[Federal Register Volume 62, Number 10 (Wednesday, January 15, 1997)]
[Notices]
[Pages 2182-2185]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-982]
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NUCLEAR REGULATORY COMMISSION
Docket No. 50-286
Power Authority of the State of New York; 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) is
considering issuance of an amendment to Facility Operating License No.
DPR-64 issued to the Power Authority of the State of New York for
operation of the Indian Point Nuclear Generating Station Unit No. 3
(IP3) located in Westchester County, New York.
The proposed amendment would revise the IP3 Technical
Specifications (TS) to allow the storage of fuel assemblies with
nominal enrichments up to 5.0 weight percent (w/o) Uranium-235 (U-235).
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:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of any accident
previously evaluated. This statement is based on an evaluation of
relevant hypothetical accident scenarios, the NRC's evaluation of
Westinghouse extended burnup fuel, and the criticality analysis of
the Indian Point 3 fresh and spent fuel pits.
Evaluation of Relevant Hypothetical Accident Scenarios
Increasing the enrichment of fuel stored in the spent fuel pit
will not increase the probability of occurrence of the following
hypothetical accident scenarios:
1. misload of a fuel assembly;
2. spent fuel assembly drop in the spent fuel pit;
3. spent fuel cask drop;
4. loss of spent fuel pit cooling system flow; or
5. seismic event.
1. Misload of a Fuel Assembly
Detailed instructions and administrative controls govern
refueling operations, precluding the misload of an assembly. The
proposed storage of extended burnup fuel will not result in these
administrative controls being relaxed in any manner. The probability
of inserting an assembly into the wrong location is not impacted by
the enrichment and burnup of the fuel. Consequently, the proposed
changes will not increase the probability of misloading a fuel
assembly.
2. Spent Fuel Assembly Drop in the Spent Fuel Pit
The probability of a spent fuel assembly drop in the spent fuel
pit is a function of the structural integrity of the fuel storage
building overhead crane and the integrity of the crane-assembly
coupling. The probability of such a drop is not affected by the
enrichment or burnup of the fuel. Therefore, the use and storage of
extended burnup fuel will not increase the probability of a fuel
assembly drop.
3. Spent Fuel Cask Drop
The probability of a spent fuel cask drop will not be affected
by the increased enrichment of the fuel. The probability of such an
event occurring is a function of the overhead crane's integrity,
which will not be affected by this amendment. In addition,
administrative controls are in place to preclude the occurrence of
such an event.
4. Loss of Spent Fuel Pit Cooling System Flow
A reevaluation of the Indian Point Unit 3 decay heat removal
analysis to address the storage of extended burnup fuel concluded
that the existing spent fuel pit cooling system is adequate to
handle the heat load associated with extended burnup fuel since any
incremental increase in decay heat for extended burnup fuel is more
than compensated for by the greater time interval between refueling
outages. In the unlikely event the cooling system should experience
a failure, adequate time is available to provide an alternate
cooling system, which is not affected by the fuel's enrichment. In
addition, an existing off normal operating procedure (ONOP) is
available to compensate for any postulated loss of spent fuel pit
cooling. Consequently, the storage of extended burnup fuel in the
spent fuel pit will not involve a significant increase in the
probability or consequences of a loss of cooling system flow event.
5. Seismic Event
The enrichment of the fuel has no effect on the probability of a
seismic event occurring. In support of Amendment 90 to Indian Point
3's Operating License, a seismic analysis of the spent fuel storage
racks was performed. This analysis, which was summarized in
[[Page 2183]]
Reference 3 [See application dated November 22, 1996] is still
applicable.
NRC Evaluation of Westinghouse Extended Burnup Fuel
Westinghouse's analysis of the use of extended burnup fuel is
documented in WCAP-10125 (Proprietary), ``Extended Burnup Evaluation
of Westinghouse Fuel''. On October 11, 1985, the NRC issued a Safety
Evaluation Report (SER) on this WCAP (Reference 2), which concluded
that: 1) fuel damage is not expected to occur as a result of normal
operation and anticipated operational occurrences (Condition I and
II events); 2) fuel damage during postulated accidents (Condition
III and IV events) would not be severe enough to prevent control rod
insertion when it is required; and 3) core coolability will always
be maintained, even after postulated accidents (Condition III and IV
events). These conclusions support the determination that the use of
extended burnup fuel will not increase the probability or
consequences of any accident previously evaluated.
The consequences from accidents involving extended burnup fuel,
both during operations and fuel handling, are evaluated in Reference
6 [See application]. This report, which was the basis for the NRC's
determination of no environmental impact, documents the amount of
radioactivity released from extended burnup fuel during an accident
may be greater than that released from lower burnup fuel. However,
the projected offsite dose incurred during accidents with extended
burnup fuel is still within 10 CFR 100 criteria. Reference 6 [See
application] concludes that since there is an order of magnitude
uncertainty in the risk estimates for accidents, any increased risk
from the increased fission products in extended burnup fuel is small
compared to the uncertainties associated with risk estimates.
Consequently, the proposed changes do not significantly increase the
consequences of any accident previously evaluated.
Criticality Analysis of the Indian Point 3 Fresh and Spent Fuel Pits
Westinghouse performed a criticality analysis of the Indian
Point 3 fresh and spent fuel storage racks to determine whether the
storage of Westinghouse 15x15 fuel assembly designs with nominal
enrichments up to 5.0 w/o U-235 would result in the effective
neutron multiplication factor, Keff, exceeding design and
licensing basis criticality limits. The analysis demonstrated that
these criteria would be met during design basis conditions using the
fuel storage configurations proposed in this submittal.
Although the analysis identified three scenarios which would
exceed the criticality limits, each of these scenarios are outside
the design and licensing basis, since they entail the occurrence of
two, independent, concurrent events. Specifically, the analysis
assumes the occurrence of the initiating accident event and the loss
of all soluble boron in the spent fuel pit water. However, the
analysis also documents that 700 ppm of soluble boron in the spent
fuel pit water will maintain Keff within acceptable limits. The
Indian Point Unit 3 spent fuel pit boron concentration is maintained
at a minimum of 1000 ppm during fuel handling operations, which is
more than adequate to offset the potential reactivity increases
incurred from even the most limiting criticality accident scenarios.
If credit for integral burnable neutron absorbers is taken, the
boron concentration to maintain Keff less than or equal to 0.95
is considerably reduced.
Consequently, as supported by the NRC's issuance of similar
license amendments to other plants whose criticality analyses have
identified similar issues, the proposed amendment does not
significantly increase the probability or consequences of any
accident previously evaluated.
The administrative changes proposed by this amendment request do
not involve a significant increase in the probability or
consequences of any accident previously evaluated as they do not
involve any plant hardware changes, nor do they change the way the
plant systems function.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any previously
evaluated?
Response:
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated. This
determination is based on the NRC's SER regarding Westinghouse
extended burnup fuel, Indian Point 3 decay heat removal analysis,
and spent fuel pit criticality analysis.
The only aspect of the plant that will be physically changed by
the proposed amendment will be the enrichment and burnup of the
fuel, which will not introduce any new fuel failure mechanisms.
While some characteristics of fuel performance change with extended
burnup, these considerations have been factored into the design of
the fuel. The NRC issued a Safety Evaluation Report (SER) regarding
the Westinghouse extended burnup fuel design on October 11, 1985
(Reference 2). In addition, Reference 6 [See application] documents
that each fuel vendor has adequately considered the performance of
extended burnup fuel to preclude the introduction of a new or
different type of fuel failure mechanism.
Two site specific evaluations demonstrate the storage of spent
and/or fresh extended burnup fuel will not introduce any new fuel
storage accidents at Indian Point Unit 3. First, the Authority has
verified the existing spent fuel pit cooling system can adequately
handle the heat load associated with extended burnup fuel. Second,
the criticality analysis performed by Westinghouse demonstrates the
criticality limits will continue to be satisfied during design basis
conditions. While three scenarios outside of the design basis have
been identified as potentially resulting in an increase in spent
fuel pit criticality, spent fuel pit soluble boron concentrations
are maintained sufficiently high to preclude even the most limiting
criticality scenarios from occurring. Consequently, the proposed
amendment will not create a new or different kind of accident from
any previously evaluated.
The administrative changes proposed by this amendment request do
not create the possibility of a new or different kind of accident
from any previously evaluated as the changes do not affect current
plant configuration or how the plant operates.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed changes do not involve a significant reduction in a
margin of safety. This determination is based on the fact that the
spent fuel pit racks are not being physically altered, the results
of the Indian Point 3 spent fuel pit criticality analysis, the spent
fuel pit decay heat analysis, and the NRC issuance of similar
amendments to other licensees.
The main safety function of the fresh and spent fuel racks is to
maintain the fuel assemblies in a safe configuration through all
normal and abnormal conditions. The proposed changes will not result
in any changes to the fresh and spent fuel racks or the manner in
which they perform. Thus, the margin of safety associated with the
fresh and spent fuel racks'' ability to physically maintain the fuel
in a safe configuration is not significantly reduced by the proposed
changes.
A criticality analysis was performed regarding the Indian Point
3 fresh and spent fuel storage racks' ability to store extended
burnup fuel within design and licensing basis criticality limits.
The analysis concludes during design basis conditions these limits
would not be violated. However, it identified three events outside
the design and licensing basis which would violate these limits.
Nevertheless, if credit is taken for the soluble boron in the spent
fuel pit water, criticality is adequately controlled even during
these three events. Consequently, as supported by the NRC issuance
of similar license amendments to other plants whose criticality
analyses have identified similar issues, the proposed amendment does
not involve a significant reduction in the margin of safety
associated with the control of criticality.
An evaluation was performed to address the spent fuel pit heat
load associated with the storage of extended burnup fuel. The
analysis concluded the existing spent fuel cooling system will
adequately dissipate the heat. Thus, there is no significant
reduction in the margin of safety with regards to spent fuel
cooling.
The administrative changes proposed by this amendment request do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
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.
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
[[Page 2184]]
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.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to
4:15 p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By February 14, 1997, 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 the 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 White Plains Public Library, 100 Martine
Avenue, White Plains, New York 10610. If a request for a hearing or
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/or 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.
A request for a hearing or 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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to S. Singh Bajwa: petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. 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.
Charles M. Pratt, 10 Columbus Circle, New York, New York 10019,
attorney for the licensee.
Nontimely 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
[[Page 2185]]
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)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment dated November 22, 1996, which is 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 White Plains Public Library, 100
Martine Avenue, White Plains, New York 10610.
Dated at Rockville, Maryland, this 9th day of January 1997.
For the Nuclear Regulatory Commission.
George F. Wunder,
Project Manager, Project Directorate 1-1, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-982 Filed 1-14-97; 8:45 am]
BILLING CODE 7590-01-P