[Federal Register Volume 67, Number 237 (Tuesday, December 10, 2002)]
[Notices]
[Pages 75867-75889]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-30921]


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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 
[email protected]. 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 [email protected]. 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 [email protected].

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.

    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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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.

    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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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.

    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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    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.

    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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.

    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.

[[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.

    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief : Anthony J. Mendiola.

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.

    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 E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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 
BoralTM 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 (BoralTM) 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.

    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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James Andersen, Acting.

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.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James Andersen, Acting.

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.

    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, PO Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James Andersen, Acting.

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.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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 [email protected].

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