[Federal Register Volume 63, Number 226 (Tuesday, November 24, 1998)]
[Notices]
[Pages 64973-64976]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-31336]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-220]
Niagara Mohawk Power Corporation; Notice of Consideration of
Issuance of Amendment to Facility Operating License, Proposed No
Significant Hazards Consideration Determination, and Opportunity for a
Hearing
The U.S. Nuclear Regulatory Commission (the Commission or NRC) is
considering issuance of an amendment to Facility Operating License No.
DRP-63 issued to Niagara Mohawk Power Corporation (NMPC or the
licensee) for operation of Nine Mile Point Nuclear Station, Unit 1
(NMP1), located in the town of Scriba, Oswego County, New York.
The proposed amendment would change Technical Specification (TS)
5.5, ``Storage of Unirradiated and Spent Fuel,'' for NMP1. The changes
would reflect a planned modification to increase the number of fuel
assemblies that can be stored in the spent fuel pool from 2776 to 4086.
The changes would also delete an erroneous reference within TS 5.5 to
10 CFR 70.55 for calculational methods approved by the Commission
involving special arrays.
Before issuance of the proposed license amendment, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment
request involves no significant hazards consideration. Under the
Commission's regulations in 10 CFR 50.92, this means that operation of
the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
The operation of NMP1, in accordance with the proposed
amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Analysis of issues concerning the expanded spent fuel pool
storage capacity modification has considered the following potential
scenarios:
1. A spent fuel assembly drop in the spent fuel pool.
2. Loss of spent fuel pool cooling flow.
3. A seismic event.
4. A cask drop in the spent fuel pool.
5. An accidental drop of a rack module during construction
activity in the pool.
The probability that any of the first four scenarios in the
above list can occur is not significantly increased by the proposed
Technical Specification changes and the associated modification
activities. Spent fuel pool activities such as fuel assembly
movement as well as Spent Fuel Pool Cooling System operation will
continue to be performed in accordance with approved plant
procedures. A cask drop into the pool is considered an unlikely
event based on the design/maintenance of the main hoist, the
controlled cask movement path and the cask drop protection system
(hydraulic guide cylinder). None of these features are affected by
the proposed change. Concerning installation activities, whether
conducted during power operation or shutdown, the reactor building
crane will be utilized for handling all heavy loads (i.e., old and
new racks) during the reracking operation. The main hoist is
equipped with a redundant hoisting system which will prevent the
dropping of heavy loads in the event that a cable or other critical
part of the main hoist equipment should fail. Operability of the
cranes will be checked and verified before the re-racking operation.
All lift rigging and the refueling crane/hoist system will be
inspected and all heavy load lifts will comply with NUREG-0612,
``Control of Heavy Loads at Nuclear Power Plants,'' per plant
procedures. Accordingly, the probability of a heavy load drop will
not significantly increase.
Therefore, the proposed modification and associated Technical
Specification changes do not involve a significant increase in the
probability of an accident previously evaluated.
UFSAR [Updated Final Safety Analysis Report] Section 15.c.3,
``Refueling Accident,'' discusses the accident in which a fuel
bundle is accidently dropped onto the top of the core during
refueling operations and the subsequent radiological effects. Fuel
assembly density in the core is essentially equivalent to that of
the assemblies stored in the replacement spent fuel racks.
Accordingly, the consequence of a fuel assembly dropped on the core
(as analyzed in UFSAR Section 15.c.3), is not significantly
[[Page 64974]]
increased. Also, analysis shows that such an accident will not
distort the racks sufficiently to impair their functionality and the
minimum subcriticality margin, keff [neutron
multiplication factor] [less than or equal to] 0.95, will be
maintained. Thus, the consequences of such an accident remain
acceptable and are not greater than those of previously evaluated
accidents.
The consequences of a loss of spent fuel pool cooling have been
evaluated and found acceptable. In the unlikely event that all
pooling cooling is lost, sufficient time is available for the
operators to re-establish cooling before the onset of pool boiling.
Also, the consequences of a design basis seismic event have been
evaluated and found acceptable. The new and the existing racks have
been analyzed in their new configuration and found safe and impact-
free during seismic motion. The structural capability of the pool
will not be exceeded under dead weight, thermal, and seismic loads
and the reactor building and the crane structure will retain the
necessary safety margins during a seismic event. Thus, the
consequences of a seismic event are not significantly increased.
Movements of heavy loads over the pool will continue to comply
with applicable guidelines (e.g., NUREG-0612) and procedures. As
previously mentioned, no heavy loads (e.g., racks, casks) will be
transported over any region of the spent fuel pool containing fuel.
The consequences of an accidental drop of a rack module into the
pool during reracking activities have been evaluated indicating that
very limited damage to the liner could occur. Therefore, the
consequences of a heavy load drop are not increased.
During rack removal and installation activities, interim
configurations will exist (i.e., various combinations of old and new
racks). These combinations have been evaluated and indicate that no
thermal-hydraulic, criticality and structural concerns exist.
The last paragraph in Section 5.5 states that calculations for
keff values have been based on methods approved by the
NRC covering special arrays (10 CFR 70.55). 10 CFR 70.55,
``Inspections,'' discusses inspections of special nuclear material
and the premises and facilities where special nuclear material is
used; not methods used to determine keff. Therefore, this
is an inaccurate reference. Also, although the NRC does review and
approve our methods to determine keff (as part [of] the
Technical Specification Amendment approval process) this information
is not considered critical design feature information. Accordingly,
it does not belong in Section 5.0, ``Design Features,'' of the
Technical Specifications. Based on the above, deletion of this
paragraph will not have any adverse affect on safety and will
eliminate any potential confusion involving the reference to 10 CFR
70.55.
Therefore, the proposed changes do not significantly increase
the consequences of any accident previously evaluated.
The operation of NMP1, in accordance with the proposed
amendment, will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed modification activities and associated Technical
Specification amendment does not introduce any new modes of plant
operation or accident precursors which could initiate a new or
different kind of accident, affect the operation or function of any
equipment necessary for the safe operation or shutdown of the plant,
or involve any changes to plant operating parameters. The only
physical alterations of plant configuration will involve the removal
of currently installed non-poison and Boraflex spent fuel racks and
the installation of new high density Boral racks. Heavy load
movements (i.e., the old and new racks, casks) will continue to be
performed in accordance with NUREG-0612. Accordingly, a drop of
heavy loads onto spent fuel during and following installation
activities need not be considered. As previously discussed,
installation of the new racks does not constitute a thermal-
hydraulic, criticality or structural concern. Therefore, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The operation of NMP1, in accordance with the proposed
amendment, will not involve a significant reduction in a margin of
safety.
The proposed modification activities and associated Technical
Specification Amendment involves replacing the currently install
non-poison flux trap and Boraflex storage racks with new high
density Boral racks. The proposed Technical Specification changes
will not reduce the equipment required by Technical Specifications,
affect any Technical Specification system setpoints, or adversely
affect the ability of plant equipment to respond to an accident.
The design and technical considerations applied to the reracking
modification included addressing the following areas:
1. Nuclear criticality considerations.
2. Thermal-hydraulic considerations.
3. Mechanical, material and structural considerations.
Concerning criticality considerations, the replacement high
density spent fuel storage racks are designed to assure that the
neutron multiplication factor ( keff ) is equal to or
less than 0.95 with the racks fully loaded with fuel of the highest
anticipated reactivity and the pool flooded with unborated water at
a temperature corresponding to the highest reactivity. The maximum
calculated reactivity includes a margin for uncertainty in
reactivity calculations and in mechanical tolerances, statistically
combined, such that the true keff will be equal to or
less than 0.95 with a 95% probability at a 95% confidence level.
Reactivity effects of abnormal and accident conditions have also
been evaluated to assure that under credible abnormal conditions,
the reactivity will be less than the limiting design basis value.
Accordingly, the proposed change does not involve a significant
reduction in a margin of safety in that the existing racks maintain
a keff of less than 0.95.
Amendment No. 54 to the NMP1 [Operating License which changed
the] Technical Specifications, dated February 1, 1984, increased the
spent fuel storage capacity to the current maximum of 2776
assemblies. In [its] Safety Evaluation, Section 2.4, ``Spent Fuel
Pool Cooling Considerations,'' the NRC indicated acceptance of
NMPC's thermal-hydraulic analysis based on: (1) with the maximum
normal heat load assumed and one cooling train in operation, pool
water is calculated to 125 degrees F which is below the 140 degrees
F limit recommended in Standard Review Plan (SRP) Section 9.1.3; and
(2) with the maximum abnormal heat load assumed and two cooling
trains operating, the maximum pool temperature is calculated to be
below 124 degrees which is below the boiling temperature limit set
forth in SRP Section 9.1.3.
The SRP requires that with a maximum normal heat load and a
single failure, pool temperatures should be kept below 140 degrees F
and that with an abnormal heat load, pool temperatures should be
kept below boiling. For the abnormal heat load case, consideration
of a single failure is not required. The analysis provided in
Section 5, Attachment C of this submittal [the licensee's May 15,
1998] indicates how the proposed change meets the requirements of
the SRP and, accordingly, that no significant decrease in a margin
of safety occurs.
In SRP 9.1.3, a normal spent fuel pool heat load is considered
to be a core shuffle. NMPC has evaluated the core shuffle using the
SRP guidance as Case 1, in previously referenced Section 5 of
Attachment C. This evaluation indicates that a maximum pool
temperature of 119 degrees F will be reached, thereby meeting the
SRP maximum temperature requirement of 140 degrees F. Because a
``normal heat load'' now potentially involves a full core offload,
NMPC has also reviewed this discharge scenario (Case 3, Section 5)
as a normal case and therefore assumed a single failure. As
delineated in Case 3, calculations will be performed to determine
the days after reactor shutdown when all assemblies can be
transferred to the pool, as a function of reactor building cooling
water temperatures, such that a 140 degrees F bulk pool temperature
will not be exceeded. Therefore, the SRP bulk pool temperature limit
of 140 degrees F for a maximum normal heat load (both shuffle and
full core offload) will not be exceeded.
The SRP also requires that for an abnormal maximum heat load
(emergency condition), without a single failure, that pool
temperatures should be maintained below boiling. Using the
guidelines provided in the SRP, calculations were performed that
found the maximum pool temperature to be 135 degrees F which is well
below the SRP criteria (Case 2).
The mechanical, material, and structural design of the spent
fuel racks is in accordance with applicable portions of NRC's
position in ``OT Position for Review and Acceptance of Spent Fuel
Storage and Handling Applications,'' dated April 14, 1978 (as
modified January 18, 1979), as well as other applicable NRC guidance
and industry codes. The primary safety function of the spent fuel
racks is to maintain the fuel assemblies in a safe configuration
through
[[Page 64975]]
normal and abnormal loading conditions. Abnormal loadings that have
been evaluated with acceptable results include the effect of an
earthquake and the impact due to the drop of a fuel assembly. The
rack materials used are compatible with the fuel assemblies and the
environment in the spent fuel pool. The structural design for the
new racks provides tilting, deflection, and movement margins such
that the racks do not impact each other or the spent fuel pool walls
in the active fuel region during the postulated seismic events.
Also, the spent fuel assemblies themselves remain intact and no
criticality concerns exist. In addition, the structural adequacy of
the spent fuel pool was demonstrated.
During rack removal and installation activities, interim
configurations will exist (i.e., various combinations of old and new
racks). These combinations have been evaluated and indicate that no
thermal-hydraulic, criticality and structural concerns exist.
Therefore, the proposed change will not result in a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based upon
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
By December 24, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in such proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at 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 Reference and Documents Department,
Penfield Library, State University of New York, Oswego, New York 13126.
If a request for a hearing and petition for leave to intervene is filed
by the above date, the Commission or an Atomic Safety and Licensing
Board, designated by the Commission or by the Chairman of the Atomic
Safety and Licensing Board Panel, will rule on the request and
petition; and the Secretary or the designated Atomic Safety and
Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
A request for a hearing and a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to Mr. Mark J. Wetterhahn, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502, attorney for the
licensee.
[[Page 64976]]
Untimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(l)-(v) and 2.714(d).
Pursuant to the Commission's regulations, 10 CFR 2.1107, the
Commission hereby provides notice that this is a proceeding on an
application for a license amendment falling within the scope of section
134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C. 10154.
Under section 134 of the NWPA, the Commission, at the request of any
party to the proceeding, must use hybrid hearing procedures with
respect to ``any matter which the Commission determines to be in
controversy among the parties.''
The hybrid procedures in section 134 provide for oral argument on
matters in controversy, preceded by discovery under the Commission's
rules and the designation, following argument of only those factual
issues that involve a genuine and substantial dispute, together with
any remaining questions of law, to be resolved in an adjudicatory
hearing. Actual adjudicatory hearings are to be held on only those
issues found to meet the criteria of section 134 and set for hearing
after oral argument.
The Commission's rules implementing section 134 of the NWPA are
found in 10 CFR Part 2, Subpart K, ``Hybrid Hearing Procedures for
Expansion of Spent Fuel Storage Capacity at Civilian Nuclear Power
Reactors'' (published at 50 FR 41662 dated October 15, 1985). Under
those rules, any party to the proceeding may invoke the hybrid hearing
procedures by filing with the presiding officer a written request for
oral argument under 10 CFR 2.1109. To be timely, the request must be
filed within ten (10) days of an order granting a request for hearing
or petition to intervene. The presiding officer must grant a timely
request for oral argument. The presiding officer may grant an untimely
request for oral argument only upon a showing of good cause by the
requesting party for the failure to file on time and after providing
the other parties an opportunity to respond to the untimely request. If
the presiding officer grants a request for oral argument, any hearing
held on the application must be conducted in accordance with the hybrid
hearing procedures. In essence, those procedures limit the time
available for discovery and require that an oral argument be held to
determine whether any contentions must be resolved in an adjudicatory
hearing. If no party to the proceeding timely requests oral argument,
and if all untimely requests for oral argument are denied, then the
usual procedures in 10 CFR Part 2, Subpart G apply.
For further details with respect to this action, see the
application for amendment dated May 15, 1998, as supplemented September
25 and October 13, 1998, which are available for public inspection at
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 Reference and Documents Department, Penfield Library,
State University of New York, Oswego, New York 13126.
Dated at Rockville, Maryland, this 18th day of November 1998.
For the Nuclear Regulatory Commission.
Darl S. Hood,
Senior Project Manager, Project Directorate I-1, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 98-31336 Filed 11-23-98; 8:45 am]
BILLING CODE 7590-01-P
