FR Doc 02-30921

[Federal Register: December 10, 2002 (Volume 67, Number 237)]
[Notices]
[Page 75867-75889]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr10de02-59]

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of
1954, as amended (the Act), to require the Commission to publish notice
of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, November 15, 2002, through November 29,
2002. The last biweekly notice was published on November 26, 2002 (67
FR 70762).

Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing

The Commission has made a proposed determination that the following
amendment requests involve 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. The basis
for this proposed determination for each amendment request is shown
below.
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
before action is taken. 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 and
Directives Branch, Division of Administrative 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 Commission's Public
Document Room (PDR), located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By January 9, 2003, 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,\1\ which is
available at the Commission's PDR, located at One White Flint North,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/.
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 a hearing or an
appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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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 a party to the proceeding; (2) the

[[Page 75868]]

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: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to (301) 415-1101 or by e-mail to
hearingdocket@nrc.gov. A copy of the request for hearing and petition
for leave to intervene should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and because of continuing disruptions in delivery of mail to United
States Government offices, it is requested that copies be transmitted
either by means of facsimile transmission to (301) 415-3725 or by e-
mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
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 which is available for public inspection at
the Commission's PDR, located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. Publicly available records
will be accessible from the Agencywide Documents Access and Management
System's (ADAMS) Public Electronic Reading Room on the Internet at the
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not
have access to ADAMS or if there are problems in accessing the
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona

Date of amendment requests: November 7, 2002.
Description of amendment requests: The amendments would revise
Technical Specification (TS) 3.2.4, ``Departure From Nucleate Boiling
Ratio (DNBR),'' TS 3.3.1, ``Reactor Protective System (RPS)
Instrumentation--Operating,'' and TS 3.3.3, ``Control Element Assembly
Calculators (CEACs).'' The proposed changes are to Limiting Conditions
for Operation (LCOs), LCO Actions, and LCO Surveillance Requirements.
The amendments support the replacement of the Core Protection
Calculator System (CPCS). The replacement CPCS will perform
functionally identical safety-related algorithms as the existing CPCS,
although on a newer platform, and the CPCS design function will remain
unchanged. Because the replacement CPCS for each unit will be installed
in refueling outages for the three units over at least a year, starting
with the Unit 2 fall 2003 outage, the licensee has proposed to have the
TSs contain both the current requirements and the new requirements with
the phrases ``(Before CPC Upgrade)'' and ``(After CPC Upgrade)'' on the
TSs to show which requirements apply to which case.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Core Protection Calculator System (CPCS) is being replaced
due primarily to parts obsolescence. The replacement CPCS will
perform functionally identical safety-related algorithms as the
existing CPCS, but on a newer platform. The CPCS design function
will remain unchanged.
The physical location of the replacement CPCS will be the same
as the existing CPCS in the auxiliary protective cabinets.
Installation will occur during refueling outages when the system is
not required for service. [The] majority of the testing will be
performed prior to installation.

[[Page 75869]]

The CPCS is not an initiator of any analyzed accident, but is
used for mitigation of a large number of anticipated operational
occurrences and a small number of accidents. Since the CPCS is not
an accident initiator, and the replacement CPCS is functionally
unchanged, the CPC replacement will not increase the probability of
an accident.
The functionality of the existing CPCS safety related algorithms
are replicated in the System Requirements Specification for the
Common Q [Common Qualified] Core Protection Calculator System. The
basic Common Q CPCS design concept was approved by NRC Safety
Evaluation (SE), Acceptance For Referencing Of Topical Report CENPD-
396-P, Rev. 01, ``Common Qualified Platform'' and Appendices 1, 2, 3
and 4, Rev. 01, dated August 11, 2000 (Ref. 2 [listed in the
enclosure to the amendment request]), and there have been no
significant functional changes to the design as presented. The
requirements for response time and accuracy that are assumed in the
Palo Verde Nuclear Generating Station (PVNGS) Updated Final Safety
Analysis Report (UFSAR) accident analysis will continue to be met.
Therefore, since the new [replacement] CPCS will be capable of
performing the same safety-related functions within the same
response time and accuracy as the existing CPCS, the proposed change
does not involve a significant increase in the consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The CPCS provides a monitoring and detection function and is not
an initiator for any accident. The CPCS provides Reactor Protection
System (RPS) trips on Low Departure from Nucleate Boiling Ratio
(DNBR) and High Local Power Density (LPD) in response to
calculations involving several input variables. It also provides a
Control Element Assembly Withdrawal Prohibit (CWP) signal to the
Plant Protection System (PPS), and provides indication and
annunciation. The CPCS performs no other plant functions, and is not
used to initiate any ESF [(Engineered Safety Feature)] functions.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The [new] CPCS is a replacement for the existing CPCS. It will
retain the same safety-related functionality as the existing CPCS.
The equipment will be qualified in accordance with requirements
described in the Palo Verde UFSAR.
The replacement CPCS will perform functionally identical safety-
related algorithms as the existing CPCS, will trip in response to
the same inputs with equivalent accuracy, and will meet the same
four channel separation requirements. The only significant area of
difference involves the platform. The Common Q platform uses a
consistent set of qualified building blocks (Advant Controllers,
Flat Panel Displays, Power Supplies, and Communication Systems) that
can be used for any safety system application. For Palo Verde
purposes, the only application of this platform at this time will be
for use as a CPCS. The new platform will include improved human
factors and fault tolerance within each CPCS channel.
In summary, the replacement CPCS performs the same function as
the existing CPCS, meets the qualification requirements of the
existing CPCS, and meets the accuracy standards of the existing
CPCS. Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, APS [(the licensee)] concludes that the
proposed amendment(s) present no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and, accordingly,
a finding of ``no significant hazards consideration'' is justified.

The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, PO Box 53999, Mail Station
9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.

Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam
Electric Plant, Unit 2, Brunswick County, North Carolina

Date of amendment request: November 7, 2002.
Description of amendment request: The proposed amendment would
revise the Minimum Critical Power Ratio (MCPR) Safety Limit contained
in Technical Specification 2.1.1.2 from 1.09 to 1.11 for two
recirculation loop operation and from 1.10 to 1.13 for single
recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: 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:

CP&L [Carolina Power and Light Company] has evaluated whether or
not a significant hazards consideration is involved with the
proposed amendment by focusing on the three standards set forth in
10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The MCPR Safety Limit values are calculated to ensure that
greater than 99.9 percent of the fuel rods in the core avoid
transition boiling during any plant operation if the safety limit is
not violated. The derivation of the MCPR Safety Limit values
specified in the Technical Specifications, and their use to
determine cycle-specific thermal limits, has been performed using
the methodology discussed in ``General Electric Standard Application
for Reactor Fuel,'' NEDE-24011-P-A-14 (i.e., GESTAR-II), and U.S.
Supplement, NEDE-24011-P-A-14-US, June 2000, which incorporates
Amendment 25. Amendment 25 was approved by the NRC in a March 11,
1999, safety evaluation report. Operational MCPR limits are applied
that ensure the MCPR Safety Limit is not exceeded during all modes
of operation and anticipated operational occurrences.
The revised MCPR Safety Limit values do not affect the
operability of any plant systems nor do these revised values
compromise any fuel performance limits; therefore, the probability
of fuel damage will not be increased as a result of this change.
The MCPR Safety Limit values do not impact the source term or
pathways assumed in accidents previously evaluated, and there are no
adverse effects on the factors contributing to offsite or onsite
radiological doses. In addition, the revised MCPR Safety Limit
values do not affect the performance of any equipment used to
mitigate the consequences of a previously evaluated accident and do
not affect setpoints that initiate protective or mitigative actions.
Therefore, the proposed Technical Specification change does not
involve a significant increase in the probability or consequences of
a previously evaluated accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. The proposed revision of the MCPR
Safety Limit values does not involve any facility modifications, and
plant equipment will not be operated in a different manner. No new
initiating events or transients will result from the revised MCPR
Safety Limit values. As a result, no new failure modes are being
introduced. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components; through the parameters
within which the plant is operated; through the establishment of
setpoints for actuation of equipment relied upon to respond to an
event; and through margins contained within the safety analyses. The
revised MCPR Safety

[[Page 75870]]

Limit values will not adversely impact the performance of plant
structures, systems, components, and setpoints relied upon to
respond to mitigate an accident or transient. The MCPR Safety Limit
values are calculated to ensure that greater than 99.9 percent of
the fuel rods in the core avoid transition boiling during any plant
operation if the safety limit is not violated, thereby ensuring that
fuel cladding integrity is maintained. The revised MCPR Safety Limit
values have been calculated using NRC approved methods and
procedures and preserve the existing margin to transition boiling.
Based on the assurance that the fuel design criteria are being met,
the revised MCPR Safety Limit values do not involve a reduction in a
margin of safety.
Based on the above, CP&L has concluded that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.

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.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: November 14, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specification Surveillance Requirement (SR)
3.3.1.3 to add a correlation slope to the formula for imbalance error.
The SR is also being changed to require an adjustment of the power
range channel output if the absolute value of the imbalance error is
=2 percent rated thermal power.
Basis for proposed no significant hazards consideration
determination: 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:

Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the
determination that this amendment request involves a No Significant
Hazards Consideration by applying the standards established by the
NRC regulations in 10 CFR 50.92. This ensures 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.
No. This change will add a correlation slope (CS) to Imbalance
Error that is derived from the Power Imbalance Detector Correlation
(PIDC) test performed during the cycle startup testing. The formula
currently exists in the technical specification. The CS will add
nuclear conservatism to the error calculation.
Since the calculation already exists and the CS adds more
conservatism, this proposed change does not involve an increase in
the probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated.
No. As stated above, the proposed revision adds a conservative
CS to the existing error calculation. This change is bounded by all
of the existing accidents and does not create the possibility of a
new or different kind of accident from any kind of accident
previously evaluated.
(3) Involve a significant reduction in a margin of safety.
No. The proposed change does not adversely affect any plant
safety limits, set points, or design parameters. The change also
does not adversely affect the fuel, fuel cladding, Reactor Coolant
System, or containment integrity. Therefore, the proposed change
does not involve a significant in a margin of safety.
Duke has concluded, based on the above, that there are no
significant hazards considerations involved in this amendment
request.

Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington.

Date of amendment request: October 22, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to change TS Section 5.0,
``Administrative Controls,'' to adopt Technical Specification Task
Force (TSTF) -258, Revision 4. The proposed changes would: (1) Revise
TS Section 5.2.2, ``Unit Staff,'' to delete the details of the staffing
requirements and delete the requirements for the Shift Technical
Advisor (STA) as a separate position while retaining the function, (2)
revise TS Section 5.5.4, ``Radioactive Effluent Controls Program,'' to
be consistent with the intent of 10 CFR Part 20, (3) revise TS Section
5.6.4, ``Monthly Operating Reports,'' to delete periodic reporting
requirements for main steam safety/relief valve challenges to be
consistent with Generic Letter 97-02, ``Revised Contents of the Monthly
Operating Report,'' and (4) revise TS Section 5.7, ``High Radiation
Area,'' in accordance with 10 CFR 20.1601(c). A new TS Section 5.3.2
would be added to incorporate regulatory definitions for the senior
reactor operator (SRO) and reactor operator (RO) positions.
Basis for proposed no significant hazards consideration
determination: 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change is an administrative clarification of
existing TS requirements which clarifies and modifies administrative
controls in the areas of operator staffing requirements, working
hour limits, STA position, Radioactive Effluent Controls Program,
periodic reporting requirements for relief valve openings, and
radiological control requirements. These changes do not impact the
operation, physical configuration, or function of plant equipment or
systems. These TS revisions do not affect analysis inputs or
mitigation for analyzed accidents and transients. Therefore, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design. The
proposed change does not introduce any new modes of plant operation
or make any changes to system setpoints. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative in nature and does not
involve physical changes to plant structures, systems, or components
(SSCs), or the manner in which these SSCs are operated, maintained,
modified, tested, or inspected. The proposed change does not involve
a change to any safety limit, limiting safety system setting,
limiting condition for operation, or design parameters for any SSC.
The proposed change does not impact any safety analysis assumptions
and does not

[[Page 75871]]

involve a change in initial conditions, system response times, or
other parameters affecting any accident analysis.
For these reasons, the proposed amendment does not involve a
significant reduction in the margin of safety.

Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington

Date of amendment request: October 22, 2002.
Description of amendment request: The proposed amendment deletes a
reference to Section 2.E in Section 2.F of Facility Operating License
No. NPF-21. Section 2.E requires the licensee to fully implement and
maintain in effect all provisions of the Commission-approved physical
security, guard training and qualification, and safeguards contingency
plans. Section 2.E is redundant because the reporting requirements and
criteria for the Physical Security Programs are specified in 10 CFR
73.71 and Appendix G of 10 CFR part 73.
Basis for proposed no significant hazards consideration
determination: 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed Operating License amendment request is
administrative in nature and merely deletes a duplicative and
unnecessary reporting requirement. The proposed amendment deletes a
reference to Operating License Section 2.E in Operating License
Section 2.F. Operating License Section 2.F presently requires the
Columbia Generating Station to report any violations of the
requirements contained in Section 2.C (with the exception of 2.C(2))
and 2.E of the License. Operating License Section 2.E requires
Columbia Generating Station to fully implement and maintain in
effect all provisions of the Commission-approved physical security,
guard training and qualification, and safeguards contingency plans.
The requirement to report a violation of Section 2.E is redundant
and unnecessary because the reporting requirements and criteria for
the physical security program are specified in [10 CFR 73.71 and 10
CFR 73] Appendix G. This change to the Operating License has no
impact on the manner in which the Columbia Generating Station is
operated. No actual plant equipment or accident analyses will be
affected by the proposed change. There will be no increase in
radiological dose to plant workers or the public. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed Operating License amendment request is
administrative in nature and merely deletes a duplicative and
unnecessary reporting requirement. The proposed amendment deletes a
reference to Operating License Section 2.E in Operating License
Section 2.F. Operating License Section 2.F presently requires the
Columbia Generating Station to report any violations of the
requirements contained in Section 2.C (with the exception of 2.C(2))
and 2.E of the License. Operating License Section 2.E requires
Columbia Generating Station to fully implement and maintain in
effect all provisions of the Commission-approved physical security,
guard training and qualification, and safeguards contingency plans.
The requirement to report a violation of Section 2.E is redundant
and unnecessary because the reporting requirements and criteria for
the Physical Security Program are specified in 10 CFR 73.71 and 10
CFR 73 Appendix G. This request is administrative in nature. This
change to the Operating License has no impact on the manner in which
the Columbia Generating Station is operated. No actual plant
equipment or accident analyses will be affected by the proposed
change. No failure modes not bounded by previously evaluated
accidents will be created. Therefore, the proposed amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no direct effect on any safety analyses assumptions, and no
adverse effect on the performance of any system, structure, or
component relied upon for accident mitigation. The proposed
amendment deletes a reference to Operating License Section 2.E in
Operating License Section 2.F. Deletion of the reference to Section
2.E eliminates a redundant and unnecessary reporting requirement,
because the reporting requirements and criteria for the physical
security program are specified in 10 CFR 73.71 and 10 CFR 73
Appendix G. Additionally, there would be no effect on baseline core
damage probability. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.

Date of amendment request: September 18, 2002.
Description of amendment request: The proposed change will revise
the Technical Specifications (TS) Limiting Conditions for Operation and
Administrative sections to correct or clarify certain requirements and
information.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are primarily to correct word omissions,
typographical errors, reflect current terminology, and make the TS
consistent with other NRC [U.S. Nuclear Regulatory Commission]
approved documents. These changes are all of an administrative
nature and have no effect on any plant equipment or structures.
Therefore, these changes do not increase the probability or
consequences of an accident previously evaluated.
The proposed amendment also revises the allowed drywell-to-
primary containment differential pressure limit. This limit is
intended to ensure that containment conditions are consistent with
safety analyses. The proposed smaller negative pressure ensures that
the design assumptions for the containment will be met if and when a
postulated loss of coolant [accident] (LOCA) should occur. Moving
the limit in a conservative direction will not increase the
probability or consequences of previously evaluated accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant. No new or different equipment or modes of operation are being
introduced by this proposed change. Thus, the changes do not create
the

[[Page 75872]]

possibility of a new or different kind of accident from any accident
previously evaluated.
The change to the allowed drywell-to-primary containment
differential pressure limit does not adversely impact the ability of
the containment to perform its intended function. The establishment
of a more conservative limit for this parameter ensures that the
plant stays within current safety analysis and therefore, can not
create the possibility of a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed changes are primarily administrative in nature
and can not affect any safety barriers. The proposed change to the
allowed drywell-to-primary containment differential pressure limit
establishes a more conservative limit for a key parameter for the
containment than is currently specified in the TS. The revised
differential pressure limit is consistent with current assumptions
of the accident analysis. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.

Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi; Entergy Gulf States, Inc., and Entergy Operations,
Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana
Parish, Louisiana; and Entergy Operations, Inc., Docket No. 50-382,
Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 6, 2002.
Description of amendment request: The proposed change will delete
the content of the Appendix B, Environmental Protection Plan (Non-
Radiological) (EPP), and the appropriate sections of the Facility
Operating License (FOL) referring to the EPP will be modified to delete
reference to the EPP.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The EPPs are concerned with monitoring the effect that plant
operations have on the environment for the purpose of protecting the
environment and has no affect on any accident postulated in the
Updated Final Safety Analysis Report (UFSAR). Accident probabilities
or consequences are not affected in any way by the environmental
monitoring and reporting required by the EPPs. The deletion of
Appendix B of the FOL will not impact the design or operation of any
plant system or component. The NRC [Nuclear Regulatory Commission]
relies on other Federal, State, and local agencies for environmental
protection regulation. No environmental protection requirements
established by these other agencies are being reduced by this
license amendment. The programs and reporting requirements of the
EPPs do not affect the initiation or mitigation of any accidents
previously analyzed.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This license amendment is administrative in nature.
Environmental monitoring and reporting has no affect on accident
initiation. The deletion of the EPPs will not produce any changes to
the design or operation of the plant. There will be no effect on the
types and amounts of any effluent that will be released.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change is administrative in nature. The change in annual
reporting requirements has no impact on margin of safety.
Environmental Evaluations will still be performed, where necessary,
on changes to plant design or operations to assess the effect on
environmental protection.
Therefore, the proposed change does 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.
Attorneys for licensee: (Grand Gulf Nuclear Station, Unit 1, and
Waterford Steam Electric Station, Unit 3) Nicholas S. Reynolds, Esq.,
Winston & Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-
3502; and (River Bend Station, Unit 1) Mark Wetterhahn, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts

Date of amendment request: August 16, 2002.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3/4.10.A, ``Refueling Interlocks''
to provide an alternative required action if the refueling interlocks
became inoperable during fuel movements in the reactor vessel. The
proposed amendment would also modify TS 3/4.10.D, ``Multiple Control
Rod Removal.'' The proposed changes would allow fuel movements in the
reactor vessel should the refueling equipment interlocks become
inoperable.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The refueling interlocks function to prevent prompt reactivity
excursions during refueling. Criticality and, therefore, subsequent
prompt reactivity excursions are prevented during the insertion and
during control rod movement provided the other control rods in core
cells containing one or more fuel assemblies are fully inserted. The
refueling interlocks accomplish this by preventing loading of fuel
into the core with any control rod withdrawn, by preventing
withdrawal of a rod from the core during fuel loading, or preventing
multiple control rod withdrawal. The proposed requirements ensure
that these functions can be performed when required. Therefore, the
probability of an accident previously evaluated is not significantly
increased.
The refueling interlocks addressed by these specifications do
not mitigate the consequences of any accident. Therefore,
consequences of an accident previously evaluated are not
significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of

[[Page 75873]]

accident [from] any accident previously evaluated?
Response: No.
The proposed change does not involve a change to the plant
design. The refueling interlocks function to prevent prompt
reactivity excursions during refueling. The proposed requirements
ensure that these functions can be performed when required. As a
result, the proposed changes do not affect any of the parameters or
conditions that could contribute to the initiation of any new or
different kind of accident. Therefore, this proposed [change] does
not create the possibility of a new or different kind of accident
[from] any accident previously evaluated.
3. Does the change involve a significant reduction in [the]
margin of safety?
Response: No.
The refueling interlocks function to prevent prompt reactivity
excursions during refueling. Criticality and, therefore, subsequent
prompt reactivity excursions are prevented during the insertion of
fuel, provided all control rods are fully inserted during the fuel
insertion and during control rod movement provided the other control
rods in core cells containing one or more fuel assemblies are fully
inserted. The refueling interlocks accomplish this by preventing
loading of fuel into the core with any control rod withdrawn, by
preventing withdrawal of a rod from the core during fuel loading, or
preventing multiple control rod withdrawal. The proposed
requirements ensure that these functions can be performed when
required. Therefore, the proposed changes do 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: James W. Andersen, Acting.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts

Date of amendment request: August 16, 2002.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 3.10.D.1.d from TS 3/4.10.D,
``Multiple Control Rod Removal,'' and the associated Surveillance
Requirement 4.10.D.1.d. The proposed changes involving the deletion of
this requirement would reduce the number of fuel movements or valve
manipulations, thereby, increasing safety and reducing worker dose. In
addition, the proposed amendment would also make an editorial change to
correct a reference to TS 3.3.B.3 instead of TS 3.3.B.4 in TS 3/
4.10.D.1.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Following the deletion of the requirement that all control rods
in a 3x3 array centered on each of the control rods being removed be
fully inserted and electronically or hydraulically disarmed, or have
the surrounding four fuel assemblies removed from the core cell,
sufficient barriers will be in place to prevent the possibility of
an unacceptable reactivity excursion.
As a backup to licensee procedures and controls to prevent an
unacceptable reactivity excursion, the Technical Specifications (TS)
will continue to have two layers of controls to ensure that an
unacceptable reactivity excursion cannot occur. The first layer of
control is on the local reactivity effects of withdrawing the
control rod while the second is on any potential core wide effects.
The local reactivity effects of removing the control rod are
addressed by the requirement that the four fuel assemblies be
removed from the core cell surrounding each control rod or control
rod drive mechanism to be removed from the core and/or the reactor
vessel. The requirement that the fuel assemblies in the cell
controlled by the control rod be removed from the reactor core
ensures withdrawal of another control rod cannot result in an
unacceptable reactivity excursion.
Any potential core wide effects of removing the control rod will
also continue to be controlled by the TS. The TS will continue to
require control rods that are not withdrawn in accordance with 3/
4.10.D remain fully inserted, the core remain sub-critical with a
margin with the highest worth control rod withdrawn, and no more
than one control rod can be inadvertently withdrawn. These
requirements together ensure an operator error that resulted in the
withdrawing of a control rod from a fueled cell would not result in
an unacceptable reactivity excursion and the operator cannot
withdraw a second control rod in error. Therefore, these
requirements ensure that adequate [Shutdown Margin] SDM will be
maintained, thereby, preventing unacceptable reactivity excursions
during refueling.
In addition to these two barriers preventing an unacceptable
reactivity excursion, the TS will continue to require that the
source range monitors be operable. This requirement ensures that
neutron monitoring information is available to the operators
providing them with the information necessary to identify an
unacceptable reactivity excursion is occurring and take action to
terminate the event.
The control remaining provide sufficient assurance an
unacceptable reactivity excursion will not occur during these
activities. Therefore, the probability of an accident previously
evaluated is not significantly increased.
The control being deleted did not mitigate the consequences of
any accident. Therefore, consequences of an accident previously
evaluated are not significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident [from] any accident previously evaluated?
Response: No.
The proposed change does not involve a change to the plant
design or a new mode of equipment operation. As a result, the
proposed change does not affect parameters or conditions that could
contribute to the initiation of any new or different kind of
accident. Therefore, this proposed [change] does not create the
possibility of a new or different kind of accident [from] any
accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
Following the deletion of the requirement that all control rods
in a 3x3 array centered on each of the control rods being removed be
fully inserted and electrically or hydraulically disarmed, or have
the surrounding four fuel assemblies removed from the core cell,
sufficient barriers will be in place to prevent the possibility of
an unacceptable reactivity excursion.
The TS will continue to have controls as a backup to licensee
procedures and controls to prevent an unacceptable reactivity
excursion. The requirement that the fuel assemblies in the cell
controlled by the control rod be removed from the reactor core
ensures withdrawal of another control rod cannot result in an
unacceptable reactivity excursion. Also the TS will ensure that an
operator error which results in the withdrawing of a control rod
from a fueled cell will not result in an unacceptable reactivity
excursion and that the operator cannot withdraw a second control rod
in error.
In addition to these two barriers preventing an unacceptable
reactivity excursion, the TS will continue to require that the
source range monitors be operable. This requirement ensures that
neutron monitoring information is available to the operators
providing them with the information necessary to identify that an
unacceptable reactivity excursion is occurring and take action to
terminate the event.
The controls remaining provide sufficient assurance an
unacceptable reactivity excursion will not occur during these
activities. Therefore, the proposed changes do not involve a
significant reduction in [a] margin of safety.

[[Page 75874]]

Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: James W. Andersen, Acting.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 24, 2002.
Description of amendment request: The amendment would revise
Technical Specifications (TSs) relating to positive reactivity
additions while in shutdown modes by clarifying TSs involving the
positive reactivity additions. The proposed changes are based on
Technical Specification Task Force (TSTF)-286, Revision 2, and allow
for small, controlled, safe insertions of positive reactivity while in
shutdown modes. In addition, two administrative-type changes are
proposed.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) changes revise actions
that either require suspension of operations involving positive
reactivity additions or preclude reduction in boron concentration
less than the reactor coolant system (RCS). Reactivity excursions
are analyzed events. The proposed changes limit positive reactivity
additions into the RCS such that the required shutdown margin (SDM)
or refueling boron concentration continue to be met. Reactivity
changes performed during shutdown modes are currently governed by
strict administrative controls. Although the proposed changes will
allow procedural flexibility with regards to RCS temperature and
boron concentration, these operations will still be under
administrative control. The changes proposed by these amendments are
within the scope and assumptions of the existing analyses.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS revisions relate to positive reactivity
additions while in shutdown modes of operation. Reactivity
excursions are analyzed events. The operational flexibility allowed
in these proposed license amendments will be performed under strict
administrative controls in order to limit the potential for
excessive positive reactivity addition. Although the existing
procedural controls will need modification, no new or different
operational failure modes will be introduced by these changes.
Additionally, implementation of these proposed changes does not
require any physical plant modifications, so no new or different
hardware-related failure modes are introduced. The changes proposed
by these amendments are within the scope and assumptions of the
existing analyses.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previosly evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes conform closely to the industry and NRC
approved TSTF-286, Rev[ision] 2, and relate to small, controlled,
safe insertions of positive reactivity additions while in shutdown
modes. These changes revise actions that either require suspension
of operations involving positive reactivity additions, or prohibit
RCS boron concentration reduction. The proposed changes provide
operational flexibility while controlling positive reactivity
additions. The proposed changes provide for continued safe reactor
operations and preserve the required SDM or refueling boron
concentration.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois

Date of amendment request: October 16, 2002.
Description of amendment request: The proposed amendment would
revise the Completion Time for Required Action A.1 of TS 3.8.7,
``Inverters--Operating,'' from the current 24 hours for one instrument
bus inverter inoperable to 14 days. The change is being proposed to
support on-line maintenance of the instrument bus inverters and will
have a negligible impact on plant safety. The current Completion Time
for restoration of an inoperable instrument bus inverter is
insufficient to support the required maintenance and post-maintenance
testing windows.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed action allows continued unit operation, for up to
14 days, with an inoperable instrument bus inverter. An inoperable
instrument bus inverter is not considered as an initiator of any
analyzed event. Extending the Completion Time for an inoperable
instrument bus inverter would not have a significant impact on the
frequency of occurrence for any accident previously evaluated. The
proposed change will not result in changes to the plant activities
associated with instrument bus inverter maintenance, but rather will
allow increased flexibility in the scheduling and performance of
preventive maintenance. Therefore, this change will not
significantly increase the probability of occurrence of any event
previously analyzed in the current Byron/Braidwood Stations' Updated
Final Safety Analysis Report (UFSAR) safety analyses.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed in the analysis, the availability and
successful functioning of equipment assumed to operate in response
to the analyzed event, and the setpoints at which these actions are
initiated. With an instrument bus inverter inoperable, the affected
instrument bus is capable of being fed from its dedicated safety-
related constant voltage transformer (CVT), which is powered from a
480 VAC Engineered Safety Feature (ESF) bus. In the event of a Loss
of Offsite Power (LOOP), the affected instrument bus will experience
a momentary loss of power until the associated diesel generator (DG)
re-energizes the 480 VAC ESF bus. A LOOP with an inoperable
instrument bus inverter (i.e., instrument bus being powered by its
CVT) will result in a loss of power to the associated instrument bus
until the associated DG re-energizes the 480 VAC ESF bus. All
instruments supplied by the instrument bus would be restored with no
adverse impact to the units because no other instrument channels in
the opposite train would be expected to be inoperable or in a
tripped condition during this time, with the exception of routine
surveillances. In the event the DG failed (i.e., failed to re-
energize the 480 VAC ESF bus), power could still be established to
the 4 kV ESF bus by powering the 480 VAC ESF bus from the opposite
unit 4 kV ESF bus cross-tie breaker. In the event of a failure to
re-energize the 480 VAC ESF bus or of a CVT failure, the most
significant impact on the unit is the failure of one train

[[Page 75875]]

of ESF equipment to actuate. In this condition, the redundant train
of ESF equipment will automatically actuate to mitigate the
accident, and the affected unit would remain within the bounds of
the accident analyses. Therefore, the request for extending the
Completion Time will not significantly increase the consequences of
an accident previously evaluated in the Byron/Braidwood Stations'
UFSAR.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed action does not involve physical alteration of the
station. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
is no change being made to the parameters within which the units are
operated. There are no setpoints at which protective or mitigative
actions are initiated that are affected by this proposed action. The
use of the CVT as an alternate power source for the instrument bus
is consistent with the Byron and Braidwood Stations' plant designs.
This proposed action will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No alteration in the procedures, which ensure
the unit remains within analyzed limits, is proposed, and no change
is being made to procedures relied upon to respond to an off-normal
event. As such, no new failure modes are being introduced. The
proposed action does not alter assumptions made in the safety
analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
or actions. There is no change in the design of the affected
systems, no alteration of the setpoints at which alarms or actions
are initiated, and no change in plant configuration from original
design. With one of the required instrument buses being powered from
the CVT, there is no significant reduction in the margin of safety.
Testing of the DGs and associated electrical distribution equipment
provides confidence that the DGs will start and provide power to the
associated equipment in the unlikely event of a LOOP during the
extended 14-day Completion Time.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above evaluation, we have concluded that the
proposed change does not involve a significant hazards
consideration.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: October 10, 2002.
Description of amendment request: The proposed amendments would
change technical specifications to increase the number of safety valves
required to be operable from eight to nine and add surveillance
requirements for the ninth safety valve.
Basis for proposed no significant hazards consideration
determination: 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. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed Technical Specifications (TS) changes require an
additional safety valve to be operable. The proposed change also
adds the requirement to verify the lift setpoint of this additional
safety valve. TS requirements that govern operability or routine
testing of plant components are not assumed to be initiators of any
analyzed event because these components are intended to prevent,
detect, or mitigate accidents. Therefore, these changes will not
involve an increase in the probability of an accident previously
evaluated.
The proposed changes ensure that the reactor pressure vessel
(RPV) steam dome pressure response is maintained within established
limits in order to maintain the analyzed response of the RPV steam
dome pressure below the safety limit for this parameter during the
most severe pressurization transient. This ensures that the reactor
coolant system integrity will be maintained during this transient.
Thus, the proposed change does not involve an increase in the
consequences of an accident previously evaluated.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not affect the manner in which plant
systems will be operated under normal and abnormal operating
conditions. Therefore, these changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes ensure that the RPV steam dome pressure
response is maintained within established limits in order to
maintain the analyzed response of the RPV steam dome pressure below
the safety limit for this parameter during the most severe
pressurization transient. Ensuring the safety limit is met for this
transient ensures that RCS integrity will be maintained. Therefore,
the proposed changes do not result in a reduction in the margin of
safety.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: October 28, 2002.
Description of amendment request: The proposed amendments would
authorize changes to the Updated Final Safety Analysis Report (UFSAR)
to address the use of cast iron components in the containment cooling
service water and emergency diesel generator cooling water systems.
These changes were submitted to the Nuclear Regulatory Commission (NRC)
for review and approval in accordance with 10 CFR 50.59(c)(2).
Basis for proposed no significant hazards consideration
determination: 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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes allow for the use of cast iron materials in
the Containment Cooling Service Water (CCSW) and Diesel Generator
Cooling Water (DGCW) Systems at Dresden Nuclear Power Station
(DNPS). The use of cast iron materials in these systems would be
subject to acceptance criteria proposed for incorporation into the
DNPS Updated Final Safety Analysis Report (UFSAR).
A failure in the CCSW or DGCW systems is not an initiator of any
analyzed accident described in the UFSAR. Therefore, these proposed
changes would not involve an increase in the probability of an
accident previously evaluated. Additionally, these

[[Page 75876]]

proposed changes would not increase the consequences of an accident
previously evaluated because the proposed changes would not
adversely impact structures, systems, or components. The proposed
UFSAR acceptance criteria establish requirements for cast iron use
that ensure the CCSW and DGCW systems would be capable of performing
their intended safety-related functions of supplying cooling water
to essential plant equipment, even during a design basis earthquake.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes allow for the use of cast iron materials in
the CCSW and DGCW systems at DNPS by adding acceptance criteria to
the UFSAR for such material. No other changes in requirements are
being proposed. The added acceptance criteria establish requirements
for cast iron that ensure the CCSW and DGCW systems would be capable
of performing their safety-related functions of supplying cooling
water to essential plant equipment, even during a design basis
earthquake. No new failure modes are introduced by the proposed
change. No new sources of energy are added. There is no change being
made to the parameters within which DNPS is operated, nor do the
proposed changes physically alter the plant. The proposed changes do
not adversely impact the manner in which the CCSW or DGCW systems
will operate under normal and abnormal operating conditions. The
plant response to any single failure is not changed. The proposed
changes will not alter the function demands on credited equipment.
No alteration in the procedures, which ensure DNPS remains within
analyzed limits, is proposed, and no change is being made to
procedures relied upon to respond to an off-normal event. Therefore,
these proposed changes provide an equivalent level of safety and
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The CCSW and DGCW systems are addressed in Technical
Specifications (TS) Sections 3.7.1 and 3.7.2. However, the Bases of
these TS sections do not discuss the codes to which the systems are
designed. Margins of safety are established in the design of
components, the configuration of components to meet certain
performance parameters, and in the establishment of setpoints to
initiate alarms and actions. The proposed cast iron acceptance
criteria will ensure that any implied margin of safety is maintained
regarding the ability of the CCSW and DGCW systems to perform their
safety functions during all design basis conditions. Therefore, it
is concluded that the proposed changes do not result in a
significant reduction in the margin of safety.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 24, 2002.
Description of amendment request: The proposed amendments would
revise Technical Specification 5.5.13, ``Primary Containment Leakage
Rate Testing Program,'' to reflect a one-time deferral of the primary
containment Type A test to no later than June 13, 2009, for Unit 1 and
no later than December 7, 2008, for Unit 2.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes will revise LaSalle County Station,
Units 1 and 2, Technical Specification (TS) 5.5.13, ``Primary
Containment Leakage Rate Testing Program'' to reflect a one-time
deferral of the primary containment Type A test to no later than
June 13, 2009, for Unit 1 and no later than December 7, 2008, for
Unit 2. The current Type A test interval of ten years, based on past
performance, would be extended on a one-time basis to 15 years from
the last Type A test.
The function of the primary containment is to isolate and
contain fission products released from the reactor Primary Coolant
System (PCS) following a design basis Loss-of-Coolant Accident
(LOCA) and to confine the postulated release of radioactive material
to within limits. The test interval associated Type A testing is not
a precursor of any accident previously evaluated. Type A testing
does provide assurance that the LaSalle County Station primary
containments will not exceed allowable leakage rate values specified
in the Technical Specifications and will continue to perform their
design function following an accident. The risk assessment of the
proposed changes has concluded that there is an insignificant
increase in total population dose rate and an insignificant increase
in the conditional containment failure probability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed changes for a one-time extension of the Type A
tests for LaSalle County Station, Units 1 and 2 will not affect the
control parameters governing unit operation or the response of plant
equipment to transient and accident conditions. The proposed changes
do not introduce any new equipment, modes of system operation or
failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
No. LaSalle County Station, Units 1 and 2, are General Electric
BWR/5 plants with Mark II primary containments. The Mark II primary
containment consists of two compartments, the drywell and the
suppression chamber. The drywell has the shape of a truncated cone,
and is located above the cylindrically shaped suppression chamber.
The drywell floor separates the drywell and the suppression chamber.
The primary containment is penetrated by access, piping and
electrical penetrations.
The integrity of the primary containment penetrations and
isolation valves is verified through Type B and Type C local leak
rate tests (LLRT) and the overall leak tight integrity of the
primary containment is verified by a Type A integrated leak rate
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor
Containment Leakage Testing for Water-Cooled Power Reactors.'' These
tests are performed to verify the essentially leak tight
characteristics of the primary containment at the design basis
accident pressure. The proposed changes for a one-time extension of
the Type A tests do not effect the method for Type A, B or C testing
or the test acceptance criteria.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania

Date of amendment request: June 5, 2002, as supplemented August 19,
2002.

[[Page 75877]]

Description of amendment request: The requested amendments would
change the plant technical specifications (TSs) to allow plant
operation with the associated containment at atmospheric pressure. The
plant TSs currently require the containment to be maintained at sub-
atmospheric pressures when its associated unit is in operation. Minor
editorial, formatting, and pagination changes will also be made as
necessary to incorporate the revisions into the TSs.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Beaver Valley Power Station (BVPS) containments are designed
to withstand the internal pressure and temperature resulting from a
loss of coolant accident (LOCA), main steamline break (MSLB),
feedwater line break, and a control rod ejection accident (CREA).
All of these accidents have been previously analyzed in the Updated
Final Safety Analysis Report (UFSAR) except the feedwater line
break. This is not analyzed because the MSLB is most limiting. The
effect on containment pressure and temperature due to a CREA is
bounded by a LOCA, since a CREA is modeled as a small break LOCA.
The probability of occurrence for these accidents is independent of
the type of containment. Therefore a change from a subatmospheric to
an atmospheric containment will not increase the probability of
these accidents.
The revised containment integrity analysis demonstrates that the
pressures and temperatures associated with the applicable design
basis accidents identified above are within the existing containment
design limits. From a containment integrity viewpoint, the limiting
design basis accidents (DBA) presently are the MSLB for Unit 1 and
the LOCA for Unit 2. Following the conversion to an atmospheric
containment, the limiting DBA will be the MSLB for both units. The
effects of the proposed changes on plant structures, systems and
components (SSC) have been evaluated and verify that the capability
of the SSCs to perform their design functions will be retained
following approval of the proposed changes. The revised radiological
analysis reflects a selective application of the Alternative Source
Term (AST) of Regulatory Guide 1.183, ``Alternative Radiological
Source Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors,'' and incorporation of the ARCON96 methodology for on-site
atmospheric dispersion factors. The revised radiological analysis
concludes that normal operation of the BVPS units with atmospheric
containments will not impact either unit's compliance with the
operator exposure limits set forth in 10CFR20, or with the public
exposure limits set forth by 10CFR50, Appendix I.
For accident conditions, the proposed changes will potentially
impact the reported dose consequences of the LOCA, CREA and MSLB for
both BVPS units, and the locked rotor accident (LRA) for BVPS Unit
1. The radiological consequences of the remaining design bases
accidents are not adversely impacted by the proposed changes.
The revised radiological analysis concludes that site boundary
and control room dose consequences of the LOCA and the CREA remain
within the regulatory requirements of 10CFR50.67, as supplemented by
Regulatory Guide 1.183. It also concludes that the control room
doses for the MSLB for both BVPS units, and LRA for BVPS Unit 1 will
continue to remain within the regulatory limits provided in SRP 6.4
[NUREG-0800, ``Standard Review Plan for the Review of Safety
Analysis Reports for Nuclear Power Plants,'' section 6.4].
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The design basis accidents, which could be adversely affected by
the proposed changes, have been reanalyzed. These analyses
demonstrate that all acceptance criteria have been satisfied. The
revised containment integrity analysis demonstrates that the
containment will not be subjected to temperatures or pressures that
are beyond its design limits. Converting to an atmospheric
containment will not result in any new or different kind of
accidents because no new accident initiators will be introduced.
Changes to instrumentation setpoints, system flow rates,
surveillance requirements, and the elimination of certain
operability requirements will not have any [effect] that could
create the possibility of a new or different type of accident since
none of these changes would result in any changes to the manner in
which the affected equipment is operated. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety attributed to the containment involves both
the pressures and temperatures the containment is subjected to
following a DBA, and the on-site and offsite dose consequences
associated with normal and post DBA operations.
The revised containment integrity analysis conducted to support
the proposed changes demonstrate that the containment peak pressure
and temperature following a DBA will not exceed the containments'
design limits. Since the containment design limits are not exceeded,
the existing margin of safety between these limits and the
containment failure limits is not reduced.
The revised radiological analysis concludes that the existing
dose consequence margin of safety is not significantly reduced.
Therefore, the proposed change does 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.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania

Date of amendment request: October 31, 2002.
Description of amendment request: The proposed amendments would
revise the Beaver Valley Technical Specifications (TS) to allow
extending the Type A Containment Integrated Leak Rate Test (ILRT)
interval from 10 years to 15 years on a one-time basis.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated.
Response: No.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change allows a one-time extension to the current
surveillance interval for the Type A Containment Integrated Leak
Rate Test (ILRT). The current test interval of ten years, based on
performance history, would be extended on a one-time basis to 15
years from the last Type A test. The proposed change will not result
in a significant increase in the risk of plant operation. The risk
analysis was performed in accordance with Regulatory Guide 1.174 and
shows that the increase in total plant risk due to the extended ILRT
interval is 0.005 percent (Unit 1) and 0.02 percent (Unit 2). The
delta-large early release frequency (LERF) is 1.91E-9 /yr (Unit 1)
and 1.35E-9 /yr (Unit 2) when the test interval is increased from 10
to 15 years. These delta-LERF values meet the Regulatory Guide 1.174
acceptance criterion of less than 1.0E-07 per year for LERF. The
proposed extension to Type A testing does not increase the
probability of an accident previously

[[Page 75878]]

evaluated, since the containment Type A test does not involve any
modifications, nor a change in the way that any plant structures,
systems or components (SSC) function, and does not involve an
activity that could lead to equipment failure or accident
initiation. The proposed extension of the test interval does not
involve a significant increase in the consequences of an accident,
since the study documented in NUREG-1493, has found that
generically, very few potential leak paths are not identified with
Type B and C tests. NUREG-1493 concluded that an increase in the
Type A test interval to twenty years resulted in an imperceptible
increase in risk. Containment testing and inspection provide a high
degree of assurance that the containment will not degrade in a
manner only detectable by Type A testing. Inspections required by
the ASME Code and the Maintenance Rule are performed in order to
identify indications of containment degradation that could affect
leak tightness. Type B and C testing requirements and intervals
required by 10 CFR 50 Appendix J are not affected by this proposed
extension to the Type A test interval, and will identify any
potential openings in containment penetrations that would otherwise
require a Type A test. The increase in risk of the proposed change,
as measured by the change in LERF is within the acceptance criterion
of Regulatory Guide 1.174, therefore there will not be a significant
increase in the consequences of any accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in operation of the units in
a way that would create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
extension to Type A testing does not create a new or different type
of accident because no physical modifications are being made, and no
compensatory measures are being imposed that could potentially lead
to a failure. There are no changes to unit operation that could
introduce a new failure mode or create a new or different kind of
accident. The proposed change only allows a one-time extension to
the current interval for Type A testing and does not change the
implementation aspects of the subsequent test.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety.
Response: No.
The proposed change will not result in a significant reduction
in a margin of safety. The proposed change is for a one-time
extension to the current interval for Type A testing. The current
test interval of ten years, based on historical performance, will be
extended on a one-time basis to 15 years from the last Type A test.
The NUREG-1493 study of the effects of extending the Type A test
interval out to 20 years concluded that there is an imperceptible
increase in plant risk. Additionally, the extended test interval
will have a minimal effect on plant risk, since Type B and C testing
detect over 95% of potential leakage paths. The plant specific risk
analysis determined results that are consistent with the conclusions
of NUREG-1493. The overall increase in the risk contribution due to
the proposed change was determined to be 0.005 percent (Unit 1) and
0.02 percent (Unit 2). The delta-LERF is 1.91E-9/yr (Unit 1) and
1.35E-9/yr (Unit 2) when the test interval is increased from 10 to
15 years. The calculated impact on risk is insignificant, and meets
the acceptance criterion of Regulatory Guide 1.174.
Therefore, the proposed change does 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.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: June 4, 2002.
Description of amendment request: The proposed amendment proposes a
revision of pressure/temperature (P/T) limit curves for non-nuclear
heatup/cooldown, core critical operation, and pressure testing for
reactor coolant systems (RCSs); including an exemption request pursuant
to 10 CFR 50.60(b).
Basis for proposed no significant hazards consideration
determination: 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed P/T limit curves are based upon the use of an
alternate material fracture toughness curve and the use of an NRC-
approved methodology for calculation of neutron fluence. The
proposed RCS P/T limit curves are valid through 22 Effective Full-
Power Years (EFPY) and 32 EFPY.
The American Society of Mechanical Engineers (ASME) Boiler &
Pressure Vessel (B&PV) Code Case N-640 permits the use of
Klc as defined in ASME B&PV Code, Section XI, Appendix A,
Figure A-4200-1 instead of Kla as defined in ASME B&PV
Code, Section XI, Appendix G, Figure G-2210-1. The use of the
Klc curve in determining the lower bound fracture
toughness in the development of P/T limit curves is more technically
correct than the Kla curve. The Klc curve
models the slow heatup and cooldown processes that a Reactor
Pressure Vessel (RPV) normally undergoes. These slow heatup and
cooldown limits are enforced through the use of the PNPP [Perry
Nuclear Power Plant] Technical Specification 3.4.11, ``RCS Pressure
and Temperature (P/T) Limits.'' Surveillance Requirement 3.4.11.1
states that heatup and cooldown rates will be <=100 [deg]F in any
one hour period. The use of the Klc curve is applicable
to PNPP and is inconsistent with the ASME B&PV. Therefore, the use
of Klc will provide an adequate margin of safety to
protect against potential RPV failure.
NRC [Nuclear Regulatory Commission] regulations require the
vessel material transition temperature be adjusted to account for
the effects of neutron radiation. Regulatory Guide 1.190,
``Calculational and Dosimetry Methods for Determining Pressure
Vessel Neutron Fluence,'' provides a methodology for calculating the
neutron fluence, while Regulatory Guide 1.99, ``Radiation
Embrittlement of Reactor Vessel Materials,'' provides the guidance
for calculating the adjusted transition temperature using the
fluence factor. The methodologies satisfy the requirements of 10 CFR
[part] 50, Appendices G and H, and General Design Criteria 31,
``Fracture Prevention of Reactor Coolant Pressure Boundary.'' The
methodologies used to develop the proposed P/T limit curves satisfy
the requirements of the regulations.
The predicted lowest upper shelf energy at 32 EFPY was greater
than the minimum of 50 ft-lbs required by 10 CFR [part] 50, Appendix
G. The adjusted reference temperature for the limiting material was
less than the 200 [deg]F limit required by Regulatory Guide 1.99,
Revision 2. Therefore, the integrity of the RCS has been maintained.
As such, the proposed curves ensure that adequate reactor vessel
safety margins against nonductile failure exist during normal
operation, anticipated operational occurrences, and hydrostatic
testing. There are no plant modifications associated with these
changes. Thus, the proposed changes do not involve a significant
increase in the probability of occurrence of an accident previously
evaluated.
The proposed changes do not adversely affect the integrity of
the reactor vessel. Hence, the function of the reactor vessel to act
as a radiological barrier during an accident is not affected.
Therefore, the proposed changes do not involve a significant
increase in the consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed P/T limit curves are based upon the use of an
alternate material fracture toughness curve and the use of an NRC-
approved methodology for calculation of neutron fluence.

[[Page 75879]]

The ASME B&PV Code Case N-640 permits the use of the
Klc curve in determining the lower bound fracture
toughness in the development of P/T limit curves. The Klc
curve models the slow heatup and cooldown processes that a RPV
normally undergoes. These slow heatup and cooldown limits are
enforced through the use of the PNPP Technical Specifications.
Therefore, the use of Klc will provide an adequate margin
of safety to protect against potential RPV failure.
NRC regulations require the vessel material transition
temperature be adjusted to account for the effects of neutron
radiation. The methodologies used to develop the proposed P/T limit
curves satisfy the requirements of the regulations. The predicted
lowest upper shelf energy at 32 EFPY was greater than the minimum of
50 ft-lbs required by 10 CFR [part] 50, Appendix G. The adjusted
reference temperature for the limiting material was less than the
200 [deg]F limit required by Regulatory Guide 1.99, Revision 2.
Therefore, the integrity of the RCS has been maintained. As such,
the proposed curves ensure that adequate reactor vessel safety
margins against nonductile failure exist during normal operation,
anticipated operational occurrences, and hydrostatic testing.
There are no plant modifications associated with these changes.
The proposed changes to the P/T limit curves do not affect the
assumed accident performance of any structure, system, or component
previously evaluated. The proposed changes do not introduce any new
modes of system operation or failure mechanisms. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
NRC regulations require that P/T limits provide an adequate
margin of safety to the conditions at which brittle fracture may
occur. These regulations are set forth in 10 CFR [Part] 50, Appendix
A, General Design Criteria (GDC) 31, and 10 CFR [Part] 50,
Appendices G and H. Regulatory Guides 1.99 and 1.190 provide
guidance for the compliance of GDC 31 and Appendices G and H. The
appendices reference the requirements and guidance of ASME B&PV
Code, Section XI, Appendix G for the development of P/T limit
curves. The methodologies described within the regulatory guides and
the ASME Code will provide P/T limit curves with the requisite
margin against brittle fracture. The proposed P/T limit curves are
based on these methodologies as modified by application of ASME Code
Case N-640.
Although the code case proposes a change to a requirement
contained in ASME, Section XI, Appendix G, the alternative allowed
by Code Case N-640 is based upon industry experience gained since
the inception of 10 CFR [Part] 50, Appendix G. The more appropriate
assumptions and provisions allowed by the code case maintain a
margin of safety that is consistent with the intent of 10 CFR [Part]
50, Appendices G and H. Therefore, the proposed changes do not
involve a significant reduction in the margin of safety.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire

Date of amendment request: October 11, 2002.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3.4.9.1, ``Reactor Coolant System
[RCS]--Pressure/Temperature Limits'' and TS 3.4.9.3, ``Reactor Coolant
System--Overpressure Protection Systems'' and their associated Bases
sections. Specifically, the proposed changes will replace TS Figure
3.4-2, ``Reactor Coolant System Heatup Limitations,'' Figure 3.4-3,
``Reactor Coolant System Cooldown Limitations,'' and Figure 3.4-4,
``RCS Cold Overpressure Protection Setpoints,'' to allow operation to
20 Effective Full Power Years (EFPY). The proposed change to TS 3.4.9.3
will also revise the Cold Overpressure Protection System arming
temperature from 329[deg]F to 290[deg]F to reflect the higher allowable
low temperature overpressure protection pressure limit afforded by the
use of ASME Code Case N-641.
Basis for proposed no significant hazards consideration
determination: 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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to TS 3.4.9.1 and TS 3.4.9.3 do not result
in a condition where the design, material, and construction
standards that were applicable prior to the proposed changes are
altered. The probability of occurrence of an accident previously
evaluated for Seabrook Station is not altered by the proposed
amendment to the TSs. The accidents remain the same as currently
analyzed in the UFSAR [Updated Final Safety Analysis Report] as a
result of changes to the P/T limits as well as those for Cold
Overpressure Mitigation System (COMS). The new P/T limits are based
on NRC [Nuclear Regulatory Commission] accepted methodology along
with [the] American Society of Mechanical Engineers (ASME) Code
alternative methodology. An exemption request to allow use of the
alternative ASME methodology is included as part of this LAR
[License Amendment Request]. The proposed COMS setpoint limit based
on the revised P/T limits satisfies the criteria specified in the
alternative ASME methodology and 10 CFR part 50 Appendix G closure
head/vessel flange region pressure limit criteria. The proposed
changes do not impact the integrity of the reactor coolant pressure
boundary (RCPB) i.e. there is no change to the operating pressure,
materials, system loadings, etc., as a result of this change. In
addition, there is no increase in the potential for the occurrence
of a loss of coolant accident. The probability of any design basis
accident is not affected by this change, nor are the consequences of
any design basis accident (DBA) affected by this proposed change.
The proposed P/T limit curves and the COMS limits are not considered
to be an initiator or contributor to any accident currently,
evaluated in the Seabrook Station UFSAR. These new limits ensure the
long term structural integrity of the RCPB.
Fracture toughness test data are obtained from beltline material
specimens contained in surveillance capsules that are periodically
withdrawn from the reactor vessel. This data allows determination of
time conditions under which the vessel can be operated with adequate
safety margins against non-ductile fracture throughout its service
life. The second Seabrook Station surveillance capsule was removed
from the reactor vessel after completion of Operating Cycle No. 5 in
May 1997 and was analyzed to predict the fracture toughness
requirements using projected neutron fluence calculations. For each
analyzed transient and steady state condition, the allowable
pressure is determined as a function of reactor coolant temperature
considering postulated flaws in the reactor vessel beltline region
material. The predicted radiation induced [Delta]RTNDT
was calculated using the respective reactor vessel beltline
materials copper and nickel contents and the neutron fluence
predicted for 20 EFPY. The RTNDT and, accordingly, the
operating limits for Seabrook Station were adjusted to account for
the effects of irradiation on the fracture toughness of the reactor
vessel beltline materials. Therefore, new operating limits are
established which are represented in the revised operating curves
for heatup/cooldown, criticality and inservice hydrostatic testing
contained in the technical specifications. The proposed P/T limit
curves and COMS setpoint limits are not considered to be an
initiator or contributor to any accident currently evaluated in the
Seabrook Station UFSAR.
Therefore based on the above discussion, it is concluded that
the proposed revisions to TS 3.4.9.1 and TS 3.4.9.3 do not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.

[[Page 75880]]

The proposed changes to the P/T and COMS limits will not create
a new accident scenario. The requirements to have P/T and COMS
protection are part of the licensing basis for Seabrook Station. The
proposed technical specification amendment reflects the change in
reactor vessel material properties as determined by evaluation of
the most recently withdrawn surveillance capsule. Based on the
surveillance capsule data, the adjusted RTNDT values for
the plate and weld material were within the two standard deviations
of Regulatory Guide 1.99, Revision 2 predictions. As all the
requisite criteria of Regulatory Guide 1.99, Revision 2 was
satisfied, it was concluded that the surveillance data was credible
and the beltline material was responding as empirically predicted.
The new P/T limits are based on NRC accepted methodology along with
American Society of Mechanical Engineers (ASME) Code alternative
methodology. An exemption request to allow use of the alternative
ASME methodology is included as part of this LAR. The proposed COMS
setpoint limit based on the revised P/T limits satisfies the
criteria specified in the alternative ASME methodology and 10 CFR
part 50 Appendix G closure head/vessel flange region pressure limit
criteria. The proposed changes will not alter the way any structure,
system or component functions, and will not significantly alter the
manner in which the plant is operated. There will be no adverse
effect on plant operation or accident mitigation equipment.
Since no new failure modes are created by the proposed revisions
to TS 3.4.9.1 and TS 3.4.9.3, this change does not create the
possibility of a new or different kind of accident from any that was
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The existing P/T and COMS limit curves in the technical
specifications are reaching their expiration for the number of years
at effective full power operation. The revision of the P/T limits
and COMS will ensure that Seabrook Station continues to operate
within the operating limits allowed by 10 CFR 50.60 and the ASME
Code. The material properties used in the development of the revised
limit curves are based on the evaluation of the most recently
withdrawn surveillance capsule. The application of ASME Code Case N-
641 presents alternative methods for calculating P/T and COMS
temperature and pressure limits in lieu of those established in ASME
Section XI, Appendix G-2215. This ASME Code alternative allows
analysis features that are less restrictive than those associated
with previous methodologies, however these features remain
conservative with respect to the requirements delineated ASME
Section XI. Therefore it is concluded that the revised P/T and COMS
limit curves proposed by this technical specification amendment
still provide sufficient margin to preclude non-ductile fracture of
the reactor vessel.
Thus, it is concluded that these proposed revisions to TS
3.4.9.1 and TS 3.4.9.3 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
PO Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief (Acting): James W. Andersen.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire

Date of amendment request: October 11, 2002.
Description of amendment request: The proposed amendment would
relocate Technical Specifications (TSs) 3.1.2.1, ``Reactivity Control
Systems-Borations Systems-Flow Paths-Shutdown;'' 3.1.2.2, ``Reactivity
Control Systems-Boration Systems-Flow Paths-Operating;'' 3.1.2.3,
``Reactivity Control Systems-Boration Systems-Charging Pumps-
Shutdown;'' 3.1.2.4, ``Reactivity Control Systems-Boration Systems-
Charging Pumps-Operating;'' 3.1.2.5, ``Reactivity Control Systems-
Boration Systems-Borated Water Sources-Shutdown;'' 3.1.2.6,
``Reactivity Control Systems-Boration Systems-Borated Water Sources-
Operating;'' and 3.4.7, ``Reactor Coolant System-Chemistry,'' to the
Seabrook Station Technical Requirements Manual (SSTR) and would revise
TS 3.1.2.7, ``Reactivity Control Systems-Boration Systems-Isolation of
Unborated Water Sources-Shutdown.'' The proposed amendment would also
revise TSs 3.4.1.2, ``Reactor Coolant System-Reactor Coolant Loops and
Coolant Recirculation-Hot Standby,'' 3.4.3 ``Reactor Coolant System-
Pressurizer,'' 3.4.7, ``Reactor Coolant System-Chemistry,'' and 3.9.2,
``Refueling Operations-Instrumentation,'' to adopt a portion of NUREG-
1431, Revision 2, ``Standard Technical Specifications, Westinghouse
Plants,'' involving a wording revision to more closely match Standard
Technical Specifications. The revision to TS 3/4.9.2 would also involve
surveillance changes. The associated Bases would also be modified as a
result of the proposed changes.
Basis for proposed no significant hazards consideration
determination: 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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The TS changes propose the relocation of the boration subsystem
and chemistry requirements to a licensee-controlled document. The
relocation of these requirements will not cause an accident to occur
and will not result in any change in the operation of the associated
accident mitigation equipment. Therefore, the proposed changes will
not increase the probability or consequences of an accident
previously evaluated.
The TS changes propose the modification of the TS for
``Isolation of Unborated Water Sources--Shutdown.'' Only the
demineralizers that are intended to deborate the Reactor Coolant
System will need to be isolated in MODE 4, 5, or 6. Administrative
controls, currently in use for the operation of the Boron Thermal
Regeneration System and replenishment of demineralizer resin in the
Chemical Volume and Control System, will be used to minimize the
affects of an inadvertent dilution due to operation of the
demineralizers. The Seabrook Station Updated Final Safety Analysis
currently includes a boron dilution event analysis for each MODE of
operation. Use of these administrative controls will ensure that the
operation of the BTRS [Boron Thermal Regeneration System] is bounded
by the boron dilution analysis. Therefore, the modification of the
TS requirement will not increase the probability or consequences of
an accident previously evaluated.
The TS changes propose to change the source range flux monitor
requirements in MODE 6. The proposed change does not significantly
affect the operability of the associated equipment. The source range
neutron flux monitors are components not assumed to be initiators of
analyzed events. Therefore, the change in the TS requirement for the
source range instrumentation in MODE 6 will not increase the
probability or consequences of an accident previously evaluated.
The additional proposed changes to the TS that will standardize
terminology, relocate information to the Bases, remove extraneous
information, modify the requirements to prevent rod withdrawal for
operational flexibility, and make minor format changes will not
result in any technical changes to the current requirements.
Therefore, these additional proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes to the TSs do not impact any system or
component that could cause an accident, nor will it alter the plant
configuration or require any unusual operator actions, nor will it
alter the way any structure, system, or component functions.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in [a] margin of safety.
The proposed TS changes associated with the relocation of the
boration subsystem and

[[Page 75881]]

chemistry requirements to a licensee-controlled document will not
result in a significant reduction in a margin of safety.
The proposed TS changes associated with the modification of the
TS for ``Isolation of Unborated Water Sources--Shutdown,'' are
consistent with the requirements contained in the Seabrook Station
Updated Final Safety Analysis which currently includes a boron
dilution event analysis for each MODE of operation. The changes
result in operation within the parameters specified by the analysis.
Therefore, the modification of the TS requirement will not result in
a significant reduction in a margin of safety.
The proposed TS changes associated with the source range flux
monitor do not significantly affect the operability of the
associated equipment. Therefore, the change in the TS requirement
for the source range instrumentation will not result in a
significant reduction in a margin of safety.
The additional proposed changes to the TSs that will standardize
terminology, relocate information to the Bases, remove extraneous
information, modify requirements to prevent rod withdrawal for
operational flexibility, and make minor format changes will not
result in any technical changes to the current requirements.
Therefore, these additional changes will not result in 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. S. Ross, Florida Power & Light
Company, PO Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief (Acting): James W. Andersen.

Florida Power and Light Company (FPL), et al., Docket Nos. 50-335 and
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: October 23, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) Section 5.6, ``Design
Features--Fuel Storage,'' to include the design of a new cask pit spent
fuel storage rack for each unit to increase the allowable spent fuel
wet storage capacity at both units and include the description of
Boral^TM as the neutron absorbing material used in the new
cask pit storage racks. The proposal also revises the spent fuel pool
(SFP) thermal-hydraulic analyses for core offload times of 120 hours
after reactor shutdown and for a partial core offload as the normal
offload condition. In addition the proposal includes a change in FPL's
commitments regarding the Unit 2 spent fuel cooling system design basis
described in the Updated Final Safety Analysis Report (UFSAR). A
current UFSAR commitment regarding the Unit 2 peak SFP temperature
limit during full core offloads with minimum SFP cooling will be
replaced with a new design basis.
Basis for proposed no significant hazards consideration
determination: 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. Would operation of the facility in accordance with the
proposed amendments involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes to increase the spent fuel storage
capacity with cask pit racks were evaluated for impact on the
following previously evaluated events:
a. A fuel handling accident (FHA),
b. A heavy load drop into the cask pit,
c. A loss of SFP cooling,
d. A stored fuel criticality event,
e. A seismic event.
The probability of a fuel handling accident is not significantly
increased by the proposed changes, because the same equipment (e.g.,
the spent fuel handling crane) and procedures will be used to handle
fuel assemblies and the frequency of fuel movement will be
essentially the same, with or without cask pit racks. The FHA
radiological consequences are not significantly increased because
the source term of a single fuel assembly will remain unchanged, and
the cask pit racks will be installed at the same water depth as the
existing SFP racks, with the same iodine decontamination factors
assumed in the FHA analysis. The structural consequences of dropping
a fuel assembly on a cask pit rack were also found to be no more
severe than those in the current FHA analysis.
The probability and consequences of a heavy load drop of the
cask pit rack or its platform are bounded by the existing cask drop
analyses, because a fuel transfer cask is much heavier than either
the empty rack or platform, and cask handling will be a more
frequent operation in the future than cask pit rack installation and
removal. The cask pit rack will be removed prior to any cask
handling operations, such that a cask drop scenario onto a cask pit
rack loaded with fuel is not credible. Therefore, the probability
and the consequences of a heavy load drop in the cask pit are not
significantly increased.
The probability of a loss of SFP cooling is unaffected and its
consequences are not significantly increased with cask pit racks
installed. With the cask pit rack installed, loss of forced cooling
results in a sufficient time-to-boil for the operator to recognize
the condition and establish SFP makeup to compensate for water lost
due to pool bulk boiling, and thereby maintain a sufficient water
blanket over the stored spent fuel.
The probability and consequences of a stored fuel criticality
event are not increased by the addition of a cask pit rack. The
reactivity analysis for the new racks demonstrates that reactivity
remains subcritical (below 0.95) for the worst-case fuel
mispositioning event, without credit for soluble boron. The
probability of a seismic event is unaffected and its consequences
are not significantly increased with cask pit racks installed,
because the structural analysis of the new racks demonstrates that
the fuel storage function of the rack is unimpaired by loading
combinations including seismic motion, and there is no adverse
seismic-induced interaction between the rack and adjacent
structures.
Based on the above, it is concluded that the proposed amendments
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Would operation of the facility in accordance with the
proposed amendments create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed changes to add a cask pit rack to each unit do
not alter the operating requirements of the plant or of the
equipment credited in the mitigation of design basis accidents, nor
do the proposed changes affect any of the important parameters
required to ensure the safe storage of spent fuel. A new rack
material (Boral^TM) is introduced into the pool under
these changes, but based on its operating history in SFPs, there are
no mechanisms that create a new or different kind of accident. The
potential for dropping the new rack or its platform during
installation or removal is bounded by the existing analysis for
dropping a spent fuel transfer cask into the cask pit. The same
equipment (e.g., the spent fuel handling crane) and procedures will
be used to handle fuel assemblies for the new cask pit racks as are
used for existing spent fuel storage. The fuel storage configuration
in the new racks will be similar to the configuration in the
existing SFP storage racks, and a fuel drop or mispositioning event
in the new racks does not represent a new or different kind of
accident from fuel handling and mispositioning events previously
evaluated. Therefore, the proposed amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Would operation of the facility in accordance with the
proposed amendments involve a significant reduction in a margin of
safety?
No. The effect of the proposed changes on current margins of
safety were evaluated for spent fuel storage functionality and
criticality, spent fuel and SFP cooling, and SFP/cask pit structural
integrity. The design of the new racks uses proven technology which
preserves the proper safety margins for spent fuel storage to
provide a coolable and subcritical geometry under both normal and
abnormal/accident conditions. The design complies with current
regulatory guidelines and the ANSI [American National Standards
Institute] standards, including 10 CFR 50 Appendix A GDC [General
Design Criterion] 62, NUREG-0800 Section 9.1.2, the OT Position for
Review and Acceptance of Spent Fuel Storage and Handling
Applications,

[[Page 75882]]

Regulatory Guide 1.13, and ANSI/ANS [American Nuclear Society] 8.17.
Handling the racks and platforms in accordance with the defense-in-
depth approach of NUREG-0612 with temporary lift items designed to
ANSI N14.6 preserves the proper margin of safety to preclude a heavy
load drop in the cask pit.
The proposed SFP cooling system design basis is consistent with
the regulatory guidance in NRC Standard Review Plan Section 9.1.3
for SFP temperature limits during normal and abnormal core offload
conditions. The rack and SFP thermal hydraulic analyses demonstrate
that the proposed SFP cooling system design basis is met, and that
no bulk boiling will occur in the new rack or SFP with minimum
cooling available. A loss of SFP cooling will allow sufficient time
for operators to identify the condition and initiate makeup flow or
restore cooling to preserve fuel cooling capability.
The new rack criticality analyses demonstrate that the
subcriticality safety margin is maintained below 0.95 under all
conditions, without credit for soluble boron. The structural
analyses for the new racks and adjacent structures show that the
rack and surrounding structures are unimpaired by loading
combinations during seismic motion, and there is no adverse seismic-
induced interaction between the rack and adjacent structures. Based
on these evaluations, operating the facility with the proposed
amendments does not involve a significant reduction in any margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Allen G. Howe.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska

Date of amendment request: September 26, 2002.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a missed
surveillance. The delay period would be extended from the current limit
of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the
limit of the specified Frequency, whichever is greater.'' In addition,
the following requirement would be added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater than
24 hours and the risk impact shall be managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714).
The licensee affirmed the applicability of the following NSHC
determination in its application dated September 26, 2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated

The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated

The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety

The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: November 15, 2002.
Description of amendment request: The licensee proposes to revise
the reactor coolant system pressure-temperature (P-T) limit curves and
associated limit tables specified in Section 3/4.2.2, ``Minimum Reactor

[[Page 75883]]

Vessel Temperature for Pressurization,'' of the Technical
Specifications (TSs). The P-T limit curves and tabular listing of P-T
limit values contained in the revised figures and tables are based, in
part, on an alternative methodology and will be valid for 28 effective
full-power years. The alternative methodology has been endorsed by the
American Society of Mechanical Engineers.
The associated licensee-controlled TSs Bases pages would also be
changed to reflect the above TS changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed changes, if approved by the Nuclear Regulatory
Commission (NRC), will be made in a manner such that conservatism is
maintained through compliance with applicable NRC regulations and
guidance. No hardware design change is involved with the proposed
amendment, thus there will be no adverse effect on the functional
performance of any plant structure, system, or component (SSC). All
SSCs will continue to perform their design functions with no decrease
in their capabilities to mitigate the consequences of postulated
accidents. P-T limit curves were not previously factored into the
probability of accidents, nor were they factored into scenarios of
previously analyzed accidents. Accordingly, the revised P-T limit
curves and tabular listing of P-T limit values will lead to no increase
in the consequences of an accident previously evaluated, and no
increase of the probability of an accident previously evaluated.
The second standard requires that operation of the unit 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 amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, the proposed amendment
will not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, 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.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey

Date of amendment request: July 25, 2002, as supplemented October
21, 2002.
Description of amendment request: The amendment would modify
Technical Specification (TS) requirements for missed surveillance tests
in TS 4.0.3 using the Consolidated Line Item Improvement Program,
modify TS 4.0.1 to be consistent with the Standard Technical
Specifications (STS), and incorporate a TS Bases Control Program in
Section 6.0 in accordance with the STS.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?

Specification 4.0.3

The proposed change relaxes the time allowed to perform a missed
Surveillance. The time between Surveillances is not an initiator to
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be OPERABLE and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected.

Specification 4.0.1

The proposed additional requirement equating failure to meet a
surveillance with failure to meet the LCO [limiting condition for
operation] is consistent with current interpretation of the
technical specifications. This change, along with relocation and
rewording of existing requirements from Specification 4.0.3, are
administrative in nature and do not adversely affect accident
initiators, design functions, facility configuration or the manner
of operation or control. The ability of structures, systems and
components to perform their intended function remains unaffected.

Bases Control Program

The proposed change to adopt a Technical Specification Bases
Control Program is also administrative in nature and does not
adversely affect accident initiators, design functions, facility
configuration or the manner of operation or control. The ability of
structures, systems or components to perform their intended function
remains unaffected. Future changes to the TS Bases will continue to
be administratively controlled in accordance with the requirements
of 10 CFR 50.59.
Therefore, these three changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
None of the three proposed changes involves a physical
alteration of the plant (no new or different type of equipment will
be installed) or a change in the methods governing normal plant
operation. Thus, these changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?

Specification 4.0.3

The relaxed time allowed to perform a missed Surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
Surveillance is verification that the LCO is met. Failure to perform
a Surveillance within the prescribed Frequency does not cause
equipment to become inoperable. The only effect of the additional
time allowed to perform a missed Surveillance on the margin of
safety is the extension of the time until inoperable equipment is
discovered to be inoperable by the missed Surveillance. However,
given the rare occurrence of inoperable equipment, and the rare
occurrence of a missed Surveillance, a missed Surveillance on
inoperable equipment would be very unlikely. This must be balanced
against the real risk of manipulating the plant equipment or
condition to perform the missed Surveillance. In addition, parallel
trains and alternate equipment are typically available to perform
the safety function of the equipment not tested.

Specification 4.0.1

The proposed changes to TS 4.0.1, including relocation and
rewording of

[[Page 75884]]

existing requirements from Specification 4.0.3, are administrative
in nature and do not reduce the level of programmatic or procedural
controls associated with the Surveillance Requirements. There are no
substantive differences in meaning or intent between the existing
specifications and the corresponding STS requirements. Further,
these changes have no impact on equipment design, configuration,
analytical basis, setpoints or operation.

Bases Control Program

The proposed change to adopt a Technical Specification Bases
Control Program is also administrative in nature and does not reduce
the level of programmatic or procedural controls associated with the
Bases. There is no impact on equipment design, configuration,
analytical basis, setpoints or operation.
Thus, there is confidence that the equipment can perform its
assumed safety function. Therefore, this change does not involve a
significant reduction in a margin of safety.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey

Date of amendment request: October 23, 2002.
Description of amendment request: The proposed change updates the
reference to 10 CFR 20.203 with the corresponding reference to 10 CFR
20.1601.
Basis for proposed no significant hazards consideration
determination: 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. The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not affect accident initiators or
precursors and do not alter the design assumptions, conditions,
configuration of the facility, or manner in which the plant is
operated. The proposed changes do not alter or prevent the ability
of structures, systems, or components to perform their intended
safety function to mitigate the consequences of an initiating event
within the acceptance limits assumed in the UFSAR [Updated Final
Safety Analysis Report]. The proposed changes are administrative in
nature. Technical Specification (TS) 6.12 will be updated to include
the new 10 CFR 20 (effective 06/20/91) requirements. The proposed
changes do not alter the conditions or assumptions in any of the
previous accident analyses, and as a result, the radiological
consequences associated with these analyses remain unchanged.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter the design assumptions,
conditions, configuration of the facility, or the manner in which
the plant is operated.
The proposed changes are administrative in nature and the
relocated procedural details do not change the level of programmatic
controls and procedural details. Accordingly, the proposed changes
do not create any new failure modes or limiting single failures
associated with a plant structure, system, or component important to
safety. Also, there will be no change in the types or increase in
the amounts of any effluents released offsite.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment would not involve a significant
reduction in the margin of safety.
The proposed changes do not impact equipment design or
operation, nor do the changes affect any TS safety limits or safety
system settings that could adversely affect plant safety. The
proposed changes are administrative in nature. Technical
Specification (TS) 6.12 will be updated to include the new 10 CFR 20
requirements (effective 06/20/91) and are in conformance with NUREG-
1433 [Standard Technical Specifications General Electric Plants, BWR
4]. Furthermore, the proposed changes do not result in a change in
the types or an increase in the amounts of any effluents released
offsite.
Therefore, it is concluded that the proposed changes do not
involve a significant reduction in a margin of safety.

PSEG Nuclear LLC, Docket No. 50-354, Salem Nuclear Generating Station,
Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: September 20, 2002.
Description of amendment request: The proposed amendment will add
new limiting conditions for operation for fuel storage pool boron
concentration, fuel assembly storage in the spent fuel pool, relocate
requirements for spent fuel storage, revise existing Technical
Specification (TS) 3/4.9.1 for boron concentration during refueling
operations, and revise existing administrative controls associated with
the Core Operating Limits Report described in TS 6.9.1.9.
Basis for proposed no significant hazards consideration
determination: 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The postulated accidents are basically of three types. The first
type of postulated accident is an abnormal location of a fuel
assembly, the second type of postulated accident is associated with
lateral rack movement, and the third type of postulated accident is
a dropped fuel assembly on the top of the rack. The dropped fuel
assembly and the lateral rack movement have been previously shown to
have negligible reactivity effects (<0.0001 [delta k]). The
misplacement of a fuel assembly could have a small positive
reactivity effect, however, the negative reactivity effect of a
minimum soluble boron concentration of 600 ppm [parts per million]
compensates for the increased reactivity caused by any of the
postulated accident scenarios.
There is no increase in the probability of the accidental
misloading of irradiated fuel assemblies into the spent fuel pool
racks when considering the presence of soluble boron in the pool
water for criticality control. Fuel assembly placement will continue
to be controlled pursuant to approved fuel handling procedures and
will be in accordance with the Technical Specification (TS) spent
fuel rack storage configuration limitations.
There is no increase in the consequences of the accidental
misloading of irradiated fuel assemblies into the spent fuel pool
racks because criticality analyses demonstrate that the pool will
remain subcritical following an accidental misloading if the pool
contains an adequate boron concentration. This has been previously
evaluated in the Safety Evaluation by the Office of Nuclear Reactor
Regulation related to Amendment Nos[.] 151 and 131 to Facility
Operating Licenses DPR-70 and DPR-75 for the Salem Nuclear
Generating Station Units 1 and 2, dated May 4, 1994 (Spent Fuel
Reracking, TAC [technical

[[Page 75885]]

assignment control] NOS. M85797 and M85798). The proposed TS
limitations will ensure that an adequate spent fuel pool boron
concentration will be maintained.
The proposed change will revise the Salem Generating Station
(SGS) TS to be consistent with the improved Standard Technical
Specifications for Westinghouse plants, NUREG-1431 Revision 2, 4/30/
01. The new TS are not an accident initiator. Specifying a minimum
boron concentration in a new TS and relocating fuel assembly storage
requirements in a new TS are conservative approaches to operational
control.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously analyzed.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously analyzed?
Response: No.
Criticality accidents in the spent fuel pool have been analyzed
in the previous criticality safety analyses documented in PSEG
letter NLR-N93058 dated April 28, 1993 transmitting License Change
Request (LCR) 93-02 and Attachment D, The Licensing Report for Spent
Fuel Storage Capacity Expansion, Public Service Electric and Gas
Company, Salem Generating Stations 1 & 2, USNRC [U.S. Nuclear
Regulatory Commission] Docket Nos[.] 50-272 & 50-311, prepared by
Holtec International. This is the basis for the present TS. The
addition of a Limiting Condition for Operation (LCO) for boron
concentration does not alter the assumptions or the results of the
existing spent fuel criticality analyses or accident analyses
described in the Salem Updated Final Safety Analysis Report. The
addition of TS which provide for TS control where previous
administrative controls had been in place and relocation of material
within existing TS does not alter the results of criticality safety
analyses.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The TS changes proposed and the resulting spent fuel storage
operation limits will continue to provide adequate safety margin to
ensure that the stored fuel assembly array will remain subcritical.
Those limits are based on a plant specific criticality analysis and
are unchanged by this application. The addition of TS which provides
for TS control where previous administrative controls had been in
place and relocation of material within existing TS continue to
establish conservative operational control.
Therefore, the proposed change does not involve a significant
reduction in [a] margin of safety.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama

Date of amendment request: October 25, 2002.
Description of amendment request: The proposed amendment would
revise the numerical value of the Safety Limit Minimum Critical Power
Ratio (SLMCPR) in Technical Specification (TS) 2.1.1.2 to incorporate
the results of the cycle-specific core reload analysis for Browns Ferry
Unit 2 Cycle 13 operation.
Basis for proposed no significant hazards consideration
determination: 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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment establishes a revised SLMCPR value for
two recirculation loop operation. The probability of an evaluated
accident is derived from the probabilities of the individual
precursors to that accident. The proposed SLMCPR preserves the
existing margin to transition boiling and the probability of fuel
damage is not increased. Since the change does not require any
physical plant modifications or physically affect any plant
components, no individual precursors of an accident are affected and
the probability of an evaluated accident is not increased by
revising the SLMCPR value.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The revised SLMCPR has been determined using NRC-
approved methods and procedures. The basis of the MCPR Safety Limit
is to ensure no mechanistic fuel damage is calculated to occur if
the limit is not violated. These calculations do not change the
method of operating the plant and have no effect on the consequences
of an evaluated accident. Therefore, the proposed TS change does not
involve an increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed license amendment involves a revision of the SLMCPR
for two recirculation loop operation based on the results of an
analysis of the Cycle 13 core. Creation of the possibility of a new
or different kind of accident would require the creation of one or
more new precursors of that accident. New accident precursors may be
created by modifications of the plant configuration, including
changes in the allowable methods of operating the facility. This
proposed license amendment does not involve any modifications of the
plant configuration or changes in the allowable methods of
operation. Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety.
Response: No.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPR was calculated using NRC-approved methods and
procedures, which are in accordance with the current fuel design and
licensing criteria. The SLMCPR remains high enough to ensure that
greater than 99.9 percent of all fuel rods in the core are expected
to avoid transition boiling if the limit is not violated, thereby
preserving the fuel cladding integrity. Therefore, the proposed TS
change does not involve a reduction in the margin of safety.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant
(SQN), Unit 2, Hamilton County, Tennessee

Date of amendment request: November 15, 2002.
Description of amendment request: The proposed one-time condition
would establish special provisions and requirements for safe operation
of Unit 2 while heavy load lifts are performed on Unit 1. The
provisions for heavy load lifts are described in Topical Report 24370-
TR-C-002 that was previously submitted on April 15, 2002, for NRC
review and approval. The topical report contains prerequisite actions
for heavy load movement, active monitoring during heavy load movement,
and compensatory measures in response to the unlikely event of a heavy
load drop. This submittal withdraws an amendment request dated July 10,
2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the

[[Page 75886]]

issue of no significant hazards consideration, which is presented
below:

A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
No changes in event classification as discussed in SQN Updated
Final Safety Analysis Chapter 15 will occur due to the proposed
license amendment. The one-time provision ensures that the SQN ERCW
[essential raw cooling water] system remains functional for
continued safe operation of Unit 2 during heavy load lifts performed
on Unit 1 during SGR (steam generator replacement) replacement [sic]
activities.
Accordingly, the proposed modification to SQN Unit 2 operating
license and the implementation of compensatory measures for a
postulated load drop will not significantly increase the probability
or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The possibility of a new or different accident scenario
occurring as a result of activities conducted during the SQN Unit 1
SGR project are [sic] not created. Three postulated scenarios
related to heavy load handling during the SGR project were examined
for their potential to represent a new or different kind of accident
from those previously evaluated: (1) A breach of the old steam
generator (OSG), resulting in the release of contained radioactive
material, (2) flooding in the Auxiliary Building caused by the
failure of piping in the ERCW tunnel, and (3) loss of ERCW to
support safe shutdown of the operating unit.
Failure of an OSG that results in a breach of the primary side
of the steam generator (SG) could potentially result in a release of
a contained source outside containment. The consequences of this
event, both offsite and in the control room, were examined and found
to be within the consequences of the failure of other contained
sources outside containment at the SQN site (i.e., within the SQN
design basis).
With regard to flooding of the Auxiliary Building from a heavy
load drop, the protective measure taken prior to the lifting of
heavy loads include installation of a wall in the ERCW tunnel near
the Auxiliary Building interface. The wall provides protection
against a postulated flood of the ERCW tunnel and protects against
flooding of the Auxiliary Building beyond those events previously
evaluated.
With regard to the potential for a heavy load drop causing the
loss of ERCW cooling water to the operating unit (i.e., Unit 2), TVA
is implementing provisions to preclude a load drop. A heavy load
drop is considered an unlikely accident for the following reasons:
The lifting equipment was specifically designed and chosen for
the subject heavy lifts,

--Crane operators will be specially trained in the operation of the
lift equipment and in the SQN site conditions,
--Qualifying analyses and administrative controls will be used to
protect the lifts from the effects of external events,

The areas over which a load drop could cause loss of ERCW are a
small part of the total travel path of the loads.
In addition, protection against the potential for a loss of ERCW
is established prior to any heavy load lifts. Compensatory measures
ensure the ERCW system is isolated should a pipe break occur, and
that ERCW flow is redirected to equipment essential for safe
shutdown capability of Unit 2.
Accordingly, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to the Unit 2 operating license supports
safe operation and safe shutdown capabilities of Unit 2 during
replacement of the Unit 1 SGs. These measures do not result in
changes in the design basis for plant structures, systems, and
components (SSCs). Consequently, the proposed change will not affect
any margins of safety for plant SSCs.
Accordingly, a significant reduction in the margin of safety is
not created by the proposed change.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, (301)
415-4737 or by email to pdr@nrc.gov.

Carolina Power & Light Company, Docket Nos. 50-324, Brunswick Steam
Electric Plant, Unit 2, Brunswick County, North Carolina

Date of amendment request: November 26, 2001, as supplemented
January 31, February 5, February 11, and October 8, 2002.
Description of amendment request: The amendment revises the
Improved Technical Specification 5.5.12 to allow a one-time interval
increase for the Type A Integrated Leakage Rate Test for no more than 2
years, 2 months.
Date of issuance: November 21, 2002.
Effective date: November 21, 2002.
Amendment No.: 250.
Facility Operating License No. DPR-62: The amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: January 8, 2002 (67 FR
926). The January 31 and February 5, 2002, supplements contained
clarifying information only, and did not change the initial no
significant hazards consideration determination or expand the scope of
the initial Federal Register notice. The February 11 and October 8,
2002, supplements revised the original requests but the initial no
significant

[[Page 75887]]

hazards determination bounded the revised request.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2002.
No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: July 19, 2002, as supplemented
September 6, 2002.
Brief description of amendment: The amendment revised Technical
Specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. The delay period is extended from the
current limit of ``* * * up to 24 hours'' to `` * * * up to 24 hours or
up to the limit of the specified surveillance interval, whichever is
greater.'' In addition, the following requirement is added to SR 4.0.3:
``A risk evaluation shall be performed for any surveillance delayed
greater than 24 hours and the risk impact shall be managed.'' The
amendment also made administrative changes to SRs 4.0.1 and 4.0.3 to be
consistent with NUREG-1431, Revision 2, ``Standard Technical
Specifications--Westinghouse Plants.''
Date of issuance: November 15, 2002.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 213.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 4, 2002 (67
FR 56604).
The September 6, 2002, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 15, 2002.
No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington

Date of application for amendment: July 16, 2002, as supplemented
by letter dated September 4, 2002.
Brief description of amendment: The amendment changes the technical
specifications (TS) to revise the specified minimum emergency diesel
generator (DG) steady state output voltage from 3740 volts to 3910
volts.
Date of issuance: November 14, 2002.
Effective date: November 14, 2002, to be implemented within 30 days
from the date of issuance.
Amendment No.: 181.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53985).
The September 4, 2002, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 14, 2002.
No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: May 30, 2002, as supplemented on
September 13 and November 6 and 20, 2002.
Brief description of amendment: The amendment revises the Facility
Operating License and the Technical Specifications to increase the
licensed core thermal power level to 3067.4 megawatts (MWt), which is a
1.4% increase above the currently authorized power level of 3025 MWt.
The power uprate is based on the improvement in the core power
uncertainty allowance originally required for the emergency core
cooling system (ECCS) evaluations performed in accordance with Appendix
K, ``ECCS Evaluation Models,'' to part 50 of Title 10 of the CFR.
Specifically, the reduced uncertainty is obtained by using a more
accurate measurement of feedwater flow.
Date of issuance: November 26, 2002.
Effective date: November 26, 2002.
Amendment No.: 213.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45565).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 26, 2002.
The September 13, November 6, and November 20, 2002, letters
provided clarifying information that did not enlarge the scope of the
amendment request or change the initial proposed no significant hazards
consideration determination.
No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
County, Pennsylvania

Date of application for amendment: May 24, 2002, as supplemented by
letters dated June 27, September 11, September 24, and October 16,
2002.
Brief description of amendments: The amendments increase the
licensed power level by approximately 1.62% from 3458 megawatts thermal
(MWt) to 3514 MWt. These changes are based on increased feedwater flow
measurement accuracy achieved by utilizing high accuracy ultrasonic
flow measurement instrumentation.
Date of issuance: November 22, 2002.
Effective date: For Peach Bottom Atomic Power Station, Unit 2, as
of the date of issuance and shall be implemented within 60 days of
issuance. For Peach Bottom Atomic Power Station, Unit 3, as of its date
of issuance, and shall be implemented upon startup following the Unit 3
14th Refueling Outage, currently scheduled for fall 2003.
Amendment No: 247 and 250.
Facility Operating License Nos. DPR-44 and DPR-56: The amendment
revises the Technical Specifications and License.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45568).
The June 27, September 11, September 24, and October 16, 2002,
supplemental letters provided clarifying information that did not
change the scope of the original Federal Register notice or the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 22, 2002.
No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: June 12, 2000, as supplemented
by letters dated November 7, 2000, June 19

[[Page 75888]]

and August 17, 2001, January 15, June 5, and September 20, 2002.
Brief description of amendments: The amendments replace the current
accident source term used in design-basis radiological analyses for
control room habitability with an alternative source term (AST)
pursuant to Title 10 of the CFR part 50.67, ``Accident Source Term.''
The licensee for D.C. Cook, Units 1 and 2, Indiana Michigan Power
Company has requested a selective implementation of the AST limited to
control room habitability assessments. The licensee has elected to use
the AST and its associated acceptance criteria in preparing a revised
control room dose analysis to show compliance with 10 CFR Part 50,
Appendix A Criterion 19 ``Control Room.''
In addition, the proposed amendments revise the technical
specifications (TSs) to change the standard by which charcoal used in
engineered safeguard features systems is tested. The proposed changes
to the TSs are made in accordance with Generic Letter 99-02,
``Laboratory Testing of Nuclear-grade Activated Charcoal.'' The
amendments also revise the format of the TS pages to adopt the format
of Technical Specification Task Force (TSTF) Document TSTF-287
``Ventilation System Envelope Outage Time.''
Date of issuance: November 14, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 271 and 252.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51356).
The supplemental letters provided by the licensee contained
clarifying information and did not change the initial no significant
hazards consideration and did not expand the scope of the original
Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 14, 2002.
No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: January 14, 2002.
Brief description of amendments: The amendments would revise Unit 2
technical specification (TS) 3.4.2, ``Safety Valves--Shutdown,'' and TS
3.4.3, ``Safety Valves--Operating,'' to increase the allowable as-found
setpoint tolerance for the Unit 2 pressurizer code safety valves from
plus or minus (+/-) 1 percent (%) to +/-3%. In addition, the amendment
would add an allowable +/-1% as-left setpoint tolerance for the
pressurizer code safety valves to Unit 1 and Unit 2 TS 3.4.2 and TS
3.4.3.
Date of issuance: November 26, 2002.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 272 and 253.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15624).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2002.
No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska

Date of amendment request: July 22, 2002.
Brief description of amendment: The amendment removes from
Technical Specification (TS) 2.10.4(4)a and b, ``Azimuthal Power Tilt
(T),'' the reference to a specific computer program for monitoring core
radial peaking factors when a core power tilt is present. Instead, the
functional requirement is specified. This change clarifies the
requirements for core tilt monitoring associated with a computer system
upgrade and changes in computer programs. Also, a clarification is made
in the Bases section for TS 2.10.4 regarding the application of TS
2.10.4(1)(b) when the plant computer incore detector alarms for
monitoring core linear heat rate become inoperable.
Date of issuance: October 29, 2002.
Effective date: October 29, 2002, and shall be implemented within
120 days from the date of issuance.
Amendment No.: 211.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56326).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 2002.
No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin

Date of application for amendments: January 11, 2002.
Brief description of amendments: The amendments revise Technical
Specification 3.6.4, ``Containment Pressure,'' to reduce the maximum
allowable pressure from 3 pounds per square inch gauge (psig) to 2
psig.
Date of issuance: November 26, 2002.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 206 and 211.
Facility Operating License Nos. DPR-24 and DPR-27: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 19, 2002 (67 FR
12605).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2002.
No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California

Date of application for amendment: December 28, 2000, as
supplemented by letters dated March 29, 2001; October 31, 2001;
December 21, 2001; and October 18, 2002.
Brief description of amendment: The amendment replaces the current
technical specifications with a set of permanently defueled technical
specifications (PDTS) to reflect the permanently defueled condition of
the plant.
Date of issuance: November 18, 2002.
Effective date: November 18, 2002, and shall be implemented within
60 days of issuance, including the incorporation of the revised Quality
Assurance Program description that contains the relocated
administrative control requirements as described in the licensee's
March 29, October 31, and December 21, 2001 letters.
Amendment No.: 34.
Facility Operating License No. DPR-7: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 26, 2001 (66
FR 66471).
The December 21, 2001, and October 18, 2002, supplemental letters
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed,

[[Page 75889]]

and did not change the staff original no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 18, 2002.
No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of application for amendments: August 20, 2002.
Brief description of amendments: The amendments revised TS Table
3.3.6.1-1, ``Primary Containment Isolation Instrumentation,''
Functional Unit 5.a, Reactor Water Cleanup System Isolation, Main Steam
Valve Vault Area Temperature--High, to extend the frequency of the
channel calibration surveillance requirement from 122 days to 24
months, and revised applicable Bases.
Date of issuance: November 26, 2002.
Effective date: As of date of issuance and shall be implemented
within 60 days from the completion of Browns Ferry Units 2 and 3
refueling outages currently scheduled for early 2003, and the spring of
2004, respectively.
Amendment Nos.: 277 and 236.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63698).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 26, 2002.
No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee

Date of application for amendment: January 14, 2002.
Brief description of amendment: The amendment reduced the steady-
state specific activity of the primary coolant. The amendment also
changes the allowable value for the main control room air intake
radiation monitor made necessary by reducing the specific activity.
Date of issuance: November 18, 2002.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 41.
Facility Operating License No. NPF-90: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15629).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 18, 2002.
No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 2nd day of December 2002.

For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 02-30921 Filed 12-9-02; 8:45 am]

BILLING CODE 7590-01-P