[Federal Register Volume 68, Number 62 (Tuesday, April 1, 2003)]
[Notices]
[Pages 15756-15771]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-7489]


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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 March 7, 2003 through March 20, 2003. The 
last biweekly notice was published on March 18, 2003 (68 FR 12946).

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

[[Page 15757]]

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, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By May 1, 2003, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text to 10 CFR 2.714(d), please see 67 FR 20884; April 29, 
2002.
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected].

[[Page 15758]]

A copy of the request for hearing and petition for leave to intervene 
should also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and because of 
continuing disruptions in delivery of mail to United States Government 
offices, it is requested that copies be transmitted either by means of 
facsimile transmission to 301-415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

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

    Date of amendment request: May 10, 2002, as supplemented March 12, 
2003.
    Description of amendment request: Carolina Power & Light Company 
(the licensee) is proposing changes to Appendix A, Technical 
Specifications (TS), and appendix B, Additional Conditions, of Facility 
Operating License No. DPR-23 for the H. B. Robinson Steam Electric 
Plant, Unit No. 2 (HBRSEP2). These changes will revise the licensing 
basis for HBRSEP2 to implement the Alternative Source Term (AST) 
described in Regulatory Guide 1.183, ``Alternative Radiological Source 
Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors.'' Implementation of the AST will allow for removal of the 
cycle operating length restriction from appendix B, Additional 
Conditions, of the Operating License, as the AST radiological 
consequence analyses support operation for an entire cycle at the 
increased power level approved in License Amendment No. 196. The AST is 
used by the licensee in evaluating the radiological consequences of the 
following Updated Final Safety Analysis Report Chapter 15 accidents:
    [sbull] Main Steam Line Break,
    [sbull] Reactor Coolant Pump Shaft Seizure,
    [sbull] Single Rod Control Cluster Assembly Withdrawal,
    [sbull] Steam Generator Tube Rupture,
    [sbull] Large Break Loss-of-Coolant Accident, and
    [sbull] Waste Gas Decay Tank Rupture.
    In addition, revised atmospheric dispersion factors for onsite and 
offsite dose consequences have been calculated and incorporated in the 
reanalysis of these events.
    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.
    Implementation of the Alternative Source Term does not affect 
the design or operation of HBRSEP, Unit No. 2. Rather, once the 
occurrence of an accident has been postulated, the new source term 
is an input to evaluate the consequences of the postulated accident. 
The implementation of the Alternative Source Term has been evaluated 
in revisions to limiting design basis accidents at HBRSEP, Unit No. 
2. Based on the results of these analyses, it has been demonstrated 
that, with the requested changes to the Technical Specifications, 
the dose consequences of these limiting events are within the 
regulatory guidance provided by the NRC. This guidance is presented 
in 10 CFR 50.67 and Regulatory Guide 1.183. The proposed Technical 
Specifications changes result in more restrictive requirements and 
support the revisions to the limiting design basis accident 
analyses.
    Therefore, 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 changes do not affect plant structures, systems or 
components. The Alternative Source Term and those plant systems 
affected by implementing the proposed changes do not initiate design 
basis accidents.
    Thus, this 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 the Margin of Safety.
    The proposed changes are associated with the implementation of a 
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis 
implements an Alternative Source Term in accordance with 10 CFR 
50.67 and the associated Regulatory Guide 1.183. The results of the 
revised limiting design basis analyses are subject to revised 
acceptance criteria. The analyses have been performed using 
conservative methodologies in accordance with the regulatory 
guidance. The dose consequences of the limiting design basis events 
are within the acceptance criteria found in the regulatory guidance 
associated with Alternative Source Terms.
    The proposed changes continue to ensure that doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limits. 
Specifically, the margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits, which are conservatively set below the 
10 CFR 50.67 limits. With respect to control room personnel doses, 
the margin of safety (the difference between the 10 CFR 50.67 limits 
and the regulatory limits defined by 10 CFR 50, Appendix A, [General 
Design] Criterion 19 (GDC-19)) continues to be satisfied.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based on the above discussion, Progress Energy Carolinas, Inc., 
also known as Carolina Power and Light Company, 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602-1551.
    NRC Section Chief: Allen G. Howe.

[[Page 15759]]

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: February 17, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification Surveillance Requirement 3.10.1.9 to 
require that the Standby Shutdown Facility (SSF) diesel generator (DG) 
be loaded to at least 3280 kilowatts during the surveillance. The 
current requirement is that the SSF DG be loaded to at least 3000 
kilowatts during the surveillance. The change supports resolution of an 
Oconee design basis issue associated with SSF pressurizer heater 
capacity.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the 
determination that this amendment request involves a No Significant 
Hazards Consideration by applying the standards established by the 
NRC in 10 CFR 50.92. This ensures that operation of the facility in 
accordance with the proposed amendment would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    This change revises the loading of the Standby Shutdown Facility 
(SSF) Diesel Generators (DG) to = 3280 kW. The design 
rating of the DG is currently 3500 kW. Since the proposed loading is 
within the design rating already evaluated, this proposed change 
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:
    As stated above, the proposed revision revises the DG loading to 
an analytical value that is within the equipment's design limit. 
Applicable load and support system calculations have been revised 
and results have shown that the increase does not adversely affect 
the ability of the SSF diesel generator or SSF to perform its 
intended safety function. Additionally, this change is bounded by 
all of the existing accidents and does not create the possibility of 
a new or different kind of accident from any kind of accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not adversely affect any plant safety 
limits, set points, or design parameters. The change also does not 
adversely affect the fuel, fuel cladding, Reactor Coolant System, or 
containment integrity. Therefore, the proposed change does not 
involve a 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.

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

    Date of amendment request: January 31, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS) 3.4.11, ``RCS Pressure 
and Temperature (P/T) Limits,'' to incorporate revised P/T curves. The 
revised P/T curves are based on calculations performed in accordance 
with General Electric (GE) Topical Report NEDC-32983P, ``General 
Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux 
Evaluation.'' The NEDC-32983P methodology is consistent with the 
guidance contained in Regulatory Guide (RG) 1.190, ``Calculational and 
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' 
dated March 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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes request for LaSalle County Station, Units 1 
and 2, that the pressure and temperature (P/T) limit curves in TS 
3.4.11, ``RCS Pressure and Temperature (P/T) Limits,'' and 
Surveillance Requirement (SR) 3.4.11.1 and SR 3.4.11.2 be revised. 
The revised curves were developed using the methodology of GE 
Topical Report NEDC-32983P, ``General Electric Methodology for 
Reactor Pressure Vessel Fast Neutron Flux Evaluation.'' NEDC-32983P 
methodology has been previously approved by the NRC for use by 
licensees. The P/T limits are prescribed during normal operation to 
avoid encountering pressure, temperature, and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause nonductile failure of the reactor coolant pressure boundary, a 
condition that is unanalyzed. Thus, the proposed changes do not have 
any affect on the probability of an accident previously evaluated.
    The P/T curves are used as operational limits during heatup or 
cooldown maneuvering, when pressure and temperature indications are 
monitored and compared to the applicable curve to determine that 
operation is within the allowable region. The P/T curves provide 
assurance that station operation is consistent with previously 
evaluated accidents. Thus, the radiological consequences of any 
accident previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not change the response of plant 
equipment to transient conditions. The proposed changes do not 
introduce any new equipment, modes of system operation or failure 
mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes adopt P/T curves that have been developed 
using the methodology of GE Topical Report NEDC-32983P. The NEDC-
32983P methodology is consistent with the guidance contained in RG 
1.190, ``Calculational and Dosimetry Methods for Determining 
Pressure Vessel Neutron Fluence,'' dated March 2001. In a letter 
dated September 14, 2001, the NRC approved NEDC-32983P for use by 
licensees.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the above, EGC concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief : Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: February 27, 2003.

[[Page 15760]]

    Description of amendment request: The proposed amendments revise 
the Technical Specifications to reflect a one-time deferral of the 
primary containment Type A leak rate test to no later than July 22, 
2009, for Unit 1 and no later than May 16, 2008, for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise Quad Cities Nuclear Power 
Station (QCNPS), Units 1 and 2, Technical Specification (TS) 5.5.12, 
``Primary Containment Leakage Rate Testing Program,'' to reflect a 
one-time deferral of the primary containment Type A test to no later 
than July 22, 2009, for Unit 1, and no later than May 16, 2008, for 
Unit 2. The current Type A test interval of 10 years, based on past 
performance, would be extended on a one-time basis to 15 years from 
the last Type A test.
    The function of the primary containment is to isolate and 
contain fission products released from the reactor coolant system 
(RCS) following a design basis loss of coolant accident (LOCA) and 
to confine the postulated release of radioactive material to within 
limits. The test interval associated with Type A testing is not a 
precursor of any accident previously evaluated. Therefore, extending 
this test interval on a one-time basis from 10 years to 15 years 
does not result in an increase in the probability of occurrence of 
an accident. The successful performance history of Type A testing 
provides assurance that the QCNPS primary containments will not 
exceed allowable leakage rate values specified in the TS and will 
continue to perform their design function following an accident. The 
risk assessment of the proposed change has concluded that there is 
an insignificant increase in total population dose rate and an 
insignificant increase in the conditional containment failure 
probability.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change for a one-time extension of the Type A tests 
for QCNPS, Units 1 and 2, will not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient and accident conditions. The proposed change does not 
introduce any new equipment or modes of system operation. No 
installed equipment will be operated in a new or different manner. 
As such, no new failure mechanisms are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    QCNPS, Units 1 and 2, are General Electric BWR/3 [boiling water 
reactor class 3] plants with Mark I primary containments. The Mark I 
primary containment consists of a drywell, which encloses the 
reactor vessel, reactor coolant recirculation system, and branch 
lines of the RCS; a toroidal-shaped pressure suppression chamber 
containing a large volume of water; and a vent system connecting the 
drywell to the water space of the suppression chamber. The primary 
containment is penetrated by access, piping, and electrical 
penetrations.
    The integrity of the primary containment penetrations and 
isolation valves is verified through Type B and Type C local leak 
rate tests (LLRTs) and the overall leak tight integrity of the 
primary containment is verified by a Type A integrated leak rate 
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors.'' These 
tests are performed to verify the essentially leak tight 
characteristics of the primary containment at the design basis 
accident pressure. The proposed change for a one-time extension of 
the Type A tests do not affect the method for Type A, B, or C 
testing, or the test acceptance criteria. In addition, based on 
previous Type A testing results, EGC [Exelon Generation Company, 
LLC] does not expect additional degradation, during the extended 
period between Type A tests, which would result in a significant 
reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: February 27, 2003.
    Description of amendment request: The proposed amendments add a 
surveillance requirement to perform a quarterly trip unit calibration 
of the reactor protection system scram discharge volume water level--
high differential pressure switches.
    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 proposed Technical Specifications (TS) change adds a trip 
unit calibration surveillance requirement (SR) for the analog trip 
units associated with the Scram Discharge Volume (SDV) Water Level--
High Trip Function for the Reactor Protection System (RPS) 
Instrumentation. Specifically, SR 3.3.1.1.11 is added to Function 
7.b of TS Table 3.3.1.1-1, ``Reactor Protection System 
Instrumentation.'' In addition, the proposed change revises Function 
7.a of TS Table 3.3.1.1-1 to delete a reference to thermal switches, 
applicable to Unit 1 through cycle 17. The change to Function 7.a is 
editorial, since Unit 1 SDV level instrumentation has been upgraded 
to replace Fluid Components International thermal switches with 
Magnetrol float switches.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to prevent, detect, or 
mitigate accidents. Therefore, these proposed changes will not 
involve an increase in the probability of an accident previously 
evaluated. Additionally, these proposed changes do not increase the 
consequences of an accident previously evaluated because the 
proposed changes do not adversely impact structures, systems, or 
components. The proposed changes establish requirements that ensure 
components are operable when necessary for the prevention or 
mitigation of accidents or transients. Furthermore, there will be no 
change in the types or significant increase in the amounts of any 
effluents released offsite.
    In summary, 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.
    There is no change being made to the parameters within which 
Quad Cities Nuclear Power Station (QCNPS) is operated. The proposed 
changes do not adversely impact the manner in which the SDV Water 
Level--High RPS instrumentation will operate under normal and 
abnormal operating conditions. The proposed changes will not alter 
the function demands on credited equipment. No alteration in the 
procedures, which ensure QCNPS remains within analyzed limits, is 
proposed, and no change is being made to procedures relied upon to 
respond to an off-normal event. Therefore, these proposed changes 
provide an equivalent level of safety and will not create the 
possibility of a new or different

[[Page 15761]]

kind of accident from any accident previously evaluated. The changes 
in methods governing normal plant operation are consistent with the 
current safety analysis assumptions. Therefore, these proposed 
changes do not create the possibility of 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.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
and actions. The proposed changes do not affect the probability of 
failure or availability of the affected instrumentation, and the 
proposed changes do not revise any allowable values for RPS 
functions. Therefore, it is concluded that the proposed changes do 
not result in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: January 14, 2003.
    Description of amendment request: This license amendment request 
proposes a change to Technical Specifications (TSs) 5.1.1, 5.4.1, and 
5.5.1 that would replace the requirement for the plant manager to 
approve administrative procedures and the Offsite Dose Calculation 
Manual. The plant manager approval signature would be replaced with the 
signature of a procedurally authorized individual who would be a more 
appropriate authority for approval of the activity.
    Basis for proposed no significant hazards consideration 
determination: As required by Section 50.91(a) of Title 10 of the Code 
of Federal Regulations (10 CFR), 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 replace the plant manager's approval with 
the approval by an authorized individual is consistent with the 
requirements of Regulatory Guide 1.33 and American National 
Standards Institute (ANSI) N18.7-1976/American Nuclear Society (ANS) 
3.2. The authorized individuals are management and supervisory 
personnel who satisfy the requirements of ANSI N18.1-1971. Use of 
ANSI N18.1-1971 is consistent with the requirements of the existing 
TS and Updated Safety Analysis Report (USAR). The change is 
administrative and does not impact or otherwise affect the physical 
plant.
    The proposed change to the License Condition to delete the 
reporting time frame eliminates duplication of a requirement that is 
already an integral part of 10 CFR 50.73 which is referenced in the 
License Condition. The proposed change is administrative and does 
not impact or otherwise affect the physical plant.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated. The 
proposed administrative changes do not involve any physical 
modifications to the facility nor add new equipment. The methods of 
plant operation have not been altered. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed changes will not involve a significant reduction 
in the margin of safety.
    The proposed changes are administrative in nature and have no 
direct impact upon any plant safety analyses. 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: January 30, 2003.
    Description of amendment request: This license amendment request 
would modify the existing minimum critical power ratio (MCPR) safety 
limit contained in Technical Specification (TS) 2.1.1.2. Specifically, 
the change modifies the MCPR safety limit values, as calculated by 
Global Nuclear Fuel (GNF), by decreasing the limit for two 
recirculation loop operation from 1.10 to 1.07, and decreasing the 
limit for single recirculation loop operation from 1.11 to 1.08. The 
change resulted from the core reload analysis performed for the Perry 
Nuclear Power Plant (PNPP) fuel cycle 10.
    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.
    PPNP Updated Safety Analysis Report (USAR) Section 4.2, ``Fuel 
System Design,'' states the PNPP fuel system design bases are 
provided in the General Electric Topical Report, NEDE-24011-P-A, 
``General Electric Standard Application for Reactor Fuel (GESTAR 
II).'' The MCPR Safety Limit is one of the limits used to protect 
the fuel in accordance with the design basis. The MCPR Safety Limit 
establishes a margin to the onset of transition boiling. The basis 
of the MCPR Safety Limit remains the same, ensuring that greater 
than 99.9% of all fuel rods in the core avoid transition boiling. 
The methodology used to determine the MCPR Safety Limit values is 
contained within GESTAR II and is NRC approved. The change does not 
result in any physical plant modifications or physically affect any 
plant components. As a result, there is no increase in the 
probability of occurrence of a previously analyzed accident.
    The fundamental sequences of accidents and transients have not 
been altered. The Safety Limit MCPR is established to avoid fuel 
damage in response to anticipated operational occurrences. 
Compliance with a MCPR Safety Limit greater than or equal to the 
calculated value will ensure that less than 0.1% of the fuel rods 
will experience boiling transition. This in turn ensures fuel damage 
does not occur following transitions due to excessive thermal 
stresses on the fuel cladding. The MCPR Operating Limits are set 
higher (i.e., more conservative) than the Safety Limit such that 
potentially limiting plant transients prevent the MCPR from 
decreasing below the MCPR Safety Limit during the transient. 
Therefore, there is no impact on any limiting USAR Appendix 15B 
transients. The radiological consequences remain the same as 
previously stated in the USAR. Therefore, the consequences of an 
accident do not increase over previous evaluations in the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit basis is preserved, which is to ensure 
that transition boiling does not occur in at least 99.9% of the fuel 
rods in the core as a result of the postulated limiting transient. 
The values are calculated in accordance with GESTAR II. The GESTAR 
II analyses have been accepted by the NRC.

[[Page 15762]]

The MCPR Safety Limit is one of the limits established to ensure the 
fuel is protected in accordance with the design basis. The function, 
location, operation, and handling of the fuel remain unchanged. No 
changes in the design of the plant or the method of operating the 
plant are associated with these revised safety limit values. 
Therefore, no new or different kind of accident from any previously 
evaluated is created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This change revises the PNPP MCPR Safety Limit values. The new 
MCPR Safety Limit values reflect changes due to the Cycle 10 core 
reload, but do not alter the design or function of any plant system, 
including the fuel. The new MCPR Safety Limit values were calculated 
using NRC-approved methods described in GESTAR II. The proposed MCPR 
Safety Limit values continue to satisfy the fuel design safety 
criteria which ensures that transition boiling does not occur in at 
least 99.9% of the fuel rods in the core as a result of the 
postulated limiting transient. Therefore, the proposed values for 
the MCPR Safety Limit do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: December 10, 2002.
    Description of amendment request: The proposed amendment would 
revise the Unit 2 reactor coolant system (RCS) pressure-temperature 
curves in Technical Specification (TS) Figures 3.4-2 and 3.4-3 and 
associated TS Bases. The revised curves will bound operation of the 
unit for the remainder of its current license duration and bound 
operation with planned license amendments to increase the power level 
at which the unit is allowed to operate. In support of this proposed 
amendment, Indiana Michigan Power (I&M) has submitted a request, in 
accordance with 10 CFR 50.60, ``Acceptance Criteria for Fracture 
Prevention Measures for Lightwater Nuclear Power Reactors for Normal 
Operation,'' for exemption from requirements in 10 CFR Part 50, 
Appendix G, ``Fracture Toughness Requirements.''
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.

Probability of Occurrence of an Accident Previously Evaluated

    The proposed change will revise the RCS pressure-temperature 
curves to reflect new limiting reactor vessel materials, to bound 
operation of the reactor up to 3600 MWt for the current fuel cycle 
and beyond, to reflect new fluence analysis methodology, to reflect 
the use of ASME [American Society of Mechanical Engineers] Code Case 
N-641, to include boltup limits, and to no longer include instrument 
uncertainty margins.
    The proposed change will not result in physical changes to 
structures, systems, or components (SSCs), or changes to event 
initiators or precursors. The proposed change will not affect the 
ability of personnel to control RCS pressure at low temperatures 
and, thereby, ensure the integrity of the reactor coolant pressure 
boundary. Use of Code Case N-641 in developing the proposed revision 
to the RCS pressure-temperature curves is in accordance with 
methodologies accepted by the ASME. These methodologies provide 
assurance that the reactor vessel will withstand the effects of 
normal cyclic loads due to temperature and pressure changes, and 
provide an acceptable level of protection against brittle failure.
    Additionally, the proposed changes will not impact the design or 
operation of plant systems such that previously analyzed SSCs will 
be more likely to fail. The initiating conditions and assumptions 
for accidents described in the UFSAR [updated final safety analysis 
report] will remain as previously analyzed. Therefore, the proposed 
changes will not involve a significant increase in the probability 
of an accident previously evaluated.

Consequences of an Accident Previously Evaluated

    The proposed change does not reduce the ability of any SSC to 
limit the radiological consequences of accidents described in the 
UFSAR. The proposed change will not alter any assumptions made in 
the analysis of radiological consequences of previously evaluated 
accidents, nor does it affect the ability to mitigate these 
consequences. No new or different radiological source terms will be 
generated as a result of the proposed change. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    The format changes will improve the appearance of the affected 
pages but will not affect any requirements. In summary, the 
probability of occurrence and the consequences of an accident 
previously evaluated will not be significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not result in physical changes to SSCs. 
The proposed change will not involve the addition or modification of 
plant equipment (no new or different type of equipment will be 
installed) nor will it alter the design of any plant systems. The 
proposed change solely involves RCS pressure-temperature limits. The 
types of potential accidents associated with these limits have been 
previously identified and evaluated. No new accident scenarios, 
accident or transient initiators or precursors, failure mechanisms, 
or single failures will be introduced as a result of the proposed 
changes. No new or different modes of failure will be created. The 
format changes will improve the appearance of the affected pages but 
will not affect any requirements. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed RCS pressure-temperature curves will continue to 
provide adequate margins of protection for the reactor coolant 
pressure boundary. The proposed changes have been determined, 
through supporting analyses, to be in accordance with the 
methodologies and criteria set forth in the applicable regulations, 
or in accordance with technically adequate alternatives. Compliance 
with these methodologies provides adequate margins of safety and 
ensures that the reactor coolant pressure boundary will withstand 
the effects of normal cyclic loads due to temperature and pressure 
changes as well as the loads associated with postulated faulted 
events as described in the UFSAR. The format changes will improve 
the appearance of the affected pages but will not affect any 
requirements. Therefore, the proposed change will not significantly 
reduce 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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), 
Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: February 25, 2003.
    Description of amendment request: The proposed Technical 
Specification

[[Page 15763]]

(TS) changes will add an allowed outage time (AOT) for Engineered 
Safety Features Actuation System (ESFAS) Instrumentation channels to be 
out of service in a bypassed state.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The addition of an ACTION STATEMENT and the addition of an AOT 
(and its associated actions if not met) for a TS action statement 
are neither an accident initiator nor precursor. The ESFAS actuates 
in response to an accident and has a mitigating function. Increasing 
the TS requirements for specific TS instrument loops provides 
additional assurance that the channels will be capable of performing 
their design function in the event of a DBA [design-basis accident]. 
The ability of the operations staff to respond to an evaluated 
accident or plant transient will not be hampered. This change 
provides conservative requirements to assure that the design basis 
of the plant is maintained.
    Addition of conservative changes to the Engineered Safety 
Feature Actuation System Instrumentation does not contribute to the 
initiation of any accident evaluated in the FSAR [Final Safety 
Analysis Report]. Supporting factors are as follows:
    [sbull] The changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. These changes are conservative and 
consistent with the Standard Technical Specifications, NUREG-1431, 
Rev. 2. There are no deletions from the Technical Specifications 
made by these changes, nor relaxation in any applicability, action, 
or surveillance requirements.
    [sbull] Overall plant performance and operation is not altered 
by the proposed changes. There are to be no plant hardware changes 
as a result of this proposed change and only minimal procedural 
changes.
    Therefore, since the Engineered Safety Feature Actuation System 
Instrumentation are treated more conservatively, the probability of 
occurrence or consequences of an accident evaluated in the VCSNS 
FSAR will be no greater than the original design basis of the plant.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. Additionally, the addition of an 
ACTION STATEMENT and an AOT with conservative requirements are 
intended to assure that the plant is in a safe configuration and can 
meet accident analyses assumptions. These changes are conservative 
and consistent with the Improved Technical Specifications, NUREG-
1431, Rev. 2. No new accident initiator mechanisms are introduced 
since:
    [sbull] No physical changes to the Engineered Safety Feature 
Actuation System Instrumentation are made.
    [sbull] No deletions from the Technical Specifications are made.
    [sbull] No relaxations in any applicability, action, or 
surveillance requirements are made.
    Since the safety and design requirements continue to be met and 
the integrity of the reactor coolant system pressure boundary is not 
challenged, no new accident scenarios have been created. Therefore, 
the types of accidents defined in the FSAR continue to represent the 
credible spectrum of events to be analyzed, which determine safe 
plant operation.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change requires that an instrument channel for an 
Engineered Safety Feature remain operable or be restored to 
operability within a reasonable time period, otherwise a controlled 
shutdown is required. This conforms to the safety analysis where the 
plant and its systems, structures and components must be capable of 
performing the safety function while a DBA is occurring, in the 
presence of a worst case single failure.
    This is not a reduction in a margin of safety, since it restores 
the margin that was designed into the plant.
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. These changes are conservative and 
consistent with the Standard Technical Specifications, NUREG-0452, 
Rev. 5. The proposed changes impose more restrictive operating 
limitations, and their use provides increased assurance that the 
Engineered Safety Feature Actuation System Instrumentation remains 
operable. Since the changes are conservative additions, it is 
concluded that the changes do not involve a significant reduction in 
the margin of safety. This is not a reduction in a margin of safety, 
since it restores the margin that was designed into the plant.
    Pursuant to 10 CFR 50.91, the preceding analyses provides a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone 
County, Alabama

    Date of amendment request: February 13, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 4.2.1, Fuel Assemblies, to modify 
the fuel design description to encompass Framatome Advanced Nuclear 
Power (FANP) fuel assemblies and also to modifiy TS 4.3, Fuel Storage, 
to remove nomenclature specific to Global Nuclear Fuels analysis 
methods. The proposed TS changes are needed to allow the receipt and 
storage of Framatome fuel assemblies.
    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 in accordance with the three standards set forth in 10 
CFR 50.92(c), which are presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment revises TS 4.2.1, Fuel Assemblies, to 
modify the fuel design description to accommodate FANP fuel designs. 
The change to TS 4.2.1 is administrative and simply adds descriptive 
text to reflect that FANP fuel assemblies have a water channel.
    To make the fuel storage TS compatible with the storage of GNF 
[Global Nuclear Fuels] and FANP fuel, the proposed amendment also 
modifies TS 4.3, Fuel Storage, to delete criteria specific to GNF 
fuel storage criticality analysis methods. BFN criticality analysis 
and storage requirements continue to be adequately described in the 
Updated Final Safety Analysis Report (UFSAR) and in existing TS 
4.3.1.1.b, TS 4.3.1.1.c, TS 4.3.1.2.b, 4.3.1.2.c, and 4.3.1.2.d. 
Hence, the proposed elimination of the GNF-specific criteria in TS 
4.3 does not affect BFN design basis requirements associated with 
ensuring adequate criticality margins are maintained for fuel 
storage.
    The requested TS changes do not involve any plant modifications 
or operational changes that could affect system reliability, 
performance, or the possibility of operator error. The requested 
changes do not affect any postulated accident precursors, do not 
affect accident mitigation systems, and do not introduce any new 
accident initiation methods. Therefore, the proposed TS change does 
not involve an 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 changes to TS do not affect the performance of 
any BFN structure,

[[Page 15764]]

system, or component credited with mitigating any accident 
previously evaluated. Fuel storage criticality analyses will 
continue to be performed in accordance with established UFSAR 
commitments that are independent are fuel vendor specific methods. 
The TS changes do not introduce new modes of operation or involve 
plant modifications.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed amendment modifies TS 4.3, Fuel Storage, to 
remove nomenclature specific to GNF criticality analysis methods. 
Fuel storage criticality analyses will continue to be performed in 
accordance with UFSAR commitments and the remaining TS commitments 
in accordance with FANP accepted methods, which specify appropriate 
criteria and conservatisms. Therefore, the proposed TS change does 
not involve a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: December 19, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Chapter 5.0, ``Administrative 
Controls,'' to incorporate three approved TS Task Force (TSTF) changes: 
TSTF-258, Revision 4; TSTF-299, Revision 0; and TSTF-308, Revision 1. 
These changes have been incorporated in Revision 2 of NUREG 1431, 
``Standard Technical Specifications Westinghouse Plants.''
    In addition, the amendment proposes two editorial changes. These 
changes either update personnel titles with the titles currently used 
at WBN and TVA's other nuclear units or clarify required staffing 
levels.
    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 in accordance with the three standards set forth in 10 
CFR 50.92(c), which are presented below:

    1. Does the proposed change involve a significant increase in 
the probability of consequences of an accident previously evaluated?
    No. The proposed changes affect only administrative requirements 
or programs. As indicated below, the justification for five of the 
changes (Parts 2 through 4 of Change Number 2 and Change Numbers 3, 
5 [only Parts 1 and 2 of Change 5], 6, and 7) is based on the 
existence of a regulation or other regulatory document which 
controls the administrative requirements. For these changes, the 
proposed amendment modifies the administrative TS to make it 
consistent with the current regulations or NRC guidance document. 
Two changes (Change Number 1 and Part 1 of Change Number 2) are 
strictly editorial. In addition, two changes (Change Number 4 and 
Part 3 of Change Number 5) add a requirement to make the program 
consistent with the criteria for Surveillance Requirements in the 
Improved Standard Technical Specifications (ISTS). Based on the 
preceding information, the proposed amendment does not involve 
technical changes to the configuration or operation of the plant 
there is not a significant increase in the probability or 
consequences of an accident previously evaluated:

------------------------------------------------------------------------
                                  Administrative   Justification for the
          Change No.             section affected          change
------------------------------------------------------------------------
1.............................  5.1,               Editorial update of
                                 ``Responsibility   staff titles.
                                 ,'' Section
                                 5.1.2.
2.............................  5.2.2, ``Unit      Part 1 of Change
                                 Staff''.           number 2--Editorial
                                                    clarification of the
                                                    number of non-
                                                    licensed operators
                                                    required for the
                                                    operation of WBN
                                                    Unit 1. Parts 2
                                                    through 4 of Change
                                                    Number 2--The
                                                    existing
                                                    administrative
                                                    requirements are
                                                    revised to align the
                                                    requirements with 10
                                                    CFR 50.54.
3.............................  5.3, ``Unit Staff  Adds TS 5.3.2 which
                                 Qualifications,'   clarifies the
                                 ' Section 5.3.2.   ``Operator'' and
                                                    ``Senior Operator''
                                                    definitions in 10
                                                    CFR 55.4 and ties
                                                    these positions to
                                                    the requirements of
                                                    10 CFR 50.54.
4.............................  5.7.2.4,           WBN TS 5.7.2.4 serves
                                 ``Primary          the same function as
                                 Coolant Sources    a Surveillance
                                 Outside            Requirement (SR).
                                 Containment.       The proposed change
                                                    structures TS
                                                    5.7.2.4 so that it
                                                    is consistent with
                                                    other ISTS SRs and
                                                    the frequency
                                                    extension allowed by
                                                    SR 3.0.2.
5.............................  5.7.2.7,           The intent of the
                                 ``Radioactive      revisions to this TS
                                 Effluent           are to: 1) eliminate
                                 Controls           possible confusion
                                 Program''.         or improper
                                                    implementation of
                                                    the requirements of
                                                    10 CFR 20; 2)
                                                    clarifies the
                                                    wording to not
                                                    require dose
                                                    projections for a
                                                    calendar quarter and
                                                    a calendar year
                                                    every 31 days; 3)
                                                    structures the TS so
                                                    that it is
                                                    consistent with
                                                    other ISTS SRs.
6.............................  5.9.4, ``Monthly   The proposed change
                                 Operating          makes the TS
                                 Reports''.         reporting
                                                    requirements
                                                    consistent with the
                                                    reporting
                                                    requirements in
                                                    Generic Letter 97-
                                                    02.
7.............................  5.11, ``High       The proposed revision
                                 Radiation Area''.  updates the TS to be
                                                    consistent with 10
                                                    CFR 20.1601(c) and
                                                    updates the
                                                    acceptable alternate
                                                    controls to those
                                                    given in 10 CFR
                                                    20.1601.
------------------------------------------------------------------------

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. As indicated above, the proposed changes do not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) or changes in methods controlling 
normal plant operation. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes will not reduce the margin of safety 
because they have no effect on assumptions made in WBN's safety 
analysis or the configuration of plant equipment important to 
safety. Additionally, several of the proposed revisions adjust the 
administrative requirements to be consistent with existing 
regulations or NRC guidance documents and therefore, will not 
adversely impact plant safety. The balance of the proposed changes 
are editorial updates or adjust a program to be consistent with the 
ISTS.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority,

[[Page 15765]]

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: February 14, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for Watts Bar Nuclear Plant 
(WBN), Unit 1. The proposed TS change would allow WBN Unit 1 to be 
refueled and operated using the Westinghouse 17x17 Robust Fuel 
Assembly-2 (RFA-2) design commencing with Cycle 6 in September 2003.
    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 Loss of Coolant Accident (LOCA) and non-LOCA transients 
and accidents which are potentially affected by the parameters and 
assumptions associated with the use of RFA-2 (including the effects 
of Tritium Producing Burnable Absorber Rods, TPBARs) have been 
evaluated/analyzed and all design standards and applicable safety 
criteria are met. The consideration of these changes does not result 
in a situation where the design material, and construction standards 
that were applicable prior to the change are altered. Therefore, the 
changes occurring with the use of RFA-2 will not result in any 
additional challenges to plant equipment that could increase the 
probability of any previously evaluated accident.
    The changes associated with the use of RFA-2 do not affect plant 
systems such that their function in the control of radiological 
consequences is adversely affected. TVA's evaluation documents that 
the design standards and applicable safety criteria limits continue 
to be met and, therefore, fission barrier integrity is not 
challenged. The fuel rod design (the first fission product barrier) 
is not changed. Compared to the current grid design on the resident 
fuel, the RFA-2 grid design provides improved resistance to fuel rod 
fretting. The RFA-2 fuel changes have been shown not to adversely 
affect the response of the plant to postulated accident scenarios. 
These changes will therefore not affect the mitigation of the 
radiological consequences of any accident described in the Final 
Safety Analysis Report (FSAR).
    Therefore, since the actual plant configuration, performance of 
systems, and initiating event mechanisms are not being changed as a 
result of this evaluation, TVA has concluded that 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 possibility for a new or different type of accident from 
any accident previously evaluated is not created since the changes 
associated with the use of RFA-2 do not result in a change to the 
design basis of any plant component or system. The evaluation of the 
effects of the use of RFA-2 shows that all design standards and 
applicable safety criteria limits are met. Specifically, the results 
of the evaluations/analyses lead to the following conclusions:
    1. The RFA-2 fuel design for Watts Bar Unit 1 is mechanically 
compatible with the current fuel assemblies, core components, the 
control rods and the reactor internals interfaces.
    2. The structural integrity of the RFA-2 fuel design has been 
evaluated for seismic/LOCA loadings for Watts Bar Unit 1. Evaluation 
of the RFA-2 fuel assembly component stresses and grid impact forces 
due to postulated faulted condition accidents verified that the fuel 
assembly design is structurally acceptable.
    3. The changes to the nuclear characteristics due to the 
transition to the RFA-2 fuel assembly design will be within the 
range normally seen from cycle to cycle due to fuel management.
    4. The RFA-2 fuel assembly design is hydraulically compatible 
with the current fuel assemblies.
    5. The core design and safety analyses documented in this report 
demonstrate the capability of the core to operate safely at the 
rated Watts Bar Unit 1 design thermal power with either a mixed core 
of RFA-2 fuel and the current fuel product or with a full core of 
RFA-2 fuel.
    6. TVA's amendment request establishes a reference upon which to 
base Westinghouse reload safety evaluations for future reloads with 
the RFA-2 fuel assembly design.
    7. Reload core designs with either a mixed core of RFA-2 fuel 
and the current fuel product or with a full core of RFA-2 fuel are 
compatible with the planned introduction of Tritium-Producing 
Burnable Absorber Rods (TPBARs) into Watts Bar Unit 1.
    These changes therefore do not cause the initiation of any 
accident nor create any new failure mechanisms. All equipment 
important to safety will operate as designed. Component integrity is 
not challenged. The changes do not result in any event previously 
deemed incredible being made credible. The use of RFA-2 is not 
expected to result in more adverse conditions and is not expected to 
result in any increase in the challenges to safety systems.
    Therefore, TVA concludes that this proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    No. The margin of safety is maintained by assuring compliance 
with acceptance limits reviewed and approved by the NRC. All of the 
appropriate acceptance criteria for the various analyses and 
evaluations have been met, therefore, there has not been a reduction 
in any margin of safety.
    Therefore, TVA concludes that 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide

[[Page 15766]]

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 e-mail to [email protected].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: April 10, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to relocate emergency diesel generator maintenance 
inspection requirements from Section 4.7 to the Updated Final Safety 
Analysis Report.
    Date of Issuance: March 7, 2003.
    Effective date: March 7, 2003 shall be implemented within 30 days 
of issuance, except the relocation of the emergency diesel generator 
maintenance requirements of Technical Specification 4.7, which shall be 
incorporated into the Updated Final Safety Analysis Report in 
accordance with the schedule specified by 10 CFR 50.71.
    Amendment No.: 236.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36926).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2003.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: August 30, 2002, as supplemented 
November 21 and December 16, 2002, and January 23, 2003.
    Brief description of amendment: This amendment revises the 
Technical Specifications by eliminating the requirements to perform 
response time testing for several reactor protection system and 
engineered safety feature functions in conformance with previously 
approved topical reports.
    Date of issuance: March 7, 2003.
    Effective date: March 7, 2003.
    Amendment No.: 112.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2002 (67 FR 
61676).
    The November 21 and December 16, 2002, and January 23, 2003, 
letters provided clarifying information and did not change the initial 
proposed no significant hazards consideration determination or expand 
the scope of the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2003.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1 (HNP), Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: August 28, 2002.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.9.9, ``Containment Ventilation Isolation 
System,'' to allow the same administrative controls for this TS as were 
approved previously by the NRC in Amendment No. 104 to the HNP TS for 
TS 3/4.9.4, ``Containment Building Penetrations,'' to provide 
consistency between the two TS.
    Date of issuance: March 12, 2003.
    Effective date: March 12, 2003.
    Amendment No.: 113.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61676).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendments: October 10, 2002, as 
supplemented by letters dated February 7 and February 26, 2003.
    Brief description of amendments: The amendment authorizes the 
licensee to continue to use, for operational cycle 13 beginning in 
March 2003, and subsequent cycles of operation, the reactor coolant 
system cold leg elbow tap flow coefficients that were approved by the 
NRC on an interim basis for cycle 12 in Amendment No. 186.
    Date of issuance: March 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 199.
    Facility Operating License No. NPF-52: Amendment authorizes 
revision of the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70765).
    The supplements dated February 7 and February 26, 2003, provided 
clarifying information that did not change the scope of the October 10, 
2002, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 19, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: May 29, 2002, as supplemented 
by letters dated September 25 and November 12, 2002, and January 8 and 
January 29, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow a one-time change in the Appendix J, 
Type A containment integrated leakage rate test interval from the 
currently required 10-year interval to a test interval of 15 years.
    Date of issuance: March 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 205/198.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45563).
    The supplements dated September 25 and November 12, 2002, and 
January 8 and January 29, 2003, provided clarifying information that 
did not change the scope of the May 29, 2002, application or the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: May 29, 2002, as supplemented 
by letters dated September 25 and

[[Page 15767]]

November 12, 2002, and January 8 and January 29, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow a one-time change in the Appendix J, 
Type A containment integrated leakage rate test interval from the 
currently required 10-year interval to a test interval of 15 years.
    Date of issuance: March 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 211/192.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45563).
    The supplements dated September 25 and November 12, 2002, and 
January 8 and January 29, 2003, provided clarifying information that 
did not change the scope of the May 29, 2002, application or the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 2003.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: April 24, 2002, as supplemented by 
letters dated July 18, December 18 and 20, 2002, and February 19, 2003.
    Brief description of amendment: The amendment reflects a full-scope 
implementation of the alternative source term, as described in 
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors,'' pursuant 
to 10 CFR 50.67, ``Accident source term.''
    Date of issuance: March 14, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 132.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40021).
    The July 18, December 18 and 20, 2002, and February 19, 2003, 
supplemental letters provided clarifying information that did not 
change the scope of the original Federal Register notice or the 
original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2003.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: May 14, 2002, as supplemented by letters 
dated February 12 and 28, 2003.
    Brief description of amendment: The amendment modifies the 
surveillance requirements (SRs) pertaining to the testing of the 
Division 3 standby emergency diesel generator (EDG). The change allows 
performance of some required surveillance tests for the Division 3 EDG 
during any mode of plant operation (previously allowed only in Modes 4 
(Cold Shutdown) and 5 (Refueling)).
    Date of issuance: March 14, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 133.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications and Surveillance Requirements.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42824).
    The February 12, 2003, supplemental letter provided clarifying 
information and the February 28, 2003, supplemental letter withdrew the 
requested change to the Note associated with SR 3.8.1.8. The 
supplemental letters did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 5, 2002, as supplemented on 
January 9 and March 4, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to implement the alternate source term methodology 
for the fuel-handling accident analysis. Specifically, the amendment 
revises TS 3.9.3, ``Containment Penetrations,'' to: (1) Permit the 
equipment closure hatch opening and the personnel airlock doors to be 
capable of being closed during movement of irradiated fuel, (2) allow 
use of administrative controls for unisolating containment penetrations 
during movement of irradiated fuel, (3) delete the containment purge 
and containment pressure relief requirements and associated 
surveillances with the reactor subcritical for less than 550 hours, and 
(4) eliminate the TS applicability ``during core alterations.'' In this 
regard, the amendment adopts TS Task Force (TSTF) Standard TS Change 
Travelers TSTF-68, ``Containment Personnel Airlock Doors Open During 
Fuel Movement,'' TSTF-312, ``Administratively Control Containment 
Penetrations,'' and, in part, TSTF-51, ``Revise Containment 
Requirements During Handling Irradiated Fuel and Core Alterations.'' 
The amendment also revises the Applicability Statements for Limiting 
Condition for Operation (LCO) 3.3.8 for the fuel storage building 
emergency ventilation system (FSBEVS) actuation instrumentation and LCO 
3.7.13 for the FSBEVS to also add the term ``recently'' before 
``irradiated fuel assemblies.'' In addition, the LCO Required Action 
would likewise be modified to add the term ``recently'' to now require 
the suspension of movement of recently irradiated fuel in the FSB.
    Date of issuance: March 17, 2003.
    Effective date: March 17, 2003.
    Amendment No.: 215.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45567).
    The January 9 and March 4 letters provided clarifying information 
that did not expand the scope of the proposed amendment or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 16, 2002.
    Brief description of amendment: The amendment relocates certain 
Control Rod Block functions from Technical Specifications 3/4.2.C, 
``Control Rod Block Actuation,'' Tables 3.2.C.1, 3.2.C-2, and 4.2.C to 
the Updated Final Safety Analysis Report.
    Date of issuance: March 17, 2003.

[[Page 15768]]

    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68735).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 24, 2002, as supplemented by 
letter dated February 4, 2003.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) relating to positive reactivity additions while in 
shutdown modes by clarifying TSs involving positive reactivity 
additions. In addition, the borated water volume requirements in TS 
3.1.2.7 is now presented in ``percent level'' units and an obsolete 
reference from Surveillance Requirement 4.8.2.2 is deleted.
    Date of issuance: March 7, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 185.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75874).
    The February 4, 2003, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 2, 2001, as supplemented by 
letters dated September 24, 2001, and February 27, July 31, and 
December 19, 2002.
    Brief description of amendment: The Refueling Water Storage Pool 
(RWSP) purification system is aligned to the RWSP to maintain the 
purity and clarity of the borated water contained in the pool. It is 
also one of two means of makeup to the Spent Fuel Pool, with the 
Condensate Storage Pool being the primary makeup source. Entergy 
Operations Inc. has proposed to revise its Waterford Steam Electric 
Station, Unit 3, Updated Final Safety Analysis Report (UFSAR) to allow 
the manual valves (FS-423 and FS-404) that isolate the RWSP from the 
RWSP purification system and provide the boundary between the 
seismically qualified, safety related RWSP and the non-seismic, non-
safety related RWSP purification system to be maintained open.
    Date of issuance: March 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 186.
    Facility Operating License No. NPF-38: The amendment revised the 
UFSAR.
    Date of initial notice in Federal Register: May 16, 2001 (66 FR 
27176).
    The September 24, 2001, and February 27, July 31, and December 19, 
2002, supplemental letters provided clarifying information that did not 
change the scope of the original Federal Register notice or the 
original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 27, 2002.
    Brief description of amendments: These amendments delete technical 
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminate the requirements to have and maintain the post accident 
sampling system at the Dresden Nuclear Power Station, Units 2 and 3. 
The amendments also address related changes to TS 5.5.2, ``Primary 
Coolant Sources Outside Containment.''
    Date of issuance: March 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 197/190.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2802).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: November 21, 2002, as 
supplemented February 25, 2003.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) for the safety limit for the minimum 
critical power ratio from its current value of 1.09 to 1.07 for two 
recirculation-loop operations, and from 1.11 to 1.09 for single 
recirculation-loop operation.
    Date of issuance: March 11, 2003.
    Effective date: As of the date of issuance, to be implemented prior 
to startup for Cycle 8 operations, scheduled for March 2003.
    Amendment No.: 127.
    Facility Operating License No. NPF-85: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
802). The February 25, 2003, letter provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on January 7, 2003 (68 FR 802).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: November 27, 2002.
    Brief description of amendments: These amendments delete technical 
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminate the requirements to have and maintain the post accident 
sampling system at the Quad Cities Nuclear Power Station, Units 1 and 
2. The amendments also address related changes to TS 5.5.2, ``Primary 
Coolant Sources Outside Containment.''
    Date of issuance: March 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 212/206.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2802)

[[Page 15769]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 14, 2002, as supplemented 
by letters dated July 17 and September 12, 2002, and January 24, 2003.
    Brief description of amendment: This amendment supplements License 
Amendment No. 100, which was issued on February 24, 1999, by placing 
restrictions on removing the inclined fuel transfer system (IFTS) blind 
flange during Operational Modes 1, 2, and 3. The amendment includes a 
time limit on the removal of the IFTS blind flange, provides a 
requirement to install the upper pool IFTS gate prior to IFTS blind 
flange removal, and limits the unbolted configuration of the IFTS blind 
flange when it is rotated.
    Date of issuance: March 7, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 123.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5675).
    The supplemental information contained 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 30, 2002.
    Brief description of amendment: This amendment deletes Technical 
Specification (TS) 5.5.3, ``Post Accident Sampling System (PASS),'' and 
thereby eliminates the requirements to have and maintain the PASS at 
the Perry Nuclear Power Plant, Unit 1. The amendment also addresses 
related changes to TS 5.5.2, ``Primary Coolant Sources Outside 
Containment.''
    Date of issuance: March 7, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 124.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2803).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2003.
    No significant hazards consideration comments received: No.

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear Station, 
Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: November 14, 2002, supplemented by a 
letter dated January 24, 2003, that supersedes previous applications 
dated August 9, 2000, June 13, 2002.
    Brief description of amendment request: The amendment revises TS 
6.5.4 and 6.5.3 to eliminate the requirements for the Independent 
Onsite Safety Review Group (IOSRG) which is not needed for safe 
monitoring of TMI-2 based on consideration that the reactor has been 
defueled to the extent reasonably achievable and the fuel shipped 
offsite. The amendment also revises TS 6.4 to delete the requirements 
for unit staff training that are outdated based on the adoption of a 
systems approach to training consistent with 10 CFR 50.120, ``Training 
and Qualification of Nuclear Power Plant Personnel.''
    Date of issuance: March 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 59.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50955).
    The November 14, 2002, application and supplemental letter dated 
January 24, 2003, replace in their entirety the previous applications 
dated August 9, 2000, June 13, 2002. The November 14, 2002, application 
supplemented by the January 24, 2003, letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated March 5, 2003.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of application for amendment: July 23, 2002, as supplemented 
November 15, 2002, and January 24, 2003.
    Brief description of amendment: The amendment revises the Unit 2 
reactor coolant system pressure-temperature curves in Technical 
Specification (TS) Figures 3.4-2 and 3.4-3 and associated TS Bases. The 
revised curves will bound operation of the unit for the remainder of 
its current license duration and bound operation with planned license 
amendments to increase the power level at which the unit is allowed to 
operate.
    Date of issuance: March 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from Unit 2 refueling outage 14.
    Amendment No.: 255.
    Facility Operating License No. DPR-74: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66010).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 20, 2003.
    No significant hazards consideration comments received: No.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: January 13, 2000, and 
supplemented by letters dated June 1, 2001, August 13, 2001, and 
October 15, 2002.
    Brief description of amendment: The amendment adds License 
Condition 2.B.(9) to the MY license. This new license condition 
incorporates the, Nuclear Regulatory Commission (NRC) approved, 
``License Termination Plan Rev 3.'' (LTP), and associated addendum, 
into the MY license and allows the licensee to make certain changes to 
the approved LTP without prior NRC review and approval.
    Date of issuance: February 28, 2003.
    Effective date: Date of issuance; to be implemented within [30] 
days from the date of issuance.
    Amendment No.: 168.

[[Page 15770]]

    Facility Operating License No. DPR-36: The amendment adds License 
Condition 2.B.(9).
    Date of initial notice in Federal Register: March 19, 2002.
    The supplemental letters provided additional clarifying 
information, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination published in the 
Federal Register on March 19, 2002.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation Report dated February 28, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 26, 2002.
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period 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 any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: March 6, 2003.
    Effective date: March 6, 2003.
    Amendment No.: 197.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75882).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 15, 2002, as supplemented by 
letter dated February 24, 2003.
    Brief description of amendment: The amendment revises the safety 
limit minimum critical power ratio values in Technical Specification 
2.1.1.2.
    Date of issuance: March 17, 2003.
    Effective date: March 17, 2003.
    Amendment No.: 198.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78521).
    The supplemental letter provided clarifying information that was 
within the scope of the original Federal Register Notice (67 FR 78521) 
and did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 26, 2002, as supplmented 
December 19, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 1.0, ``Definitions,'' TS 2.1, ``Safety Limits, 
Reactor Core,'' TS 2.3, ``Limiting Safety System Settings, Protective 
Instrumentation,'' TS 3.1, ``Reactor Coolant System,'' TS 3.8, 
``Refueling Operations,'' TS 3.10, ``Control Rod and Power Distribution 
Limits,'' TS 6.9, ``Reporting Requirements,'' and their associated 
Bases. These modifications allow the licensee to implement a Core 
Operating Limits Report (COLR) by relocating cycle-specific, reactor 
coolant system-related parameter limits from the TSs to the COLR. In 
addition, the amendment makes administrative changes to the above TSs.
    Date of issuance: March 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 165.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56322).
    The supplemental information dated December 19, 2002, contained 
clarifying information and did not change the scope of the July 26, 
2002, application nor the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 19, 2002, supplemented by 
letters dated September 13 and October 21, 2002.
    Brief description of amendment: The amendment revises the current 
radiological consequence analyses for the Kewaunee Nuclear Power Plant 
(KNPP) design-basis accidents to implement the alternate source term 
(AST) as described in Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors'' and Pursuant to 10 CFR 50.67, ``Accident 
Source Term.''
    Date of issuance: March 17, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 166.
    Facility Operating License No. DPR-43: Amendment revised the 
current radiological consequence analyses for the KNPP design-basis 
accidents to implement the AST.
    Date of initial notice in Federal Register: April 16, 2002 (67 FR 
18646).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2003.
    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: January 4, 2002, as supplemented 
January 9, 2003.
    Brief description of amendment: The amendment adds a limiting 
condition for operation of the mechanical vacuum pump instrumentation 
to trip the pumps on indication of high radiation levels in the main 
steam line and adds associated Surveillance Requirements.
    Date of issuance: March 11, 2003.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 143.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7421).
    The January 9, 2003, supplement contained clarifying information 
and

[[Page 15771]]

did not change the staff's proposed finding of no significant hazards 
consideration. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 11, 2003.
    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: August 20, 2002.
    Brief description of amendment: The amendment modifies the diesel 
generator action statements and surveillance requirements defined in 
the plant's Technical Specifications, in order to reduce degradation of 
the diesel generators associated with fast starting and rapid loading.
    Date of issuance: March 17, 2003.
    Effective date: March 17, 2003.
    Amendment No.: 144.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61684).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2003.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: March 20, 2002.
    Brief description of amendment: The proposed amendment would change 
Technical Specification (TS) Section 1.10, ``Definitions, Dose 
Equivalent I-131,'' to allow the use of the thyroid dose conversion 
factors listed in the International Commission on Radiological 
Protection Publication No. 30 (ICRP-30), ``Limits for Intakes of 
Radionuclides by Workers,'' 1979, in determining the iodine-131 dose 
equivalent reactor coolant activity in TS Section 3/4.4.8 and in 
calculating the radiological consequences from postulated design basis 
accidents.
    Date of issuance: March 6, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 162.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53991).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2003.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: August 9 2002, as supplemented 
by letters dated January 8 and February 6, 2003.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report to incorporate the Boiling Water Reactor 
Vessel and Internals Project Integrated Surveillance for the 
surveillance of the material capsules.
    Date of issuance: March 10, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 237 and 179.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61684).
    The supplements dated January 8 and February 6, 2003, provided 
clarifying information that did not change the scope of the August 9, 
2002, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 10, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 24th day of March, 2003.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-7489 Filed 3-31-03; 8:45 am]
BILLING CODE 7590-01-P