[Federal Register Volume 60, Number 4 (Friday, January 6, 1995)]
[Notices]
[Pages 2160-2162]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-319]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-528, STN 50-529, and STN 50-530]
Arizona Public Service Company; Palo Verde Nuclear Generating
Station, Units 1, 2, and 3; Notice of Consideration of Issuance of
Amendment to Facility Operating Licenses, Proposed no Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of amendments to Facility Operating License Nos.
NPF-41, NPF-51, and NPF-74 issued to Arizona Public Service Company for
Operation of the Palo Verde Nuclear Generating Station, Units 1, 2, and
3, located in Maricopa County, Arizona.
The proposed amendments would change the refueling machine overload
cutoff limit from less than or equal to 1556 pounds to less than or
equal to 1600 pounds. The change is a consequence of the fuel assembly
weight increase which resulted from design and fabrication
improvements.
Before issuance of the proposed license amendments, 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 amendments 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:
Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed Technical Specification amendment to sections 3.9.6
and 4.9.6.1 provides a revised refueling machine hoist overload
cutoff limit that is appropriate for the increased weight of the
fuel assemblies. The increased weight of fuel assemblies results
from design and fabrication improvements such as denser fuel
pellets, laser welded GUARDIANTN grids, and laser welded spacer
grids. The weight of a fuel assembly is identified in the UFSAR as a
parameter in the analysis for a Fuel Handling Accident. The
radiological consequences of a Fuel Handling Accident were
reevaluated in order to incorporate fuel assembly design changes
including increases in the fuel assembly weight and increases of the
maximum fuel enrichment. The analysis used a fuel assembly enriched
to 4.3 weight percent and the power assigned to the assembly was
1.65 times the average power per assembly. The accident is assumed
to occur 100 hours after reactor shutdown and it is also assumed
that all 236 fuel rods fail. The resultant thyroid dose at the 2
hour exclusion area boundary is 71.5 rem which meets the Standard
Review Plan 15.7.4 limit of 75 rem. The conclusions for the
radiological consequences of a Fuel Handling Accident remain
consistent with the results in the Safety Evaluation Report. The
increased weight of the fuel assemblies was reviewed, separate from
this proposal, in accordance with the provisions of 10 CFR 50.59 and
found to be acceptable, as described above.
The increase in the refueling machine overload cutoff limit does
not impact the manner in which the refueling machine is operated or
the manner in which the fuel assemblies are engaged and lifted. The
overload cutoff limit is not a parameter used in the analysis of a
Fuel Handling Accident. The overload cutoff limit was incorporated
on the refueling machine hoist to protect the core internals and
pressure vessel from [[Page 2161]] possible damage in the event the
fuel assembly becomes mechanically bound as it is withdrawn from the
reactor vessel. The proposed overload cutoff limit was determined as
follows:
Overload Cut Off limit=(Hoist Wet Weight)+(Grapple Wet Weight)+(Max
Wet Fuel Weight)+90lbs.
Where:
(a) Hoist and Grapple Wet Weight=176 lbs.
(b) Maximum Wet Fuel Weight=1334 lbs.
The basis for the 90 pounds had two considerations: (1) to be
large enough to account for friction loads during fuel assembly
withdrawal; and, (2) to be small enough to ensure that while lifting
a minimum weight fuel assembly, the loads imposed on a mechanically
bound fuel assembly are below the design limit specified by the fuel
manufacturer. The maximum value for the existing overload cut off
limit was specified by the fuel manufacturer to be 1602 pounds.
The revised overload cut off limit does not decrease the factor
of safety for the refueling machine hoist below the Crane
Manufacturer's [sic] Association of America (CMAA) Standard 70
required value of 5/1.
Therefore, the proposed change for the refueling machine
overload cut off limit will not significantly increase the
probability or consequences of an accident previously evaluated and
will remain bounded by the accident analysis of Chapter 15 of the
Updated Final Safety Analysis Report (UFSAR).
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
The proposed Technical Specification amendment to Sections 3.9.6
and 4.9.6.1 would provide a revised refueling machine hoist overload
cut off limit that is appropriate for the increased weight of the
fuel assemblies. The increased weight of fuel assemblies results
from design and fabrication improvements such as denser fuel
pellets, laser welded GUARDIANTM grids, and laser welded spacer
grids. The fuel overload cut off limit was incorporated on the
refueling machine hoist to protect the core internals and pressure
vessel from possible damage in the event the fuel assembly becomes
mechanically bound as it is withdrawn from the reactor vessel. The
proposed overload cut off limit was determined as follows:
Overload Cut Off limit=(Hoist Wet Weight)+(Grapple Wet Weight)+(Max
Wet Fuel Weight)+90 lbs.
Where:
(a) Hoist and Grapple Wet Weight=176 lbs.
(b) Maximum Wet Fuel Weight=1334 lbs.
The basis for the 90 pounds had two considerations: (1) to be
large enough to account for friction loads during fuel assembly
withdrawal; and, (2) to be small enough to ensure that while lifting
a minimum weight fuel assembly, the loads imposed on a mechanically
bound fuel assembly are below the design limit specified by the fuel
manufacturer. The maximum value for the existing overload cut off
limit was specified by the fuel manufacturer to be 1602 pounds to
limit the potential for damage to the fuel assemblies.
The accident of concern related to the change in the refueling
machine overload cut off limit is the Fuel Handling Accident. This
accident occurs when a fuel bundle becomes disengaged from the
refueling machine grapple. The change of the refueling machine
overload cut off limit does not change the way in which the
refueling machine grapple engages the fuel assemblies. Since fuel
handling is the subject of change, no new or different kinds of
accidents are created.
Therefore, it can be concluded that the proposed change to
Sections 3.9.6 and 4.9.6.1 will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Standard 3--Does the proposed change involve a significant
reduction in a margin of safety.
The proposed Technical Specification amendment to Sections 3.9.6
and 4.9.6.1 would provide a revised refueling machine hoist overload
cut off limit that is appropriate for the increased weight of the
fuel assemblies. The increased weight of fuel assemblies results
from design and fabrication improvements such as denser fuel
pellets, laser welded GUARDIANTM grids, and laser welded spacer
grids. The overload cut off limit was incorporated on the refueling
machine hoist to protect the core internals and pressure vessel from
possible damage in the event the fuel assembly becomes mechanically
bound as it is withdrawn from the reactor vessel. The proposed
overload cut off limit was determined as follows:
Overload Cut Off limit=(Hoist Wet Weight)+(Grapple Wet Weight)+(Max
Wet Fuel Weight)+90 lbs.
Where:
(a) Hoist and Grapple Wet Weight=176 lbs.
(b) Maximum Wet Fuel Weight=1334 lbs.
The basis for the 90 pounds had two considerations: (1) to be
large enough to account for friction loads during fuel assembly
withdrawal; and, (2) to be small enough to ensure that while lifting
a minimum weight fuel assembly, the loads imposed on a mechanically
bound fuel assembly are below the design limit specified by the fuel
manufacturer. The maximum value for the existing overload cut off
limit was specified by the fuel manufacturer to be 1602 pounds.
The overload cut off limit is not a parameter used in the
analysis of a Fuel Handling Accident. The conclusion regarding the
radiological consequences of the Fuel Handling Accident remain
valid, and there is no decrease in the margin of safety.
Therefore, it can be concluded that the proposed change will
maintain the integrity of the fuel assemblies and reactor vessel
internals and does not involve a significant reduction in the 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 considered in making any final
determination.
Normally, the Commission will not issue the amendments 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 amendments before the expiration
of the 30-day notice period, provided that its final determination is
that the amendments involve 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, 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 6, 1995, the licensee may file a request for a hearing
will respect to issuance of the amendments 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 [[Page 2162]] 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 Phoenix Public Library, 12 East McDowell
Road, Phoenix, Arizona 85004. 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 amendments 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
amendments and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendments.
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, 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 Theodore R. Quay: 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, and to Nancy C. Loftin,
Esq., Corporate Secretary and Counsel, Arizona Public Service Company,
P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999,
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 October 31, 1994, as supplemented by
letter dated December 28, 1994, 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 Phoenix Public Library, 12 East McDowell
Road, Phoenix, Arizona 85004.
Dated at Rockville, Maryland, this 3rd day of January 1995.
For the Nuclear Regulatory Commission.
Linh N. Tran,
Project Manager, Project Directorate IV-2, Division of Reactor Projects
III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 94-319 Filed 1-5-95; 8:45 am]
BILLING CODE 7590-01-M
