[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