[Federal Register Volume 67, Number 170 (Tuesday, September 3, 2002)]
[Notices]
[Pages 56317-56333]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-22197]


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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 Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 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, August 9, 2002, through August 22, 2002. 
The last biweekly notice was published on August 20, 2002 (67 FR 
53983).

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 October 3, 2002, 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

[[Page 56318]]

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 subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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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 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 requested involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the rquest 
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, 11555Rockville 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 
UnitedStates 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 AtomicSafety 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 AgencywideDocuments 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].

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, CalvertCliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: June 11, 2002.
    Description of amendments request: The proposed amendments revise 
the Unit 1 and 2 Technical Specification (TS) Administrative Controls 
Section to incorporate seven changes previously approved for the 
Improved Standard TechnicalSpecifications (ISTS). These changes are 
reflected in Revision 2 of NUREG-1432(Reference a). In addition, a 
change is also being requested to correct an inconsistency introduced 
in a prior TS amendment.
    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 not involve a significant increase in the probability 
or consequences of an accident previously evaluated.

[[Page 56319]]

    The majority of changes proposed are editorial in nature, that 
is, they do not change the fundamental requirement of the 
TechnicalSpecification. They generally clarify the existing 
requirement. The remaining changes are changes to the Technical 
Specification requirements. The deletion of the pressurizer safety 
and relief valve challenges and failures report does not impact the 
operation of the pressurizer safety and relief valves and still 
permits reporting of significant failures under the provisions of 10 
CFR 50.72 and 50.73. Removal of pipe supports from the Inservice 
Testing Program description corrects the description of the program. 
It does not change the manner or timing of any evaluations of pipe 
supports or snubbers. Removal of the discussion of the Nuclear 
Regulatory Commission environmental monitoring program with the 
state reflects the cancellation of that program with the state. It 
does not alter any other environmental monitoring requirements.
    As described above, these proposed changes are generally 
editorial in nature or have no impact on plant operation. None of 
the proposed changes impact the operation of any equipment needed 
for the mitigation of an accident or any known accident initiators.
    Therefore, the probability or consequences of an accident 
previously evaluated have not significantly increased.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    As noted above, these changes are generally editorial in nature. 
That is, they do not change the fundamental requirement of the 
TechnicalSpecification. They generally clarify the existing 
requirement. The remaining changes do not impact plant operation. 
None of the proposed changes would result in new or different plant 
operation or the addition of new equipment.
    Therefore, the possibility of a new or different [kind] of 
accident from any previously evaluated is not created.
    3. Would not involve a significant reduction in a margin of 
safety.
    Since the majority of the proposed changes are editorial in 
nature, they do not change the fundamental Technical Specification 
requirement. Therefore, they do not impact the margin of safety 
represented by these Technical Specifications. The remaining changes 
do not impact plant operation and generally align these Technical 
Specification requirements with the criteria given in 10 CFR 
50.36(c)(2)(ii). The deletion of the pressurizer safety and the 
relief valve challenges and failures report does not impact the 
operation of the pressurizer safety and relief valves and still 
permits reporting of significant failures under the provision of 10 
CFR 50.72 and 50.73. Removal of pipe supports from the Inservice 
Testing Program description corrects the description of the program. 
It does not change the manner or timing of any evaluations of pipe 
supports or snubbers. Removal of the discussion of the Nuclear 
Regulatory Commission environmental monitoring program with the 
state reflects the cancellation of that program with the state. It 
does not alter any other environmental monitoring requirements. 
These changes do not impact the margin of safety.

    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company (CP&L), Docket No. 50-261, H. B. 
Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, 
South Carolina

    Date of amendment request: May 10, 2002, as supplemented August 12, 
2002.
    Description of amendment request: The proposed amendment would 
allow an increase in the authorized reactor power level for HBRSEP2.
    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:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change * * * does not involve a significant 
increase in the probability of an accident previously evaluated 
based on the results of comprehensive analytical efforts that were 
performed to demonstrate the acceptability of the proposed power 
uprate changes.
    An evaluation has been performed that identified the systems and 
components that could be affected by these proposed changes. The 
evaluation determined that these systems and components will 
function as designed and that performance requirements remain 
acceptable.
    The primary loop components (reactor vessel, reactor internals, 
control rod drive mechanisms (CRDMs), loop piping and supports, 
reactor coolant pumps, steam generators and pressurizer) will 
continue to comply with their applicable structural limits and will 
continue to perform their intended design functions. Thus, there is 
no increase in the probability of a structural failure of these 
components leading to an accident.
    The Leak-Before-Break analysis conclusions remain valid and the 
breaks previously exempted from structural considerations remain 
unchanged.
    Systems included within the scope of the Nuclear Steam Supply 
System(NSSS) will continue to perform their intended design 
functions during normal and accident conditions. Additionally, NSSS 
components will continue to comply with applicable structural limits 
and will continue to perform their intended design functions. Thus, 
there is no increase in the probability of a structural failure of 
these components.
    The NSSS/Balance of Plant interface systems will continue to 
perform their intended design functions. The MSSVs [main steam 
safety valves] will provide adequate relief capacity to maintain the 
Main Steam System within design limits. The maximum feedwater flow 
rate and the isolation time for the MFRVs [main feedwater regulating 
valves] andBypass Valves will continue to ensure that the analyzed 
containment pressure during postulated accidents remains below the 
allowable limit.
    The current loss-of-coolant [accident] (LOCA) hydraulic analyses 
remain bounding.
    The reduction in power measurement uncertainty achieved through 
the use of the Caldon Leading Edge Flow Meter (LEFM) Check-Plus 
TM system allows for certain safety analyses to continue 
to be used, without modification, at the 2346 MWt [megawatt thermal] 
power level (102 percent of 2300 MWt). Other safety analyses 
performed at a nominal power level of 2300 MWt have been either re-
performed or re-evaluated to support the 2339 MWt power level, and 
continue to meet their applicable acceptance criteria. Some existing 
safety analyses had been previously performed at a power level 
greater than or equal to 2346 MWt, and thus continue to bound the 
2339 MWt power level.
    The proposed changes to the RCS [reactor coolant system] 
pressure-temperature limit curves impose a conservative projection 
of the increase in neutron fluence associated with the power uprate. 
This projection will ensure that the requirements of 10 CFR 50, 
Appendix G, ``Fracture Toughness Requirements,'' will continue to be 
met following the proposed power uprate. The design basis events 
that were protected against by these limits have not changed, 
therefore, the probability of an accident previously evaluated is 
not increased.
    Based on the foregoing, it is concluded that this 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 previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because no new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed power uprate 
changes. Systems, structures, and components previously required for 
the mitigation of an event remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component, and do not 
challenge the

[[Page 56320]]

performance or integrity of any safety-related system.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Extensive analyses of the primary fission product barriers 
conducted in support of the proposed power uprate have concluded 
that relevant design criteria remain satisfied, both from the 
standpoint of the integrity of the primary fission product barrier 
and compliance with regulatory acceptance criteria. As appropriate, 
evaluations have been performed using methods that have either been 
reviewed and approved by the Nuclear Regulatory Commission (NRC), or 
that are in compliance with applicable regulatory review guidance 
and standards.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based on the above discussion, CP&L has determined that 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Kahtan N. Jabbour, Acting.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: August 8, 2002.
    Description of amendment request: The proposed amendment allows a 
revision of the current reactor pressure vessel (RPV) material 
surveillance program description in the Updated Final Safety Analysis 
Report for Fermi 2 to reference the Integrated Surveillance Program 
(ISP) that was developed by the Boiling Water Reactor Owners Group's 
Vessel and Internals Project (BWRVIP). The proposed amendment is 
consistent with the NRC's Regulatory Issue Summary 2002-05, ``NRC 
Approval of Boiling Water Reactor Pressure Vessel Integrated 
Surveillance Program,'' datedApril 8, 2002 (ADAMS Accession No. 
ML020660522).
    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 License Amendment involves a change in the program 
of RPV material surveillance for monitoring the effects of neutron 
embrittlement and thermal environment as required by Appendix H of 
10CFR 50. Instead of the Fermi 2 plant-specific program, the BWRVIP 
ISP is proposed for use in complying with the requirements of 
Appendix H[to 10 CFR Part 50]. Paragraph III.C of Appendix H 
provides the requirements for an ISP. The BWRVIP ISP has been 
reviewed and approved by the NRC staff as an acceptable program for 
use by all BWRs. There are many advantages for participating in the 
ISP over utilizing a plant-specific program. The advantages include 
improved compliance with the NRC requirements, better matching of 
the plant limiting material to the representative capsule material, 
additional data points for irradiated and unirradiated specimens, 
and better quality and consistency of the data and methodology. 
Additionally, future calculations of neutron fluence will be 
completed in accordance with the approved NRC methodologies in 
Regulatory Guide (RG) 1.190[``Calculational and Dosimetry Methods 
for Determining Pressure VesselNeutron Fluence''].
    The data obtained from testing the RPV surveillance capsules is 
used to define the pressure-temperature limits for the RPV and to 
ensure that fracture toughness requirements for ferritic materials 
of pressure retaining components of the reactor coolant boundary are 
met. Using the ISP for RPV material surveillance program enhances 
the RPV integrity evaluations and results in using data from better-
matching specimens. The ISP also results in better compliance with 
the NRC requirements and consistency among the BWR plants.
    The proposed change results in better compliance with the 
regulatory requirements for RPV material surveillance; therefore, it 
does not increase the likelihood of a malfunction of plant 
structures, systems and components.
    Based on the above, the proposed change does not significantly 
increase the probability or consequences of any 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 purpose of the RPV material surveillance program is to 
monitor neutron embrittlement and thermal environment effects in 
order to predict the behavioral characteristics of ferritic material 
of pressure retaining components of the reactor coolant pressure 
boundary and to ensure RPV fracture toughness and integrity 
requirements are not violated. The BWRVIP ISP was approved for use 
by all BWRs as an alternate to plant-specific programs. The change 
does not affect the design function or operation of any plant 
structure, system or component. The ISP is an approved alternate 
monitoring program that meets the regulatory requirements in 
Appendix H to 10 CFR 50. As an alternate monitoring program, the ISP 
cannot create a new failure mode involving the possibility of a new 
or different kind of accident. Therefore, the proposed change does 
not create the potential for a new or different kind of accident 
from any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The RPV material surveillance program requirements in Appendix H 
to 10 CFR 50 are designed to provide adequate margins of safety 
during any condition of normal operation, including anticipated 
operational occurrences and system hydrostatic tests, to which the 
reactor coolant pressure boundary may be subjected over its service 
lifetime. The material surveillance data for the Fermi 2 RPV 
obtained through the ISP is equal or better to that from plant-
specific programs. Paragraph III.C of Appendix H to 10 CFR 50 
delineates the regulatory requirements for an ISP. The BWRVIP ISP 
meets these requirements and has been approved by the NRC. 
Therefore, the proposed changes will 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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: July 29, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Surveillance Requirement 3.7.2.2 to 
decrease the allowable closure time for the turbine stop valves from 15 
seconds to 1 second.
    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 
HazardsConsideration 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:

[[Page 56321]]

    (1) Involve a significant increase in the probability or 
consequences of an Accident previously evaluated:
    No. The request is for a decrease in the Turbine Stop Valve 
(TSV) closure time acceptance criteria of Technical Specification 
(TS) Surveillance Requirement (SR) 3.7.2.2, from a value of [le]15 
seconds to a value of [le]1 second. This decrease in the closure 
time for the Channel B closure circuitry is more conservative and is 
being made to match the existing 1 second or less acceptance 
criteria of the closure time of the Channel A closure circuitry. The 
new Chapter 15 Transient Analysis Methodology assumes that the TSVs 
will be closed in 1 second or less by either the Channel A or 
Channel B closure circuitry. The new design has already been 
installed and tested, and is more conservative than the previous 
design. Therefore, the request for a more restrictive TS SR does not 
involve a significant 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. The 1 second or less closure time was and is acceptable 
under the existing TS SR for the Channel B circuitry since the 
existing acceptance is 15 seconds or less. This request is to change 
the TS SR and its Bases to a more restrictive requirement (1 second 
or less). This more restrictive requirement is being requested to 
ensure that the installed equipment will continue to meet the 
conditions and assumptions that are currently in the analysis model 
described in the Topical Report DPC-NE-3005-P, ``UFSAR Chapter 15 
Transient Analysis Methodology''. Therefore, this request 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, setpoints, 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 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: 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: July 18, 2002.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to 
perform a missed surveillance. The time is 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 July 18, 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 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.

[[Page 56322]]

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

    Date of amendment request: July 5, 2002.
    Description of amendment request: The proposed amendment would 
increase the licensed power level by 1.5% from 1,998 megawatts thermal 
(MWt) to 2,028 MWt based on the installation of ultrasonic flow 
measurement instrumentation resulting in improved feedwater flow 
measurement accuracy. The proposed amendment would change the Operating 
License and Technical Specifications to reflect the increase in 
licensed power level.
    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 standards of 10 CFR 50.92(c). The NRC staff's review 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 increase in power level is achieved by improving 
the accuracy of the feedwater flow measurement instrumentation 
resulting in a more accurate feedwater flow used in the heat balance 
calculation. The increased flow accuracy improves the uncertainty in 
the core power level from the existing 2% margin to [le]0.5%. The 
probability of an accident previously evaluated is not increased by 
the proposed change because the flow measurement instrumentation is 
not an initiator of design-basis accidents (DBAs) evaluated in the 
updated final safety analysis report (UFSAR). The consequences due 
to postulated DBA events previously evaluated are based on analyses 
using a 2% margin above the current licensed power level which 
bounds the proposed 1.5% power level increase. 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 increase in power level is achieved by improving 
the accuracy of the feedwater flow measurement instrumentation. 
Using the more accurate flow measurement in the heat balance 
calculation improves the core power level uncertainty. The proposed 
increase in power level will not create a change in the operation or 
function of the flow measurement instrumentation. Changes to the 
feedwater flow measurement accuracy does not create accident 
initiators not considered in the DBAs. Therefore, the proposed 
change 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 calculated loads on all affected structures, systems, and 
components have been shown to remain within design criteria at the 
increased power level for all design-basis event categories. The 
current design margins, operational margins, and margins of safety 
are not exceeded by the increased power level. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Jacob I. Zimmerman, Acting.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: July 26, 2002.
    Description of amendment request: The proposed amendment would 
allow relocation of the Kewaunee Nuclear Power Plant (KNPP) cycle 
dependent variables from the Technical Specifications (TS) to a formal 
report, Core Operating Limits Report(COLR).
    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. Operation of the Kewaunee Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The proposed changes relocate certain cycle specific parameters 
from the Technical Specifications to a Core Operating Limits Report 
(COLR) or are administrative in nature. Appropriate design and 
safety limits are retained or added to the Specifications thereby 
meeting the requirements of 10 CFR 50.36. Specific, approved 
methodologies used to determine and evaluate the parameter 
requirements are added to the Specifications and a reporting 
requirement is added to ensure the NRC is apprised of all changes. 
Approved methodologies are required to be used to evaluate and 
change parameters, and appropriate safety and design limits are 
maintained in the Technical Specifications. Thus, operation of KNPP 
will continue to meet all design and safety analysis requirements. 
Therefore, neither the probability nor consequences of an accident 
previously evaluated can be increased.
    2. Operation of the Kewaunee Nuclear Plant in accordance with 
the proposed amendment does not create a new or different kind of 
accident from any accident previously evaluated.
    Operation of KNPP, in accordance with the proposed changes, will 
continue to meet all design and safety limits. Appropriate design 
and safety limits continue to be controlled within the Technical 
Specifications. These changes will not result in a change to the 
design and safety limits under which KNPP operation has been 
determined to be acceptable. Therefore, these changes cannot result 
in a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the Kewaunee Nuclear Plant in accordance with 
the proposed amendment does not result in a significant reduction in 
a margin of safety.
    Appropriate safety limits continue to be controlled by the 
Technical Specifications. Changes to cycle specific parameters 
related to these limits will be accomplished using NRC approved 
methodologies, thereby ensuring operation will continue within the 
bounds of the existing safety analyses including all applicable 
margins of safety. Therefore, operation in accordance with the 
proposed changes cannot 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 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, (NMC) Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: July 26, 2002.
    Description of amendment request: The proposed license amendment 
would implement changes to the Kewaunee Nuclear Power Plant (KNPP) 
Technical Specifications (TS) to accommodate Westinghouse 422 VANTAGE + 
nuclear fuel with PERFORMANCE + features.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The [Nuclear Regulatory Commission] NRC generically approved 
Westinghouse 422V+

[[Page 56323]]

[Westinghouse 422 VANTAGE + nuclear fuel with PERFORMANCE + 
features] fuel assemblies for use in reactors substantially similar 
to KNPP. NMC used 422V+ fuel in the Lead-Test-Assembly Program 
during cycle 25, as permitted by existing TS. Empirical data 
acquired during Cycle 25 confirms that this fuel is both compatible 
with KNPP reactor design and with the Framatome/ANP fuel currently 
in use. Reanalysis of postulatedKNPP design basis accidents shows 
that reactor operation with 422V+ fuel remains within design basis 
limitations and safety margins. All design basis accidents and 
transients affected by the fuel upgrade were analyzed, and the 
results documented in the Westinghouse Report provided with this 
request. These analyses and evaluations show that use of 422V+ fuel 
is acceptable. The margin to safety is not exceeded in any instance. 
Pending approval of Addendum 2 to [Westinghouse CommercialAtomic 
Power ``Revision to Design Criteria''] WCAP 12488 revising the 
current transient stress strain criteria, all design basis 
acceptance criteria will be satisfied. Changes to the technical 
specification that remain within the limits of the bounding accident 
analyses cannot change the probability or consequence of an accident 
previously evaluated. Thus, nothing in this proposal will cause an 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Use of the 422V+ fuel is consistent with current plant design 
bases and does not adversely affect any fission product barrier, nor 
does it alter the safety function of safety significant systems, 
structures and components or their roles in accident prevention or 
mitigation. The operational characteristics of 422V+ fuel are 
bounded by the safety analyses (Attachment 4 [of the submittal]). 
The 422V+ fuel design performs within existing fuel design limits. 
Thus, this proposal does not create the possibility of a new or 
different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. Licensed safety margins are maintained. It 
conforms to plant design bases, is consistent with current safety 
analyses, and limits actual plant operation within analyzed and 
licensed boundaries. Analyses of design basis accidents and 
transients were performed using power level greater than that 
currently licensed, thus rendering more conservative results than 
required. All safety analysis acceptance criteria are satisfied at 
this value and all KNPP safety requirements continue to be met. Use 
of 422V+ fuel as proposed by this amendment request is bounded by 
these analyses. Thus, changes proposed by this request 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: April 22, 2002.
    Description of amendment request: The proposed amendment would 
revise the reactor vessel pressure and temperature (P/T) limit curves 
in the Monticello Technical Specifications (TSs). The revised P/T 
limits will allow required hydrostatic and leak tests to be performed 
at a significantly lower temperature. This is expected to reduce 
challenges to plant operators associated with maintaining the reactor 
coolant system within a narrow temperature band during testing.
    The Nuclear Management Company, LLC, is also requesting an 
exemption from the requirements of 10 CFR Part 50, Appendix G, to allow 
the use of American Society of Mechanical Engineers (ASME) Boiler and 
Pressure Vessel Code (Code) Case N-640 as the basis for these revised 
curves. The proposed P/T curves were developed in accordance with the 
1989 edition of the ASME Code, Section XI, Appendix G; 10 CFR Part 50, 
Appendix G; and ASME Code Case N-640. The use of this Code Case as the 
basis for the proposed P/T curves constitutes an alternative to the 
requirements of 10 CFR Part 50, Appendix G. The regulation at 10 CFR 
50.60(b) provides that the NRC may grant alternatives to the 
requirements in Appendix G by using the procedures for exemption 
specified in 10 CFR 50.12.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The P/T limits are not derived from Design Basis Accident (DBA) 
analyses. They are prescribed by the ASME Code and 10 CFR 50 
Appendix G and H as restrictions on operation to avoid encountering 
pressure, temperature, and temperature rate of change conditions 
that might cause undetected flaws to propagate and cause non-ductile 
failure of the reactor coolant pressure boundary.
    The changes to the calculation methodology for the P/T limits 
are based upon ASME Code Case N-640, ``Alternative Reference 
Fracture Toughness for Development of P-T Limit Curves for ASME 
Section XI, division 1,'' and provide adequate margin in the 
prevention of a non-ductile type fracture of the reactor pressure 
vessel (RPV). The code case was developed based upon the knowledge 
gained through years of industry experience. The P/T limits 
developed using the allowances of ASME Code Case N-640 provide more 
operating margin. However, experience gained in the areas of 
fracture toughness of materials and pre-existing undetected defects 
shows that some of the existing assumptions used for the calculation 
of P/T limits are unnecessarily conservative and unrealistic. 
Therefore, use of the allowances of ASME Code Case N-640 in 
developing the P/T limits will provide adequate protection against 
nonductile-type fractures of the RPV.
    Development of the revised Monticello P/T limits was performed 
using the approved methodologies of 10 CFR 50, Appendix G, and using 
the allowances of ASME Code Case N-640. The P/T limit curves 
generated using these methods ensure the P/T limits will not be 
exceeded during any phase of reactor operation. Therefore, the 
probability of occurrence and the consequences of a previously 
analyzed event are not significantly increased. Finally, the 
proposed change will not affect any other system or piece of 
equipment designed for the prevention or mitigation of previously 
analyzed events.
    Thus, the probability of occurrence and the consequences of any 
previously analyzed event are not significantly increased as the 
result of the proposed changes.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes provide more operating margin in the P/T 
limit curves for inservice leakage and hydrostatic pressure testing, 
non-nuclear heatup and cooldown, and criticality, with benefits 
being primarily realized during the pressure tests. Operation in the 
``new'' regions of the newly developed P/T curves has been analyzed 
in accordance with the provisions of ASME Code, Section XI, Appendix 
G; 10 CFR 50 Appendix G, and ASME Code Case N-640, thus providing 
adequate protection against a nonductile-type fracture of the RPV.
    The proposed changes do not alter any existing system 
relationships. The proposed changes do not result in any new or 
unanalyzed operation of any system or piece of equipment important 
to safety, and as a result, the possibility of a new type [of] event 
is not created.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    As mentioned previously, the revised P/T limit curves provide 
more operating margin and thus, more operational flexibility than

[[Page 56324]]

the current P/T limit curves. With the increased operational margin, 
a reduction in the safety margin results with respect to the 
existing curves. However, industry experience since the inception of 
the P/T limits in 1974 confirms that some of the existing 
methodologies used to develop P/T limit curves are unrealistic and 
unnecessarily conservative. Accordingly, ASME Code Case N-640 takes 
into account the acquired knowledge and establishes more realistic 
methodologies for the development of P/T limit curves.
    Use of ASME Code Case N-640 to develop the revised P/T curves 
utilized the KIC fracture toughness curve in lieu of the 
KIA curve as the lower bound for fracture toughness. Use 
of the KIC curve to determine lower bound fracture 
toughness is more technically correct than using the KIA 
curve. P/T curves based on the KIC fracture toughness 
limits enhance overall plant safety by expanding the P/T window in 
the low-temperature operating region. The benefits which occur are a 
reduction in the duration of the pressure test and personnel safety 
while conducting inspections in primary containment with no decrease 
to the margin of safety. Therefore, operational flexibility is 
gained and an acceptable margin of safety to RPV non-ductile type 
fracture is maintained.
    Therefore, the proposed amendment will 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N. Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: April 22, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to permit a one-time 5-year 
extension, to no later than March 2008, of the 10-year performance-
based Type A test interval established in NEI 94-01, ``Nuclear Energy 
Institute Industry Guideline for Implementing Performance-Based Option 
of 10 CFR Part 50, Appendix J,'' Revision 0, dated July 26, 1995.
    This TS change has been prepared in accordance with the guidance 
provided in Regulatory Guide 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant 
Specific Changes to the Licensing Basis.''
    A plant-specific, risk-based evaluation has been performed in 
support of this one-time exception to extend the Type A test interval. 
This evaluation uses the latest Monticello probabilistic safety 
assessment (PSA) models to estimate the changes in risk associated with 
increasing the Type A testing interval. This risk assessment is 
consistent with current PSA best practices.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to TS 4.7.A.2.b provides a one-time 
exception to the testing frequency for the Type A containment 
integrated leakage rate test. The current ten-year interval is based 
on past performance and the proposed change will only extend the 
Type A test frequency to fifteen years. The proposed change to the 
Technical Specifications does not involve a physical change to the 
plant or a change in the manner in which the plant is operated or 
controlled. The primary containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the primary containment does not involve the prevention or 
identification of any precursors of an accident and therefore does 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The consequences of the evaluated accidents are the amount of 
radioactivity that is released to secondary containment and 
subsequently to the public. The proposed change involves a one-time 
change to the interval between Type A containment leakage tests. 
Type B and C containment leakage tests will continue to be performed 
at the frequency specified in the Monticello Technical 
Specifications. As documented in NUREG-1493, ``Performance-Based 
Containment Leakage-Test Program,'' industry experience has shown 
that Type B and C containment leakage tests have identified a very 
large percentage of containment leakage paths and that the 
percentage of containment paths that are detected only by Type A 
tests is very small. An analysis of 144 integrated leak rate tests, 
including 23 failures, found that no failures were due to 
containment liner breach. NUREG-1493 also concluded, in part, that 
reducing the frequency of Type A containment leakage rate tests to 
once per twenty years was found to lead to an imperceptible increase 
in risk. The Monticello risk-based evaluation of the proposed one-
time extension to the Type A test frequency supports this 
conclusion. The integrity of the reactor containment is subject to 
two types of failure mechanisms which can be categorized as (1) 
activity based and (2) time based. Activity based failure mechanisms 
are defined as degradation due to system and/or component 
modifications or maintenance. Local leak rate test requirements and 
administrative controls such as design change control and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the primary 
containment, combined with the containment inspections performed in 
accordance with the American Society of Mechanical Engineers (ASME) 
Code, Section XI and 10 CFR 50.65, Maintenance Rule, provide a high 
degree of assurance that the primary containment will not degrade in 
a manner that is detectable only by Type A tests and therefore does 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed change to Technical Specification 4.7.A.2.b 
involves a one-time exception to the current test interval for Type 
A containment leakage rate tests. The primary containment and the 
test requirements invoked to periodically demonstrate the integrity 
of the primary containment exist to ensure the ability to mitigate 
the consequences of an accident. Additionally, the reactor 
containment and its associated test requirements do not involve the 
prevention or identification of any precursors of an accident. The 
proposed change to the leakage rate test frequency does not involve 
any physical changes being made to the facility. In addition, the 
proposed extension of the Type A leakage rate test frequency does 
not change the operation of the plant such that a new failure mode 
involving the possibility of a new or different kind of accident 
from any accident previously evaluated is created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed Technical Specification change does not involve a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The proposed change involves only 
the extension of the interval between Type A containment leakage 
tests. The current interval of ten years, based on past performance, 
would be extended on a one-time basis to fifteen years from the last 
Type A test. Type B and C containment leakage tests will continue to 
be performed at the frequency currently required by the plant 
Technical Specifications.
    The NUREG-1493 generic study of the effects of extending 
containment leakage test intervals found that a twenty-year 
extension

[[Page 56325]]

for Type A leakage tests resulted in an imperceptible increase in 
risk to the public. This study also found that, generically, the 
containment leakage paths are mainly detected by Type B and C tests. 
The proposed change involves a one-time extension of the frequency 
for Type A containment leakage tests; the overall primary 
containment leakage rate limit, specified by the Monticello 
Technical Specifications, is being maintained. The regular 
containment inspections being performed in accordance with the ASME 
Code, Section XI, and 10 CFR 50.65, Maintenance Rule, provide a high 
degree of assurance that the containment will not degrade in a 
manner that is only detectable by Type A tests. In addition, the 
containment monitoring capability that is inherent to boiling water 
reactors using an inert containment atmosphere allows for the 
detection of gross containment leakage that may develop during power 
operation. The cumulative effect of these inspections, tests and 
operating methods ensures that the margin of safety is maintained.
    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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: April 25, 2002.
    Description of amendment request: The proposed amendment would 
revise the Monticello Technical Specifications (TSs) to allow the use 
of 10 CFR Part 50, Appendix J, Option B, for Types B and C containment 
leak rate testing. The proposed amendment would also revise the 
surveillance requirements (SRs) in TS 3.7/4.7 and provide a new TS 
Section 6.8.M, ``Primary Containment Leakage Rate Testing Program,'' in 
the ``Programs and Manuals'' section of the Monticello TSs. This 
proposed new TS program is formatted to be consistent with the NRC-
approved guidance provided in Option B of the Primary Containment 
Leakage Rate TestingProgram included in NUREG-1433, ``Standard 
Technical Specifications General Electric Plants, BWR/4,'' Revision 2, 
dated April 2001.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes deal exclusively with testing of features 
related to containment isolation. The changes only affect testing 
frequency and methodology. Containment leakage is not considered as 
an initiator of any accident previously evaluated.
    Additionally, the proposed changes do not impact current plant 
operations or the design function of any system or component. The 
proposed changes do not change any accidents previously evaluated in 
the updated safety analysis report.
    The proposed changes only affect the frequency of testing the 
containment penetrations and containment isolation valves. The 
proposed changes will allow test intervals to be extended in 
accordance with program requirements and 10 CFR Part 50, Appendix J, 
Option B, with reference to Regulatory Guide 1.163, and NEI 94-01, 
Rev. 0. The change in risk resulting from the proposed change, was 
evaluated by the NRC in the rule making process for implementing the 
Option B requirements, and are characterized in NUREG-1493. For Type 
B and C tests, the NRC concluded that the extension of test 
intervals as allowed by Option B would lead to only minor increases 
in potential offsite dose consequences.
    The performance of the leakage tests themselves is not an input 
or consideration in any accident previously evaluated, thus the 
proposed change will not increase the probability of any such 
accident occurring. The same operability requirements remain in 
place for the primary containment, therefore, the consequences of an 
accident are not significantly increased. The proposed revision does 
not involve any change to the configuration or method of operation 
of any plant equipment that is used to mitigate the consequences of 
an accident, nor does it affect any assumptions or conditions in the 
accident analysis.
    Therefore, operation of the facility in accordance with the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed changes deal exclusively with testing of features 
related to containment isolation. The changes only affect testing 
frequency and methodology. The proposed changes to the TS will not 
result in any physical alterations to the plant configuration, no 
new equipment is added, no equipment interfaces are modified, and no 
changes to any equipment's function or the method of operating the 
equipment are being made. Since the proposed changes would not 
change the design, configuration or operation of the plant, they 
would not cause the containment leak rate testing to become an 
accident initiator. No new or different kinds of accident modes are 
created.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed changes deal exclusively with testing of features 
of [sic] related to containment isolation. The changes only affect 
testing frequency and methodology. Containment leakage is not 
considered as an initiator of any accident previously evaluated.
    The proposed changes do not exceed or alter a design basis or 
safety limit. The proposed changes only affect the methodology and 
frequency of Type B and C testing. The proposed performance based 
approach, provided by using Option B to 10 CFR Part 50, Appendix J, 
would continue to ensure that the containment leakage rates would 
not exceed the maximum allowable leakage rates defined in the 
Technical Specifications and assumed in the accident analysis.
    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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N. Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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 amendment request: June 11, 2002.
    Description of amendment request: The proposed amendment would 
revise TS 3.1.8, ``Physics Test Exceptions,'' to correct a 
typographical error in the numbering of a function. The existing 
typographical error inappropriately makes the TS more restrictive than 
intended.
    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. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant

[[Page 56326]]

increase in the probability or consequences of any accident 
previously evaluated.
    The primary purpose of the Mode 2 Physics Tests exceptions is to 
permit relaxations of existing LCOs [limiting conditions for 
operation] to allow certain Physics Tests to be performed. The 
proposed change will permit the number of required channels 
specified in LCO 3.3.1, ``RPS [ReactorProtection System] 
Instrumentation,'' for Power Range Neutron Flux, P-10 interlock, to 
be reduced to ``3'' required channels for Physics Tests, as 
originally analyzed and approved by NRC. LCO 3.1.8 already allows 
one power range neutron flux channel to be bypassed, reducing the 
number of required channels from ``4'' to ``3''. With this reduction 
in the number of required channels, the fuel design criteria are 
preserved as long as the power level is limited to [le]5% RTP [rated 
thermal power], the reactor coolant temperature is kept 
[ge]530 deg.F, and shutdown margin (SDM) is within the limits 
provided in the Core Operating Limits Report (COLR). These three 
conditions are not affected by the proposed change. This change only 
restores the allowance previously analyzed as acceptable.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be significantly increased as a result 
of the proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The proposed change does not involve any physical alteration of 
plant systems, structures or components, nor does it alter 
parameters governing normal plant operation. This change does not 
introduce any new or different normal operation or accident 
initiators. With the reduction in the number of required 
instrumentation channels, the fuel design criteria continue to be 
preserved as originally analyzed.
    Equipment important to safety will continue to operate as 
designed. The changes do not result in any event previously deemed 
incredible being made credible. The changes do not result in more 
adverse conditions or result in any increase in the challenges to 
safety systems. Therefore, operation of the Point Beach Nuclear 
Plant in accordance with the proposed amendment will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The primary purpose of the Mode 2 Physics Tests exceptions is to 
permit relaxations of existing LCOs to allow certain Physics Tests 
to be performed. The analysis for Physics Tests is based on one 
power range neutron flux channel being bypassed. Therefore, reducing 
the requirement for an interlock associated with the bypassed 
channel is bounded by the original analysis. There are no new or 
significant changes to the initial conditions contributing to 
accident severity or consequences. The proposed amendment will not 
otherwise affect the plant protective boundaries, will not cause a 
release of fission products to the public, nor will it degrade the 
performance of any other structures, systems or components important 
to safety. Therefore, the proposed change will 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N. Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: July 22, 2002.
    Description of amendment request: The proposed amendment removes 
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. These changes clarify the 
requirements for core tilt monitoring associated with a computer system 
upgrade and changes in computer programs. Also, it is proposed to add 
clarification in the Basis section for Technical Specification (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.
    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 change does not result in any changes to the existing 
core power distribution monitoring requirements. There is no change 
in the analysis values used in the evaluation of the transients and 
accidents. All of the evaluated transients and accidents currently 
show acceptable results and will not be affected by this change. 
Incorporating this change will not affect the probability of an 
accident, since core power distribution monitoring is not changed. 
The change to the wording of the core power distribution monitoring 
specifications will not change the failure possibilities for reactor 
protective features. The effect of the proposed change is the 
clarification of the existing core power distribution monitoring 
requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change to the wording of the core power distribution 
monitoring specifications does not provide the possibility of the 
creation of a new or different type of accident. Changing the 
wording of the core power distribution monitoring specifications 
does not change the method of core power distribution monitoring or 
the expected response of reactor protective features. The reactor 
will operate within previously analyzed limits.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to the wording of the core power 
distribution monitoring specifications does not constitute a 
significant reduction in the margin of safety due to the core power 
distribution monitoring requirements are not changed and are 
consistent with the assumptions contained in the transient and 
accident analyses contained in the UpdatedSafety Analysis Report 
shown to produce acceptable results.
    The acceptance criteria used in the analysis have been developed 
for the purpose of use in design basis accident analyses such that 
meeting these limits demonstrates adequate protection of public 
health and safety. An acceptable margin of safety is inherent in 
these licensing limits. Therefore, the proposed changes do not 
involve a 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 22, 2002.
    Description of amendment request: The proposed amendment deletes 
technical specification (TS) requirements for missed surveillances from 
TS 3.0.4 and adds TS 3.0.5 for missed surveillances consistent with the 
Improved Standard Technical Specification (ITS) for Combustion 
Engineering Plants, NUREG-1432, Revision 2, and Technical Specification 
Task Force Change Traveler TSTF-358, Revision 6. This proposed 
amendment also adds a TS requirement for a Bases Control Program 
consistent with that presented in Section 5.5 of the ITS (NUREG-1432, 
Revision 2), in

[[Page 56327]]

accordance with the guidance published in the Federal Register on 
September 28, 2001, ``Notice of Availability of Model Application 
Concerning Technical Specification Improvement to Modify Requirements 
Regarding Missed Surveillances Using the Consolidated Line Item 
Improvement Process,'' (66 FR 49714). Appropriate TS Bases changes are 
also provided in accordance with the Consolidated Line Item Improvement 
Process.
    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 to incorporate Improved Standard Technical 
Specification (ITS) SR 3.0.3 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. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to provide a Technical Specification (TS) 
Bases Control Program presents more stringent requirements than 
previously existed in the Technical Specifications. These more 
stringent requirements do not result in operation that will increase 
the probability of initiating an analyzed event. If anything the new 
requirements may decrease the probability or consequences of an 
analyzed event by incorporating the more restrictive changes. The 
changes do not alter assumptions relative to mitigation of an 
accident or transient event. The more restrictive requirements 
continue to ensure process variables, structures, systems, and 
components are maintained consistent with the safety analyses and 
licensing basis. Therefore, the changes do 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 to incorporate ITS SR 3.0.3 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. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change to provide a TS Bases Control Program 
presents more stringent requirements than previously existed in the 
TechnicalSpecifications. The changes do not alter the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The changes do impose different requirements. However, 
these changes are consistent with the assumptions in the safety 
analyses and licensing basis. Therefore, the changes do 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 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 inoperable Limiting Condition 
for Operation 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. 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 proposed change to provide a TS Bases Control Program 
presents more stringent requirements than previously existed in the 
TechnicalSpecifications. Adding more restrictive requirements either 
increases or has no impact on the margin of safety. The changes, by 
definition, provide additional restrictions to enhance plant safety. 
The changes maintain requirements within the safety analyses and 
licensing basis. As such, no question of safety is involved. 
Therefore, the 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 23, 2002.
    Description of amendment request: This proposed amendment will: (1) 
Remove the requirement to demonstrate operability of redundant 
auxiliary feedwater system components, and (2) provide an allowed 
outage time to restore operability of the emergency feedwater storage 
tank. Each of the revisions is modeled after the Improved Standard 
Technical Specifications.
    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 changes to Technical Specifications Sections 2.5 
establish an allowed outage time and actions required for restoring 
operability.The proposed Technical Specifications address the 
regulatory requirements for equipment required for Auxiliary 
Feedwater Systems per NUREG-0635[``NRC Requirements for Auxiliary 
Feedwater Systems'']. The change will ensure that proper Limiting 
Conditions for Operation are entered for equipment or functional 
inoperability. There are no physical alterations being made to the 
Auxiliary Feedwater System or related systems.Therefore, the 
proposed changes do 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 changes will not result in any physical alterations 
to the Auxiliary Feedwater System, any plant configuration, systems, 
equipment, or operational characteristics. There will be no changes 
in operating modes, or safety limits, or instrument limits. With the 
proposed changes in place, Technical Specifications will retain 
requirements for the Auxiliary Feedwater System. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes clarify the regulatory requirements for the 
Auxiliary Feedwater System as defined by NUREG-0635 and NUREG-0737. 
The times established are identical to those invoked by the present 
TechnicalSpecifications or to those previously reviewed and approved 
for use by the NRC. The proposed changes will

[[Page 56328]]

not alter any physical or operational characteristics of the 
Auxiliary Feedwater System and associated systems and equipment. 
Therefore, the proposed changes do not involve a 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
ElectricStation, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: July 25, 2002.
    Description of amendment request: The proposed amendments would 
change the Susquehanna Steam Electric Station Final Safety Analysis 
Report by revising the Reactor Pressure Vessel (RPV) Material 
Surveillance Program. Specifically, the licensee proposes to replace 
the current plant-specific RPV material surveillance program with the 
Boiling Water Reactor (BWR) Integrated Surveillance Program.
    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?
    No.
    The proposed change implements an integrated surveillance 
program that has been evaluated by the NRC staff as meeting the 
requirements of paragraph III.C of Appendix H to 10 CFR 50 [Title 10 
of the CODE OFFEDERAL REGULATIONS, Part 50]. Consequently, the 
proposed change does not significantly increase the probability of 
any accident previously evaluated. The proposed change provides the 
same assurance of RPV integrity. As a result, the consequences of 
any accident previously evaluated are not significantly increased. 
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?
    No.
    The proposed change does not involve a physical alteration of 
the plant(no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change maintains an equivalent level of RPV material surveillance 
and does not introduce any new accident initiators. 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?
    No.
    The proposed change has been evaluated as providing an 
acceptable alternative to the plant-specific RPV material 
surveillance program that meets the requirements of the regulations 
for RPV material surveillance. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    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: June 28, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by relaxing the secondary 
containment requirements and eliminating the Filtration, Ventilation, 
and Recirculation system (FRVS) charcoal filters.
    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 analyzed?
    Response: No.
    The definition of CORE ALTERATIONS has been revised to define 
that control rod movement, provided there are no fuel assemblies in 
the associated core cell is not a core alteration. This is 
consistent withStandard Technical Specifications (STS) NUREG-1433 
Vol.1, Rev. 2,Standard Technical Specifications, General Electric 
Plants, BWR/4.
    The TS presently provide[s] a period of 7 days to restore an 
inoperableFRVS ventilation unit when performing activities with the 
potential for draining the reactor vessel or discontinue such 
activities. Operation of the redundant train will ensure that the 
remaining subsystem is operable, that no failures, which could 
prevent automatic actuation, have occurred and that any other 
failures will be readily detected. This is consistent with STS, 
NUREG-1433 Vol.1, Rev. 2, Standard Technical Specifications,General 
Electric Plants, BWR/4.
    The proposed changes associated with the FHA [fuel handling 
accident] do not involve a change to structures, components, or 
systems that would affect the probability of an accident previously 
evaluated in the HopeCreek Updated Final Safety Analysis Report 
(UFSAR). The FHA for the HCGS is defined as a drop of a fuel 
assembly over irradiated assemblies in the reactor core 24 hours 
after reactor shutdown. AST [accident source term] is used to 
evaluate the dose consequences of a postulated accident. TheFHA has 
been analyzed without credit for Secondary Containment,Filtration 
Recirculation and Ventilation System (FRVS), and Control 
RoomEmergency Filtration (CREF) system. The resultant radiological 
consequences are within the acceptance criteria set forth in 10 CFR 
50.67 and Regulatory Guide [(RG)] 1.183. This amendment does not 
alter the methodology or equipment used directly in fuel handling 
operations. The equipment hatch, the personnel air locks, nor any 
other containment penetration, nor any component thereof is an 
accident initiator. Actual fuel handling operations are not affected 
by the proposed changes. Therefore, the probability of a Fuel 
Handling Accident is not affected with the proposed amendment. No 
other accident initiator is affected by the proposed changes.
    The Loss of Coolant Accident (LOCA) Dose Calculation has been 
revised to (1) eliminate credit for the FRVS recirculation charcoal 
filters, (2) reduce credited efficiency of FRVS vent charcoal 
filters, (3) reduceEngineered Safety Feature (ESF) leakage from 10 
gpm to 1 gpm and (4) reduce control room unfiltered in-leakage to 
350 cfm.
    These proposed changes do not eliminate any safety system. The 
changes are only associated with the credit provided by the system 
in reducing the radiological consequences and therefore, do not 
affect any accident initiator. The results of that analysis show 
that the Exclusion AreaBoundary (EAB), Low Population Zone (LPZ), 
and Control Room (CR) doses are of the same order of magnitude as 
the previous analysis and remain within the acceptance criteria in 
10 CFR 50.67 and Regulatory Guide 1.183.
    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 change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    Response: No.
    The proposed amendment will not create the possibility for a new 
or different type of accident from any accident previously 
evaluated. Changes to the allowable activity

[[Page 56329]]

in the primary and secondary systems do not result in changes to the 
design or operation of these systems. The evaluation of the effects 
of the proposed changes indicates that all design standard and 
applicable safety criteria limits are met.
    Equipment important to safety will continue to operate as 
designed. Component integrity is not challenged. The changes do not 
result in any event previously deemed incredible being made 
credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems. The 
systems affected by the changes are used to mitigate the 
consequences of an accident that has already occurred. The proposed 
TS changes and modifications do not significantly affect the 
mitigative function of these systems.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    The proposed changes revise the TS to establish operational 
conditions where specific activities represent situations during 
which significant radioactive releases can be postulated. These 
operational conditions are consistent with the design basis analysis 
and are established such that the radiological consequences are at 
or below the regulatory guidelines. Safety margins and analytical 
conservatisms are retained to ensure that the analysis adequately 
bounds all postulated event scenarios. The proposed TS continue[s] 
to ensure that the TEDE [total effective dose equivalent] for the 
CR, the EAB, and LPZ boundaries are below the corresponding 
acceptance criteria specified in 10 CFR 50.67 and RG1.183.
    Therefore, these changes do not involve a significant reduction 
in 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, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: Jacob Zimmerman, Acting.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

    Date of amendment request: July 25, 2002.
    Brief description of amendments: The proposed amendments would 
change the CPSES Facility Operating Licenses as follows: Section 
2.C.(4)(b) would be changed to be consistent with the license 
conditions stated in the U.S. Nuclear RegulatoryCommission (NRC) Order 
and Safety Evaluation issued December 21, 2001, which approved the 
direct transfer of ownership interest and operating authority for CPSES 
to TXU Generation Company LP; Section 2.E which requires reporting any 
violations of the requirements contained in Section 2.C of the licenses 
would be deleted. Additionally, Technical Specification Table 5.5-2 
``Steam Generator Tube Inspection,'' Table 5.5-3, ``Steam Generator 
Repaired Tube Inspection for Unit 1 Only,'' and Section 5.6.10, ``Steam 
Generator Tube Inspection Report,'' would be revised to delete the 
requirement to notify the NRC pursuant to 50.72(b)(2) of Title 10 of 
the Code of Federal Regulations (10 CFR) if the steam generator tube 
inspection results are in a C-3 classification.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The requested change to revise Section 2.C.(4)(b) of the 
OperatingLicenses is consistent with NRC Order and Safety Evaluation 
approvedDecember 21, 2001 for Facility Operating Licenses NPF-87 and 
NPR-89. The requested change to delete Section 2.E of the Operating 
Licenses and the changes to revise Technical Specification Table 
5.5-2, Table 5.5-3 and Section 5.6.10 are consistent with the 
changes recently implemented in 10 CFR 50.72 and 10 CFR 50.73.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed changes.Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed change and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, the proposed changes 
do not create a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, Reactor 
Coolant System pressure boundary, and containment structure) to 
limit the level of radiation dose to the public. This request 
involves administrative changes only.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: July 26, 2002.
    Brief description of amendment request: The proposed amendments 
would amend Operating Licenses DPR-58 and DPR-74 to add a license 
condition allowing a one-time 140-hour allowed outage time for the 
essential service water (ESW) system, to allow ESW pump replacement 
during plant operation.
    Date of publication of individual notice in Federal Register: 
August 8, 2002 (67FR 51603).
    Expiration date of individual notice: September 9, 2002.

[[Page 56330]]

Florida Power and Light Company, Docket No. 50-251, Turkey Point Plant, 
Unit 4, Miami-Dade County, Florida

    Date of amendment request: July 29, 2002.
    Description of amendment request: Revised Technical Specifications 
to allow use of an alternate method of determining rod position for a 
control rod with an inoperable rod position indication. Effective 
during the current operating cycle until repair of the indication 
system can be completed at the next outage.
    Date of publication of individual notice in the Federal Register: 
August 2, 2002(67 FR 50473).
    Expiration date of individual notice: August 16, 2002.

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 
OperatingLicense, Proposed No Significant Hazards Consideration 
Determination, andOpportunity 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 10CFR 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, SafetyEvaluation 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].

Dominion Nuclear Connecticut, Inc. et al., Docket Nos. 50-245, 50-336, 
and 50-423 Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New 
London County, Connecticut

    Date of amendment request: August 8, 2001.
    Brief description of amendment: The amendment incorporates two 
changes into each operating license. The physical protection (security) 
related license condition is revised to indicate that the physical 
security program plans listed, may, rather than do, contain safeguards 
information; and the plant name is changed from the ``Millstone Nuclear 
Power Station'' to the ``Millstone Power Station.''
    Date of issuance: August 8, 2002.
    Effective date: August 8, 2002, to be implemented within 60 days 
from the date of issuance.
    Amendment Nos.: Unit 1,-110, Unit 2-269, and Unit 3-208.
    Facility Operating License Nos. DPR-21, DPR-65 and NPF-49: The 
amendment revised the operating licenses.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52798). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 1, 2001, as supplemented 
by letters dated June 26, and August 5, 2002.
    Brief description of amendment: The amendment will revise the 
TechnicalSpecifications (TSs) limiting condition for operation and 
surveillance requirements associated with verification of reactor 
coolant system operational leakage. Conforming changes are also made to 
the associated TS Bases.
    Date of issuance: August 21, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 209.
    Facility Operating License No. NPF-49: Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57120). The supplements dated June 26 and August 5, 2002, were 
within the scope of the original application as published in the 
Federal Register and did not change the staff's proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety evaluation dated August 21, 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: April 19, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification(TS) 5.5.10, ``Technical Specification (TS) Bases Control 
Program,'' to provide consistency with the changes to 10 CFR 50.59 as 
published in the Federal Register(64 FR 53582) dated October 4, 1999, 
that became effective March 13, 2001.
    Date of issuance: August 15, 2002.
    Effective date: August 15, 2002.
    Amendment No.: 177.
    Facility Operating License No. NPF-21: The amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42821). TheCommission's related evaluation of the amendment is 
contained in a SafetyEvaluation dated August 15, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick NuclearPower Plant, Oswego County, New York

    Date of application for amendment: November 2, 2001, as 
supplemented January 9 and July 10, 2002.
    Brief description of amendment: The amendment changes the Current 
TechnicalSpecifications and the Improved Technical Specifications Main 
Steam IsolationValve Leakage Surveillance Requirement. The licensee 
will also make conforming changes to the associated Bases and the 
Primary Containment Leakage Rate TestingProgram.

[[Page 56331]]

    Date of issuance: August 13, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 275.
    Facility Operating License No. DPR-59: Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2923). TheJanuary 9 and July 10, 2002, letters provided clarifying 
information that was within the scope of the original application and 
did not change the staff's initial proposed no significant hazards 
consideration determination as published in the Federal Register. The 
January 9 supplement also corrected the original application date from 
November 2, 2000, to November 2, 2001.
    The Commission's related evaluation of the amendment is contained 
in a SafetyEvaluation dated August 13, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 22, 2001, as supplemented 
on March 5,2002.
    Brief description of amendment: The amendment revises the Technical 
Specification(TS) Surveillance Requirement (SR) 3/4.7.B.1.a.2 for the 
Standby Gas Treatment(SBGT) System and the associated TS Bases 3/
4.7.B.1, by increasing the SBGT inlet heaters minimum output testing 
requirement from 14 kW to 20 kW.
    Date of issuance: August 20, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 194.
    Facility Operating License No. DPR-35: Amendment revised the TSs.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57121). The supplement dated March 5, 2002, provided additional 
information that clarified the application, and did not expand the 
scope of the application or change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a SafetyEvaluation dated August 20, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., DocketNo. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: August 20, 2001, as supplemented 
on February 13, 2002.
    Brief description of amendment: The amendment to the Technical 
Specifications(TSs) revises certain requirements associated with 
demonstrating the operability of alternate trains when redundant 
equipment is made or found to be inoperable. The TSs revised include: 
4.4.B, 4.5.A.2, 4.5.A.3, 4.5.A.4, 4.5.B.2, 4.5.C.2, 4.5.C.3, 4.5.D.2, 
4.5.D.3, 4.5.E.2, 4.5.F.2, 4.5.H.1, 4.7.B.3.c, 4.10.B.1,4.10.B.3.b.2. 
Some format and typographical errors were also corrected.
    Date of Issuance: August 14, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-28: Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48292).The February 13, 2002, supplement was within the scope of the 
original application and did not change the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a SafetyEvaluation dated August 14, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., DocketNo. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: November 20, 2001, as 
supplemented on March 28, 2002.
    Brief description of amendment: This amendment moves Table 4.7.2, 
``PrimaryContainment Isolation Valves'' and references, to the 
Technical RequirementsManual; changes surveillance requirement 
4.7.B.1.b to reflect that the Standby GasTreatment system duct heater 
needs to meet relative humidity design-basis requirements; adds Section 
3.7.E, ``Reactor Building Automatic Ventilation SystemIsolation 
Valves,'' to the Table of Contents; removes wording in 3.5.A.4.a and b 
referencing a one-time 30-day Limiting Condition for Operation; and, 
makes administrative changes to Sections 5.3 and 6.4.
    Date of Issuance: August 21, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-28: Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66474). TheMarch 28, 2002, supplemented was within the scope of the 
original application and did not change the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 21, 2002.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 251, Turkey Point Plant, 
Unit 4, Miami-Dade County, Florida

    Date of amendment request: July 29, 2002, as supplemented August 14 
and August 16, 2002.
    Description of amendment request: The amendment revised Technical 
Specifications 3/4.1.3.1, 3/4.1.3.2 and 3/4.1.3.5 to allow the use of 
an alternate method of determining rod position for the control rod C-
9, until the end of Cycle 20 or until repairs can be conducted on the 
Analog Rod Indication System at the next outage of sufficient duration, 
whichever comes first.
    Date of issuance: August 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 216.
    Facility Operating License No. (DPR-41): Amendment revised the 
TechnicalSpecifications.
    Public comments requested as to proposed no significant hazards 
consideration(NSHC):Yes. August 2, 2002 (67 FR 50473). The licensee's 
August 14 and August 16, 2002, submittals of supplemental information 
did not affect the original no significant hazards consideration 
determination, and did not expand the scope of the request as noticed 
on August 2, 2002. The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided an opportunity to request 
a hearing by August 16, 2002, but indicated that if the Commission 
makes a final NSHC determination, any such hearing would take place 
after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent

[[Page 56332]]

circumstances, state consultation, and final NSHC determination are 
contained in a safety evaluation dated August 20, 2002.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Kahtan N. Jabbour, Acting.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point NuclearStation, Unit 2, Oswego County, New York

    Date of application for amendment: June 7, 2002.
    Brief description of amendment: The amendment deletes Section 
5.5.3, ``PostAccident Sampling,'' from the Technical Specifications and 
thereby eliminates the requirements to have and maintain the Post 
Accident Sampling System.
    Date of issuance: August 9, 2002.
    Effective date: As of the date of issuance to be implemented within 
180 days.
    Amendment No.: 106.
    Facility Operating License No. NPF-69: Amendment revises the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45570). The staff's related evaluation of the amendment is contained in 
a Safety Evaluation datedAugust 9, 2002.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
ElectricStation, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendments: April 2, 2002.
    Brief description of amendments: These amendments revised 
SurveillanceRequirement (SR) 3.0.3 to extend the delay period, before 
entering a LimitingCondition for Operation, following a missed 
surveillance. The delay period is 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 is added to SR 3.0.3: ``A risk evaluation shall 
be performed for anySurveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: August 12, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 205 and 179.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36932). TheCommission's related evaluation of the amendments is 
contained in a SafetyEvaluation dated August 12, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 29, 2002.
    Brief description of amendment: The amendment allows the use of the 
current pressure-temperature (P-T) limit curves through Cycle 12. The 
amendment also removes notes from the Technical Specifications that 
state that the curves are valid for 32 effective full power years.
    Date of issuance: August 13, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 139.
    Facility Operating License No. NPF-57: This amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34491). TheCommission's related evaluation of the amendment is 
contained in a SafetyEvaluation dated August 13, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425,Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: October 21, 2001, as 
supplemented by letters dated February 11 and May 27, 2002.
    Brief description of amendments: The amendments revised the 
TechnicalSpecifications, Table 3.3.1-1, ``Reactor Trip System 
Instrumentation'' and associated Bases B 3.3.1. A limit or ``clamp'' on 
the Overtemperature DeltaTemperature reactor trip function addresses 
design issues related to fuel rod design under transient conditions. In 
addition, editorial revisions to bases B 3.3.1 are included.
    Date of issuance: August 9, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 127 & 105.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15608). The supplements dated February 11, 2002, and May 27, 2002, 
provided clarifying information that did not change the scope of the 
October 30, 2001, application nor the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a SafetyEvaluation dated August 9, 2002.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
SteamElectric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: December 26, 2001, as supplemented by 
letters datedFebruary 4 and June 12, 2002.
    Brief description of amendments: The amendments revise TS 5.5.16, 
``ContainmentLeakage Rate Testing Program'' to extend the 10 CFR Part 
50, Appendix J, Type A,Containment Integrated Leak Rate Test date for 
Comanche Peak Steam ElectricStation, Units 1 and 2, from the fall of 
2002 to December 2008 for Unit 1, and from the fall of 2006 to December 
2012 for Unit 2. The following phrase implements this change in TS 
5.5.16.a: `` * * * as modified by the following exception: 1. NEI 94-
01--1995, Section 9.2.3: The first Type A Test performed after the 
December 7, 1993, Type A Test (Unit 1) and the December 1, 1997, Type 
ATest (Unit 2) shall be performed no later than December 15, 2008 (Unit 
1) andDecember 9, 2012 (Unit 2).''
    Date of issuance: August 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 98 and 98.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5340). TheFebruary 4 and June 12, 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 amendments is contained 
in a SafetyEvaluation dated August 15, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of August 2002.


[[Page 56333]]


For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-22197 Filed 8-30-02; 8:45 am]
BILLING CODE 7590-01-P