the procedure used to load the cask.
Although this statement, in a general sense, is true, the NRC staff agrees with the Petitioners that such statements may be misleading in that they oversimplify the description of the unloading activity. For this reason, the NRC staff included an item related to unloading procedures in its dry cask storage action plan to ensure that actual unloading procedures did
not refiect such an oversimplified representation. The unloading procedure for the dry i
storage casks at Prairie Island was inspected by the NRC staff and, as previously discussed, j was ultimately found to provide adequate guidance to control the unloading process.
IV. CONCl.USION For the reasons described above, the NRC has determined that no adequate basis exists for granting the Petitioners' request for suspension of Northem States Power Company's license for dry cask storage of spent nuclear fuel at Prairie Island or for taking the other actions requested by the Petitioners. While acknowledging that the Petitioners' concems regarding the potential difficulties in retrieving failed fuel from dry storage casks are legitimate, the NRC staff has concluded that licensees need not be required to incorporate specific guidance into the normal unloading procedure to address this unlikely situation.
A copy of this decision will be filed with the Secretary of the Commission for the Commission to review in accordance with 10 CFR 2.206(c).
As provided by this regulation, this decision will constitute the final action of the Commission 25 days after issuance, unless the Commission, on its own motion, institutes a review of the decision within that time.
Dated at Rockville, Maryland, this 29thday of Augu st 1997.
FOR THE NUCLEAR REGUI.ATORY COMMISSION a or Office of Nuclear Reactor Regulation 4