[Federal Register Volume 62, Number 173 (Monday, September 8, 1997)]
[Notices]
[Pages 47227-47232]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-23696]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-282, 50-306 and 72-10]
Northern States Power Company, Prairie Island Nuclear Plant and
Prairie Island Independent Spent Fuel Storage Installation, Issuance of
Director's Decision Under 10 CFR 2.206
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, has issued a Director's Decision concerning a Petition
dated May 28, 1997, filed by the Prairie Island Indian Community
(Petitioners) under Section 2.206 of Title 10 of the Code of Federal
Regulations (10 CFR 2.206). The Petition requested that the NRC (1)
determine that Northern States Power Company (the licensee) 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) suspend Materials License No. SNM-2506 for
cause under 10 CFR 50.100 until such time as all significant issues in
the unloading process, as described in 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 the Petition; and (4)
update the Technical Specifications for the Prairie Island ISFSI to
incorporate mandatory unloading procedure requirements.
The Director of the Office of Nuclear Reactor Regulation has
determined that the Petition should be denied for the reasons stated in
the ``Director's Decision Under 10 CFR 2.206'' (DD-97-18), the complete
text of which follows this notice. The decision and documents cited in
the decision are available for public inspection and copying in the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC, and at the local public document room located at
the Minneapolis Public Library, Technology and Science Department, 300
Nicollet Mall, Minneapolis, MN.
A copy of this decision has been filed with the Secretary of the
Commission for the Commission's review in accordance with 10 CFR
2.206(c). As provided therein, this decision will become 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 29th day of August 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Director's Decision Under 10 CFR 2.206
I. 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
Regulations (10 CFR 2.206) requesting that the U.S. Nuclear Regulatory
Commission (NRC) take action to accomplish the following:
1. Determine that Northern 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. Suspend Materials License No. SNM-2506 for cause under 10 CFR
50.100 until such time as all significant issues in the unloading
process, as described herein [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 the Petition and provided the NRC
staff's 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 staff's evaluation of the issues and for
the reasons given below, the Petitioners' requests are denied.
II. Background
On October 19, 1993, the NRC issued Materials License No. SNM-2506
to NSP (the licensee) to allow storage of spent nuclear fuel in TN-40
dry storage casks, designed by 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 preoperational license conditions were satisfied. Among the
preoperational license conditions were a required training exercise
(dry-run) of the loading, handling, and unloading activities for the
TN-40 casks and the implementation of written procedures describing the
actions to be taken during
[[Page 47228]]
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 Materials License No. SNM-2506.
A report dated April 20, 1995, submitted by the licensee to the NRC
pursuant to 10 CFR 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 that the NRC staff had reviewed cask fabrication
records, observed portions of the preoperational test activities as
they occurred, 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 loaded 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 returning 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 licensee 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 fuel
assemblies. For example, Section 2.3, ``Maximum Cask Lifting 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 and the procedures implemented by specific licensees
support the unloading activity, whether it is being performed to allow
further processing or disposal of the spent fuel or it is required as
part of the response to an unplanned event or condition, while
preventing gross ruptures of 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 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 by the NRC
staff during a special inspection conducted from January 24 through May
11, 1995. In addition to the review of the licensee's facility and
procedures, the NRC inspectors observed preoperational testing that the
licensee was required to perform before loading casks with spent fuel
assemblies. The inspection findings are documented in NRC Inspection
Report 50-282/95002; 50-306/95002; 72-10/95002(DRP), dated 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 failing 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 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 50-282/95002; 50-306/95002; 72-10/
95002(DRP), the NRC staff concluded that as of May 1995, the licensee
had corrected the identified deficiencies and was ready to safely load
and, if necessary, unload spent nuclear fuel in TN-40 casks.
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 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, concerns identified during reviews and inspections, and
other issues related to loading and unloading procedures. The working
group completed its reviews in April 1996. The concerns related to
unloading procedures reviewed by the working group were found to
involve either (1) isolated occurrences 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
individual licensees. 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 included in NRC Inspection
[[Page 47229]]
Procedure 60855, ``Operation of an ISFSI,'' issued February 1, 1996,
states:
For unloading activities, attention should be paid to how the
licensee has prepared to deal with the potential hazards associated
with that task. Some potential issues may include: the radiation
exposure associated with drawing and analyzing a sample of the
canister's potentially radioactive atmosphere; steam flashing 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.
To address those ISFSIs that began operation before the
improvements in the NRC's review and inspection guidance, the staff
performed audits or inspections of those licensee 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 performed
as part of the inspection documented in NRC Inspection Report 50-282/
95002; 50-306/95002; 72-10/95002(DRP) adequately addressed the concerns
included in the NRC action plan.
III. Discussion
The Petition requests four actions by the NRC on the basis of the
contention 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 72.122(l)
In support of the Petition's contention that the licensee 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 concern that only portions of
the licensee's 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 concerns 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 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 staff agrees with the Petitioners' premise that actually unloading
a storage cask would likely result in licensees learning lessons that
could result in additional enhancements to unloading procedures, the
staff does not agree that an actual demonstration of the unloading
procedure at Prairie Island is warranted. In addition to the staff's
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 materials. 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; actual loading
and unloading of transportation casks; loading of storage casks;
handling of spent fuel assemblies under various conditions; and
performing relevant maintenance and engineering activities associated
with reactor facilities.
Regarding the Petitioners' concerns pertaining to the dates of the
submittal of 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; 72-10/95002(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 fuel loading
satisfied the preoperational license condition to implement written
procedures before loading spent nuclear fuel into a TN-40 cask. This
matter does not, therefore, represent a violation of NRC requirements
or introduce concerns pertaining to the technical adequacy of the
unloading procedure.
The Petitioners identified several concerns pertaining to the lack
of specific guidance in the unloading procedure to address a scenario
in which significant fuel degradation occurs 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 radioactive 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 Petitioners' legitimate concerns regarding the potential
difficulties in retrieving failed fuel from dry storage casks, 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. 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
[[Page 47230]]
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
ISFSI 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 concerns 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.
(a) Failed Fuel Considerations
As previously discussed, the NRC staff has accepted that procedures
developed by 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; 72-10/95002(DRP), the licensee had
originally 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 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 cladding 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
instructs 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 radioactive materials to parts of the reactor
facility, including the spent fuel pool, that may occur if fuel
assemblies degrade during storage. The NRC staff believes, however,
that the Petitioners have identified valid concerns 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 licensee to
address these and other problems that may occur in the unlikely event
that fuel assemblies that have degraded during storage need to be
unloaded from dry storage casks.
(b) Venting of Radioactive Gases
The possible need to vent radioactive gases from a cask is among
the issues that the 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 storage.
As with the concern regarding the contamination of the spent fuel pool,
the need to vent the cask while minimizing the radiological
consequences of unloading a cask containing failed fuel is 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 being 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
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 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 systems and additional
radiation monitoring equipment being used to minimize the release of
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-up
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 introduction of spent fuel pool water to the cask cavity
and stored fuel assemblies. The procedure instructs personnel to
continuously monitor the temperature and pressure instrumentation
installed on the vent connection and to stop pumping water if the
pressure exceeds 10 psig or the temperature exceeds 240 deg.F. In the
staff's 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 fuel is gradual and,
thereby, prevent safety problems that could hypothetically result from
damage to the
[[Page 47231]]
cask or the fuel assemblies because of stresses induced by a poorly
controlled addition of cooling water from the spent fuel pool.
The Petitioners expressed concerns 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 deg.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 deg.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' concern regarding the
need to post hazard warnings 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' concerns 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
concerns 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.
Item 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 in 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.1
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\1\ 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.
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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; 72-10/
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 staff's concern 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 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
Hearings and Allow Petitioners to Participate in the Proceedings
The licensee has provided the NRC with the unloading procedure,
including Revision 2, 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 this Petition, the
Petitioners are encouraged to continue their interactions with the NRC
staff regarding concerns or questions about the operation of the
Prairie Island Nuclear Generating Plant or the Prairie Island ISFSI.
Item 4: Update the Technical Specifications for the Prairie Island
ISFSI to Incorporate Mandatory Unloading Procedure Requirements
The TS for ISFSIs are required, by 10 CFR 72.44, to include
requirements in the following categories:
(1) Functional and operating limits and monitoring instruments and
limiting control settings;
(2) Limiting conditions;
(3) Surveillance requirements;
(4) Design features; and
(5) 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 typical in this respect since neither 10 CFR 72.44 nor the
associated regulatory guidance documents specify that technical
specifications should include special requirements for the unloading
procedure.2 Instead, the functional and operating limits,
limiting conditions, administrative controls, and other requirements
included in the TS for the Prairie Island ISFSI 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
[[Page 47232]]
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.
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\2\ 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 Format and Content for a Topical 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.
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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 Petitioners' concerns.
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 casks (see
10 CFR 72.122, 10 CFR 72.150, and 10 CFR 72.152). For example, the NRC
inspection documented in Inspection Report 50-282/95002; 50-306/95002;
72-10/95002(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 developed 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 reflect such an oversimplified representation. The
unloading procedure for the dry storage casks at Prairie Island was
inspected by the NRC staff and, as previously discussed, was ultimately
found to provide adequate guidance to control the unloading process.
IV. Conclusion
For the reasons described above, the NRC has determined that no
adequate basis exists for granting the Petitioners' request for
suspension of Northern 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' concerns 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 29th day of August 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-23696 Filed 9-5-97; 8:45 am]
BILLING CODE 7690-01-P
