[Federal Register Volume 62, Number 46 (Monday, March 10, 1997)]
[Notices]
[Pages 10882-10885]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-5852]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-255, 50-266/301, 50-313/368, 72-5, 72-7, 72-13]
Consumers Power Company, Palisades Nuclear Plant, Wisconsin
Electric Power Company, Point Beach Nuclear Plant, Units 1 and 2,
Entergy Operations, Inc., Arkansas Nuclear One, Units 1 and 2; 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 November 17, 1995, filed Ms. Fawn Shillinglaw (Petitioner) under
Section 2.206 of Title 10 of the Code of Federal Regulations (10 CFR
2.206). The Petition requested that the NRC prohibit loading of spent
nuclear fuel into VSC-24 dry storage casks at any nuclear site until
the multi-assembly sealed basket (MSB) #4 at the Palisades Nuclear
Plant is unloaded and the unloading process is evaluated.
The Director of the Office of Nuclear Reactor Regulation has
determined that Petition should be denied for the reasons stated in the
``Director's Decision Under 10 CFR 2.206'' (DD-97-05), 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.
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 review of the decision within that time.
Dated at Rockville, Maryland, this 4th day of March 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 November 17, 1995, Ms. Fawn Shillinglaw (Petitioner) filed a
Petition pursuant to Section 2.206 of Title 10 of
[[Page 10883]]
the Code of Federal Regulations (10 CFR 2.206) requesting that the U.S.
Nuclear Regulatory Commission (NRC) take action to prohibit loading of
VSC-24 casks at any nuclear site until the multi-assembly sealed basket
(MSB) #4 at the Palisades plant has been unloaded and the experience
evaluated for potential safety improvements. In addition to Consumers
Power Company, the licensee for Palisades, other licensees that use the
VSC-24 cask system are Wisconsin Electric Power Company at its Point
Beach Nuclear Plant, Units 1 and 2, and Entergy Operations, Inc., at
Arkansas Nuclear One, Units 1 and 2.
The Petition has been referred to me pursuant to 10 CFR 2.206. The
NRC letter to you dated January 18, 1996, acknowledged receipt of the
Petition. Notice of receipt was published in the Federal Register on
January 25, 1996 (61 FR 2269).
On the basis of the NRC staff's evaluation of the issues and for
the reasons given below, the Petitioner's request is denied.
II. Background
NRC regulations contain a general license that authorizes nuclear
power plants licensed by the NRC to store spent nuclear fuel at the
reactor site in storage casks approved by the NRC. (See 10 CFR Part 72,
Subpart K.) In regard to dry cask storage of spent nuclear fuel at
Palisades, Point Beach, and Arkansas Nuclear One, the licensees opted
to use the VSC-24 Cask Storage System designed by Sierra Nuclear
Corporation. The VSC-24 Cask Storage System was added to the list of
NRC certified casks in May 1993 (58 FR 17948). The associated
certificate of compliance, Certificate Number 1007, specifies the
conditions for use of VSC-24 casks under the general license provisions
of 10 CFR Part 72. Section 1.1.2, ``Operating Procedures,'' in the
certificate of compliance for the VSC-24 casks requires that licensees
prepare an operating procedure related to cask unloading. Specifically,
the condition states--
Written operating procedures shall be prepared for cask
handling, loading, movement, surveillance, and maintenance. The
operating procedures suggested generically in the SAR [safety
analysis report] are considered appropriate, as discussed in Section
11.0 of the SER [safety evaluation report], and should provide the
basis for the user's written operating procedures. The following
additional written procedures shall also be developed as part of the
user operating procedures:
1. A procedure shall be developed for cask unloading, assuming
damaged fuel. If fuel needs to be removed from the multi-assembly
sealed basket (MSB), either at the end of service life or for
inspection after an accident, precautions must be taken against the
potential for the presence of oxidized fuel and to prevent
radiological exposure to personnel during this operation. This
activity can be achieved by the use of the Swagelok valves, which
permit a determination of the atmosphere within the MSB before the
removal of the structural and shield lids. If the atmosphere within
the MSB is helium, then operations should proceed normally, with
fuel removal, either via the transfer cask or in the pool. However,
if air is present within the MSB, then appropriate filters should be
in place to permit the flushing of any potential airborne
radioactive particulate from the MSB, via the Swagelok valves. This
action will protect both personnel and the operations area from
potential contamination. For the accident case, personnel protection
in the form of respirators or supplied air should be considered in
accordance with the licensee's Radiation Protection Program.
In July 1994, the licensee for Palisades discovered radiographic
indications of possible defects in a weld in MSB #4. MSB #4 had been
loaded with spent fuel earlier that month and placed inside a
ventilated concrete cask on the independent spent fuel storage
installation (ISFSI) storage pad. The licensee evaluated the flaw
indications and determined that the MSB continued to meet its design
basis and was capable of safely storing spent fuel for the duration of
the certificate (20 years). Nevertheless, the licensee stated that MSB
#4 would be unloaded to support additional inspections and evaluations
related to its future use. 1 In preparation for the unloading of
MSB #4, the licensee reviewed the unloading procedure issued in May
1993 (Revision 0) and identified several technical deficiencies. A
revision of the unloading procedure (Revision 1) was subsequently
developed to resolve the identified technical deficiencies. The revised
unloading procedure is the subject of an ongoing NRC inspection.2
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\1\ The unloading of MSB #4 was originally planned for several
months after the discovery of the radiographic indications of
possible weld defects in July 1994. However, the unloading has been
delayed several times and in its letter of January 17, 1997, the
licensee informed the NRC staff that the unloading has been
postponed until the fuel in MSB #4 can be reloaded into a certified
storage and transportation cask. The licensee also indicated it
intends to pursue development and licensing of such a cask, has
solicited and received bids from vendors, and plans to award a
contract before the end of the first quarter of 1997.
\2\ In regard to the original (Revision 0) unloading procedure
at Palisades, the NRC staff concluded that, had the licensee
attempted to unload a cask using the original unloading procedure,
the licensee would have needed to suspend activities at one or more
times during the unloading process in order to implement revisions
to the procedure. The NRC staff found that this was a violation of
requirements that all activities affecting quality be prescribed by
procedures appropriate for the circumstances and that procedures are
reviewed for adequacy. However, given the limited safety
significance of the procedural deficiencies and the fact that the
licensee identified and corrected the deficiencies, the NRC
dispositioned the violation as a Non-Cited Violation in accordance
with the NRC Enforcement Policy. (See NRC Inspection Report 50-255/
96014 and Director's Decision 97-01.)
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Through inspections at Palisades and other facilities, the NRC
staff identified a number of concerns regarding licensees' procedures
for unloading spent fuel from dry storage casks. The NRC staff
identified examples of procedural inadequacies and quality assurance
shortcomings experienced during preoperational tests and actual cask
loading operations at several facilities. In addition, the staff
observed that some unloading procedures implemented by licensees
neglected to consider contingencies and assumptions on 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 address these concerns, the following item titled ``Cask
Loading and Unloading,'' was included in the NRC dry cask storage
action plan implemented in July 1995. 3
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\3\ Action plans are used by the NRC staff to manage the
resolution of significant generic issues. Such plans are prepared
when the anticipated resources that will be required to resolve
generic or potentially generic issues exceed certain thresholds or
when the NRC staff determines that an action plan would improve its
efficiency and effectiveness.
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Issue: Cask Loading and Unloading
As licensees have implemented their ISFSI plans, several issues
have been identified related to the loading and unloading of casks.
Loading issues have centered on procedural inadequacies and quality
assurance shortcomings. The unloading procedures developed by
licensees tend to be simplistic. This has resulted in neglecting to
consider contingencies and assumptions on failed fuel, air sampling
techniques, disassembly requirements, design problems, and radiation
protection requirements. The importance of these procedures should
be emphasized to licensees, and technical issues related to
unloading problems resolved. This issue should also be addressed for
shipping casks.
The NRC action plan developed for dry cask storage was formulated
to manage the resolution of a variety of technical and process issues
associated with the expanding use of that technology for the storage of
spent nuclear fuel. The 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.
[[Page 10884]]
To implement the 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
which were addressed by incorporating remedial measures into ongoing
staff activities, such as the preparation of revised inspection
procedures or other guidance documents.
In May 1996, an event occurred at the Point Beach plant involving
the ignition of hydrogen gas during the loading of a VSC-24 cask.4
Completion of the NRC inspection of the revised unloading procedure for
Palisades was postponed following the event at Point Beach in order to
allow licensees and the NRC staff to identify the cause of the hydrogen
ignition and implement appropriate corrective actions. Following the
event, the NRC issued confirmatory action letters (CALs) to those
licensees using or planning to use VSC-24 casks for the storage of
spent nuclear fuel (i.e., licensees for Point Beach, Palisades, and
Arkansas Nuclear One). The CALs documented the licensees' commitments
not to load or unload a VSC-24 cask without resolution of material
compatibility issues identified in NRC Bulletin 96-04, ``Chemical,
Galvanic, or Other Reactions in Spent Fuel Storage and Transportation
Casks,'' and subsequent confirmation of corrective actions by the NRC.
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\4\ On May 28, 1996, a hydrogen gas ignition occurred during the
welding of the shield lid on a VSC-24 cask at the Point Beach
Nuclear Plant. The hydrogen was formed by a chemical reaction
between a zinc-based coating (Carbo Zinc 11) and the borated water
in the spent fuel pool.
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On December 3, 1996, the NRC staff informed the licensee for
Arkansas Nuclear One that it had completed its reviews and inspections
associated with that facility and found that the licensee had
satisfactorily completed the commitments documented in the CAL. Shortly
thereafter, the licensee initiated cask-loading activities. The review
of responses to the bulletin related to Palisades and Point Beach is
ongoing and cask operations at those facilities continue to be limited
by the licensees' commitments described in CALs.
III. Discussion
In support of the Petitioner's request that VSC-24 casks not be
loaded until MSB #4 at Palisades has been unloaded and the unloading
process has been evaluated, the Petitioner cites the action plan
prepared by the NRC staff that included the staff's observation that
some unloading procedures developed by licensees tended to be
simplistic. The Petitioner asserts that because problems are discovered
through experience, the proper way to unload casks will not be known
until a cask is actually unloaded. The Petitioner also claims that the
unloading procedures should not be left to the licensees to develop and
implement but should be the subject of detailed NRC evaluations.
The NRC staff's concerns 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 Palisades, the licensee's revised procedure
addressed many of the generic staff activities on cask unloading and is
currently the subject of a thorough NRC inspection that will be
completed in the near future.
To fulfill some of the goals included in the 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. On
the basis that these discussions with the industry and other staff
actions had conveyed important operating experiences to NRC licensees,
the staff deferred issuance of an NRC information notice on the subject
of loading and unloading of dry storage casks. 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 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, Draft Report for Comment,'' issued in
February 1996 and will be included in the final version of the standard
review plan that is currently being prepared. The revised guidance
documents ensure that recent and future reviews will address the
adequacy of unloading procedures developed by licensees.
The NRC staff also reviewed the inspection history for existing
ISFSIs to determine if unloading procedures were reviewed with due
consideration given to the potential complications that may arise
during the unloading process. The NRC 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. In regard to the users of
the VSC-24 cask system, inspections of unloading procedures at Arkansas
Nuclear One (NRC Inspection Report 50-313/96-16; 50-368/96-16; 72-13/
96-01 and Notice of Violation, dated July 31, 1996) and Point Beach
(NRC Inspection Report 50-266/95011; 50-301/95011, dated November 15,
1995) considered the concerns included in the NRC action plan.
As previously mentioned, the revised unloading procedure at
Palisades is the subject of an ongoing inspection, completion of which
was delayed as a result of the hydrogen ignition event at Point Beach.
The NRC inspection of the revised unloading procedure at Palisades is
being coordinated with the staff's review of the licensee's response to
NRC Bulletin 96-04 and is expected to be completed in the near future,
notwithstanding the licensee's decision to postpone unloading MSB #4
pending the availability of a certified storage and transportation
cask. 5 Further, the NRC has committed to State officials and
members of the public that the exit meeting for the inspection of the
revised unloading procedure at Palisades will be open to the public,
the meeting will be noticed sufficiently in advance to
[[Page 10885]]
allow interested parties to attend, and the NRC staff will allocate
time to discuss issues with the public following the meeting with the
licensee.
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\5\ The licensee for Palisades responded to NRC Bulletin 96-04
by letters dated August 19 and November 12, 1996. The NRC staff is
awaiting the licensee's response to a request for information that
was issued on February 12, 1997.
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The NRC staff agrees with the Petitioner that learning from
experience is an essential part of improving the safety of nuclear
power plant activities, including those associated with dry cask
storage of spent nuclear fuel. This principle is reflected in the
regulatory requirements pertaining to preoperational testing of dry
cask storage activities, as well as various provisions of NRC-approved
quality assurance programs. The issuance of Bulletin 96-04 and the CALs
for licensees using VSC-24 casks is another example of the NRC staff's
efforts to ensure that applicable operating experience is incorporated
into procedures at facilities licensed by the NRC. In this case, the
licensees using the VSC-24 cask revised procedures to address the
technical concerns identified after the event at Point Beach and agreed
to defer cask operations pending the NRC's review of responses to the
bulletin and confirmation of corrective actions.
As previously mentioned, the licensee for Arkansas Nuclear One
loaded VSC-24 casks following the NRC staff's determination that the
licensee had satisfactorily completed the commitments documented in the
CAL. On the basis of reviews and inspections performed to verify
corrective actions associated with the bulletin, in combination with
reviews performed for cask certification and previous inspections of
preoperational testing and other aspects of the licensee's dry cask
storage program, the NRC staff determined that the licensee for
Arkansas Nuclear One could perform either cask loading or unloading
operations without undue risk to the health and safety of the public or
its own personnel. The NRC staff, through reviews and inspections to
verify corrective actions associated with NRC Bulletin 96-04, must have
confidence in the procedures implemented by the licensee for Point
Beach before the NRC permits that licensee to resume loading or
unloading of VSC-24 casks. The staff must also obtain the necessary
confidence that the licensee for Palisades has implemented the
corrective actions related to NRC Bulletin 96-04 as well as the issues
included in the NRC action plan before permitting the licensee to
resume loading or unloading VSC-24 casks.
Thus, only after resolution of the issues identified in NRC
Bulletin 96-04 and other questions that may arise during the
inspections of the licensees' revised procedures at Point Beach and
Palisades, will the NRC permit them to unload casks. As part of its
review, the NRC staff will consider matters such as the dry-run
exercises licensees performed to verify key aspects of unloading
procedures, as well as licensees' actual experience in the loading and
unloading of transportation casks, loading of storage casks, handling
of spent fuel assemblies under various conditions, and performing
relevant maintenance and engineering activities associated with reactor
facilities. Given that the NRC staff will not permit unloading of any
casks unless it obtains reasonable assurance of each licensee's ability
to do so safely, the NRC does not have reason to require unloading of
MSB #4 at Palisades before allowing resumption of normal activities
under the general licenses at Arkansas Nuclear One, Point Beach, or
Palisades.
The Petitioner's request is, therefore, denied.
IV. Conclusion
The Petitioner requested that the NRC prohibit loading of VSC-24
casks at any nuclear site until MSB #4 at the Palisades plant has been
unloaded and the experience evaluated for potential safety concerns.
Each of the claims by the Petitioner has been reviewed. I conclude
that, for the reasons discussed above, no adequate basis exists for
granting Petitioner's request for suspension of the licensees' use of
the general licenses for dry cask storage of spent nuclear fuel at
Palisades, Point Beach, or Arkansas Nuclear One until the MSB at
Palisades has been unloaded and the experience evaluated for potential
safety improvements.
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 4th day of March 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-5852 Filed 3-7-97; 8:45 am]
BILLING CODE 7590-01-P
