[Federal Register Volume 62, Number 83 (Wednesday, April 30, 1997)]
[Notices]
[Pages 23502-23504]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-11121]
[[Page 23502]]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-244]
Rochester Gas and Electric 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) is
considering issuance of an amendment to Facility Operating License No.
DRP-18 issued to Rochester Gas & Electric Corporation for operation of
the R. E. Ginna Nuclear Power Plant located in Wayne County, New York.
The proposed amendment would revise the Ginna Station Improved
Technical Specifications (ITS) to reflect a planned modification to the
spent fuel pool (SFP) storage racks. Specifications associated with SFP
boron concentration, fuel assembly storage, and the maximum limit on
the number of fuel assemblies which can be stored in the SFP would be
revised.
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. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The design basis events considered for the spent fuel pool
include both external events and postulated accidents in the pool.
The external events considered are tornado missiles and seismic
events. The evaluation of the postulated impact of a tornado missile
is detailed in Sections 3, 4, and 6 of Reference 1 [see application
dated March 31, 1997]. The structural evaluation indicates that
there are no gross distortions of the racks or any adverse effects
upon plant structures or equipment. The radiological consequences of
this event indicate that offsite doses are ``well within'' the 10
CFR 100 limits.
The structural evaluation is detailed in Section 3 of Reference
1 [see application dated March 31, 1997]. Current state of the art
methods are used in the structural analysis. The evaluation of the
storage racks is based on a conservative interpretation of the ASME
[American Society of Mechanical Engineers] Boiler and Pressure
Vessel Code. The evaluation of the spent fuel pool is based on a
conservative interpretation of requirements set forth in the
American Concrete Institute, Code Requirements for Nuclear Safety
Related Concrete Structures, and American Institute of Steel
Construction, Specification for Structural Steel Buildings. The
spent fuel storage system was designed to meet all applicable
structural criteria for normal (Level A), upset (Level B), and
faulted (Level D) conditions as defined in NUREG-0900, SRP [Standard
Review Plan] 3.8.4, Appendix D. The following loadings were
considered: dead weight, seismic, thermal, stuck fuel assembly, drop
a fuel assembly, and tornado missile impact. Load combinations were
performed in accordance with SRP 3.8.4, Appendix D. Given the
evaluated seismic events, the changes in the final position of the
racks are small as compared to the initial position prior to the
seismic event. The maximum closure of gaps is such that no
significant changes in gaps result during any single seismic event.
Furthermore, the combined gap closures resulting from a combination
of 5 OBEs [Operating Basis Earthquakes] and 1 SSE [Safe Shutdown
Earthquake] show that there are no rack-to-rack or rack-to-wall
impacts. These evaluations conclude that under these postulated
events the stored fuel assemblies are maintained in a stable,
coolable geometry, and a subcritical configuration.
As described in the bases for LCO [Limiting Condition for
Operation] 3.7.12 and 3.7.13, the postulated accidents in the spent
fuel pool are divided into two categories. The first are those
involving a loss of cooling in the spent fuel pool. The thermal-
hydraulic analysis for the maximum expected decay heat loads is
described in Section 5 of Reference 1 [see application dated March
31, 1997]. The proposed modification does not change the
configuration of the available spent fuel cooling systems, the
limiting design conditions for maximum decay heat load which occurs
during a full core offload, or the existing requirement to maintain
pool temperature below 150 deg.F. Utilizing the three available
spent fuel cooling systems, Ginna Station maintains full redundancy
during high heat load conditions. The decay heat load to the spent
fuel pool is maintained within the capacity of the operating cooling
system by appropriately delaying fuel offload from the reactor.
Should a failure occur on the operating cooling system, the
resulting heat rates allow sufficient time to place a standby
cooling system in service before the pool design limit temperature
is exceeded. Increases in spent fuel pool temperature, with the
corresponding decrease in water density and void formation from
boiling, will result in a decrease in reactivity due to the decrease
in moderation effects. In addition, the analysis demonstrates that
the storage rack geometry and required fuel storage configurations
result in a Keff [less than or equal to] .95
assuming no soluble boron allowing for the potential of makeup to
the pool with unborated water.
The second category is related to the movement of fuel
assemblies and other loads above the spent fuel pool. The limiting
accident with respect to reactivity is the fuel handling accident
which is analyzed in Section 4 of Reference 1 [see application dated
March 31, 1997]. For both the incorrectly transferred fuel assembly
(placed in an unauthorized location) or a dropped fuel assembly, the
positive reactivity effects resulting are offset by the negative
reactivity from the required minimum soluble boron concentration.
The resulting Keff is shown to be less than 0.95. The
radiological consequences of a fuel assembly drop remain as
described in Section 15.7.3 of the UFSAR [Updated Final Safety
Analysis Report] and as discussed in Section 6 of Reference 1 [see
application dated March 31, 1997]. Loads in excess of a fuel
assembly and its handling tool are administratively prohibited from
being carried over spent fuel. There are no changes anticipated for
either the fuel handling equipment or the auxiliary building
overhead crane due to the proposed modification to the fuel storage
racks. The modification is scheduled for the Year 1998 to be
performed while Ginna Station is operating. Movement of heavy loads
around the spent fuel pool are controlled by the requirements of
NUREG-0612 and the regulatory guidelines set forth in NRC Bulletin
96-02 (see Section 3 of Reference 1).
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[see application dated March 31, 1997]. Spent fuel casks and storage
racks (during removal and installation) will be moved using the
auxiliary building crane and lifting attachments satisfying the
single failure proof criteria of NUREG-0554, obviating the need to
determine the consequences for this accident.
Based on the above, it is concluded that the proposed changes do
not significantly increase the probability or consequences of any
accident previously analyzed.
2. Operation in accordance with the proposed changes does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed modification does not alter the function of any
system associated with spent fuel handling, cooling, or storage. The
proposed changes do not involve a different type of equipment or
changes in methods governing normal plant operation. The additional
restrictions placed on the acceptable storage locations for spent
fuel are consistent with the type of restriction that previously
existed. The potential violation of these restrictions (incorrectly
transferred fuel assembly) are analyzed as discussed above. The
design, analysis, fabrication, and installation meet all the
appropriate NRC regulatory requirements, and appropriate industry
codes and standards.
Based on the above, the change does not create the possibility
of a new or different kind of accident from any previously analyzed.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in the margin of
safety.
The Licensing Report enclosed as Reference 1 [see application
dated March 31, 1997] addresses the following considerations:
nuclear criticality, thermal-hydraulic, and mechanical, material,
and structural. Results of these evaluations demonstrate that the
changes associated with the spent fuel reracking do not involve a
significant reduction in the margin of safety as summarized below:
Nuclear Criticality
The established regulatory acceptance criterion is that
Keff be less than or equal to 0.95, including all
uncertainties at the 95/95 probability/confidence level, under
normal and abnormal conditions. The methodology used in the
evaluation meets NRC requirements, and applicable industry codes,
standards, and specifications. In addition, the methodology has been
reviewed and approved by the NRC in recent nuclear criticality
evaluations. Specific conditions which were evaluated include
misloading of a fuel assembly, drop of a fuel assembly (shallow,
deep drops, and side drops), pool water temperature effects, and
movement of racks due to seismic events. Results described in
Section 4 of Reference 1 [see application dated March 31, 1997]
document that the criticality acceptance criterion is met for all
normal and abnormal conditions.
Thermal-Hydraulic
Conservative methods and assumptions have been used to calculate
the maximum temperature of the fuel and the increase of the bulk
pool water temperature in the spent fuel pool under normal and
abnormal conditions. The methodology for performing the thermal-
hydraulic evaluation meets NRC regulatory requirements. Results from
the thermal-hydraulic evaluation show that the maximum temperature
of the hottest fuel assembly, intact or consolidated canister, is
less than the temperature for nucleate boiling condition. The
effects of cell blockage on the maximum temperature of intact fuel
and consolidated canisters were evaluated. Results described in
Section 5 of Reference 1 [see application dated March 31, 1997] show
that adequate cooling of the intact or consolidated fuel is assured.
In all cases the existing spent fuel pool cooling system will
maintain the bulk pool temperature at or below 150 deg.F by
delaying core offload from the reactor.
Mechanical, Material, and Structural
The primary safety function of the spent fuel pool and the racks
is to maintain the spent fuel assemblies in a safe configuration
through all normal and abnormal loads. Abnormal loadings which have
been considered in the evaluation are: seismic events, the drop of a
fuel assembly, the impact of a tornado missile, a stuck assembly,
and the drop of a heavy load. The mechanical, material, and
structural design of the new spent fuel racks is in accordance with
NRC regulatory requirements (including the NRC OT Position dated
April 14, 1978, [NRC letter to all power reactor licensees dated
April 14, 1978] and addendum dated January 18, 1979), and applicable
industry standards. The rack materials are compatible with the spent
fuel pool environment and fuel assemblies. The material used as a
neutron absorber (borated stainless steel) has been approved by the
American Society for Testing and Materials (ASTM), and licensed
previously by the NRC for use as a neutron absorber at Indian Point
3, Indian Point 2, and Millstone 2. The structural evaluation
presented in Section 3 of Reference 1 [see application dated March
31, 1997] documents that the tipping or sliding of the free-standing
racks will not result in rack-to-rack or rack-to-wall impacts during
seismic events. The spent fuel assemblies will remain intact and the
criticality criterion of keff less than or equal to 0.95
is met.
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 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 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 May 30, 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 Rochester Public Library, 115 South
Avenue, Rochester, New York 14610. If a request for a hearing or
petition for leave to intervene is filed by the above date, the
[[Page 23504]]
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
Nicholas S. Reynolds, Winston & Strawn, 1400 L Street, NW., Washington,
DC 20005, 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 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 March 31, 1997, 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 Rochester Public Library, 115 South
Avenue, Rochester, New York 14610.
Dated at Rockville, Maryland, this 24th day of April 1997.
For the Nuclear Regulatory Commission.
Guy S. Vissing,
Senior Project Manager Project Directorate I-1, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-11121 Filed 4-29-97; 8:45 am]
BILLING CODE 7590-01-P