[Federal Register Volume 68, Number 150 (Tuesday, August 5, 2003)]
[Notices]
[Pages 46239-46251]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-19487]


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NUCLEAR REGULATORY COMMISSION


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

Background

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

[[Page 46240]]

proposed to be issued from, July 11, 2003, through July 24, 2003. The 
last biweekly notice was published on July 22, 2003 (68 FR 43382).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 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 September 4, 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, 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.
    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

[[Page 46241]]

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]. 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 Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: July 21, 2003.
    Description of amendments request: The proposed license amendment 
requests approval to revise the Updated Final Safety Analysis Report, 
Section 9.4.5, ``Turbine Building Ventilation System,'' and supporting 
information in Section 6.4.4.1, ``Radiological Protection,'' and 
Section 15.6.3, ``Main Steam Line Break Accident,'' to allow the system 
to be operated in a once-through versus recirculation configuration in 
support of outage activities.
    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 accident of concern for the proposed modification is a Main 
Steam Line Break (MSLB). The probability of this event is not 
impacted by the change to the turbine building ventilation system 
configuration. The consequences of the event have been re-evaluated 
to determine the impact on control room operator doses and offsite 
doses. The re-evaluation was performed consistent with the analysis 
done in support of the adoption of Alternative Source Term (AST) 
which was approved for use at BSEP [Brunswick Steam Electric Plant] 
in Amendments 221 and 246 for Units 1 and 2, respectively. The 
results of the re-evaluation demonstrate that control room doses 
remain well below regulatory limits and [that] there is no 
significant impact on offsite doses. 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 turbine building ventilation system is non-safety related 
and its purpose is to provide an acceptable environment for 
equipment and personnel within the turbine building as well as treat 
the gaseous effluent prior to release. As such, modification of this 
system cannot (1) Alter any design basis accident initiators, (2) 
create new types of accident precursors, or (3) introduce new 
failure modes of safety related equipment. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety for this modification is considered to be 
that provided by meeting the applicable regulatory limits. Operation 
of the turbine building ventilation system in a once-through versus 
recirculation configuration does not impact the ability to ensure 
that the doses at the exclusion area and low population zone 
boundaries, as well as the control room, remain well within 
corresponding regulatory limits with respect to a MSLB event (i.e., 
the only event whose consequences can be impacted by the proposed 
modification). This was confirmed through re-evaluation of the 
consequences of a MSLB event, consistent with the analysis done in 
support of the adoption of AST. Since the proposed changes continue 
to ensure that the doses at the exclusion area and low population 
zone boundaries, as well as the control room are within 
corresponding regulatory limits, the proposed license amendment 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: April 25, 2003, as supplemented on May 
21 and June 11, 2003.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) consists of revisions to protective 
instrumentation specifications. These changes are made to resolve non-
conservative TS issues, relax overly restrictive requirements, and to 
provide consistency between TS and design and licensing bases. These 
changes also involve reformatting data, as well as relocation of some 
data to plant-controlled documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's review is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 46242]]

    The proposed changes do not significantly affect the design or 
fundamental operation and maintenance of the plant. Accident initiators 
or the frequency of analyzed accident events are not significantly 
affected as a result of the proposed changes; therefore, there will be 
no significant change to the probabilities of accidents previously 
evaluated.
    The proposed changes do not significantly alter assumptions or 
initial conditions relative to the mitigation of an accident previously 
evaluated. The proposed changes continue to ensure process variables, 
structures, systems, and components (SSCs) are maintained consistent 
with the safety analyses and licensing basis. The revised TSs continue 
to require that SSCs are properly maintained to ensure operability and 
performance of safety functions as assumed in the safety analyses. The 
design basis events analyzed in the safety analyses will not change 
significantly as a result of the proposed changes to the TSs.
    Therefore, the proposed changes do 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 accident previously evaluated.
    The proposed changes do not involve any physical alteration of the 
plant (no new or different type of equipment being installed) and do 
not involve a significant change in the design, normal configuration or 
basic operation of the plant. The proposed changes do not introduce any 
new accident initiators. In some cases, the proposed changes impose 
different requirements; however, these new requirements are consistent 
with the assumptions in the safety analyses and current licensing 
basis. Where requirements are relocated to other licensee-controlled 
documents, adequate controls exist to ensure their proper maintenance.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Margin of safety is related to the confidence in the ability of the 
fission product barriers to perform their design functions during and 
following an accident situation. The proposed changes do not 
significantly affect any of the assumptions, initial conditions or 
inputs to the safety analyses. Plant design is unaffected by these 
proposed changes and will continue to provide adequate defense-in-depth 
and diversity of safety functions as assumed in the safety analyses.
    There is no proposed change to Safety Limits and only 
administrative and more restrictive changes to Limiting Safety System 
Setting requirements. The proposed changes maintain requirements 
consistent with safety analyses assumptions and the licensing basis. 
Fission product barriers will continue to meet their design 
capabilities without significant impact to their ability to maintain 
parameters within acceptable limits. The safety functions are 
maintained within acceptable limits without any significant decrease in 
margin.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: February 27, 2003. By a letter dated 
July 17, 2003, the licensee revised its analysis about the issue of no 
significant hazards consideration.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Section 3.4.9, Reactor Coolant 
System Pressure and Temperature (P/T) Limits, and delete the license 
conditions specified in Facility Operating License Sections 2.C(8) and 
3.P, Pressure-Temperature Limit Curves, for Dresden Nuclear Power 
Station, Units 2 and 3 respectively. The P/T limit curves are proposed 
to be replaced with ones that are applicable to the remainder of the 
licensed life of the plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The P/T limits are prescribed during all operational conditions 
to avoid encountering pressure, temperature, and temperature rate of 
change conditions that might cause undetected flaws to propagate, 
resulting in non-ductile failure of the reactor coolant pressure 
boundary, which is an unanalyzed condition. The methodology used to 
determine the P/T limits has been approved by the NRC and thus is an 
acceptable method for determining these limits. Therefore, the 
proposed changes do not affect the probability of an accident 
previously evaluated.
    There is no specific accident that postulates a non-ductile 
failure of the reactor coolant pressure boundary. The loss of 
coolant accident analyzed for the plant assumes a complete break of 
the reactor coolant pressure boundary. The revision to the P/T 
limits does not change this assumption. 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 TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not 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.
    Non-ductile failure of the reactor coolant pressure boundary is 
not an analyzed accident. The proposed changes to the P/T limits 
were developed using an NRC-approved methodology, and thus the 
revised limits will continue to provide protection against non-
ductile failure of the reactor coolant pressure boundary.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety related to the proposed changes is the 
margin between the proposed P/T limits and the pressures and 
temperatures that would produce non-ductile failure of the reactor 
coolant pressure boundary. The use of an NRC-approved methodology 
together with conservatively-chosen plant-specific input parameters 
provides an acceptable 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.

[[Page 46243]]

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station, Unit 3, York County and 
Lancaster County, Pennsylvania

    Date of amendment request: June 23, 2003.
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Peach Bottom Atomic Power 
Station (PBAPS), Unit 3, Technical Specifications (TSs) contained in 
Appendix A to the Operating License. This proposed change will revise 
the TS section on safety limits to incorporate revised safety limit 
minimum critical power ratios (SLMCPRs) based on the cycle-specific 
analysis performed by Global Nuclear Fuel for PBAPS, Unit 3, Cycle 15.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The U.S. Nuclear Regulatory Commission (NRC) staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff's review is presented below.
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Changing the SLMCPRs does not require any physical plant 
modifications, physically affect any plant components, or involve 
changes in plant operation. Therefore, the probability of an accident 
previously evaluated remains unchanged.
    The operability of plant systems designed to mitigate any 
consequences of accidents has not changed, therefore, the consequences 
of an accident previously evaluated are not expected to increase.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve any modifications of the plant 
configuration for allowable modes of operation. The SLMCPRs are not 
accident initiators, and their revision will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    The proposed SLMCPRs provide a margin of safety by ensuring that no 
more than 0.1% of the rods are in a boiling transition if the operating 
limit minimum critical power ratios are violated during all modes of 
operation. The change in the SLMCPRs continues to ensure that during 
normal operation and during abnormal operational transients, at least 
99.9% of all fuel rods in the core do not experience transition boiling 
if the limit is not violated when all uncertainties are considered, 
thereby preserving the fuel cladding integrity. Therefore, the proposed 
TS change will not involve a significant reduction in a margin of 
safety.
    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of amendment request: June 24, 2003.
    Description of amendment request: The proposed amendments would 
change the Technical Specification (TS) steam generator tube inspection 
definition such that the definition of tube inspection would exclude 
the portion of the tube within the tubesheet below the W* distance and 
would change the tube plugging criteria to indicate that the plugging 
or repair criteria does not apply to service-induced degradation 
identified in the W* distance. Service-induced degradation identified 
in the W* distance would be repaired upon detection. The W* distance is 
defined in Westinghouse Topical Report, WCAP-14797, Revision 1, and is 
the distance from the top of the tubesheet to the bottom of the W* 
length including the distance to the bottom of the WEXTEX transition 
(approximately 0.25 inches from the top of the tubesheet) plus 
uncertainties. This equals approximately 7.12 inches on the hot leg 
side plus the distance to the bottom of the WEXTEX transition and 7.62 
inches on the cold leg side plus the distance to the bottom of the 
WEXTEX transition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change modifies the BVPS Unit 1 TSs to 
incorporate [an] SG [steam generator] tube inspection scope based on 
WCAP-14797, Revision 1. The proposed change only clarifies the 
current process which has been utilized in the past. The W* analysis 
takes into account the reinforcing effect that the tubesheet has on 
the external surface of an expanded SG tube. Tube-bundle integrity 
will not be adversely affected by the implementation of the W* tube 
inspection scope. SG tube burst or collapse cannot occur within the 
confines of the tubesheet; therefore, the tube burst and collapse 
criteria of Regulatory Guide (RG) 1.121 are inherently met. Any 
degradation below the W* distance is shown by analysis and test 
results to be acceptable, and therefore does not result in an 
increase in probability of a tube rupture or an increase in the 
consequences of a tube rupture.
    Tube burst is precluded for cracks within the tubesheet by the 
constraint provided by the tubesheet. However, in the unlikely event 
of a complete circumferential separation of a tube occurring below 
the W* distance, SG tube pullout is precluded, tube integrity is 
maintained and leakage is predicted to be maintained within the 
Updated Final Safety Analysis Report limits during all plant 
conditions.
    In conclusion, the incorporation of the W* inspection scope into 
BVPS Unit 1 TS[s] maintains existing design limits and does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change modifies the BVPS Unit 1 TSs to 
incorporate SG tube inspection scope based on WCAP-14797, Revision 
1. Tube-bundle integrity will be maintained during all plant 
conditions upon implementation of the proposed tube inspection 
scope. Use of this scope does not induce a new mechanism that would 
result in a different kind of accident from those previously 
analyzed. Even with the limiting circumstances of a complete 
circumferential separation of a tube occurring below the W* 
distance, SG tube pullout is precluded and leakage is predicted to 
be maintained within the design limits during all plant conditions.
    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?
    No. WCAP-14797, Revision 1 describes the testing that was 
performed to define the length of non-degraded tubing that is 
sufficient to compensate for the axial forces on the tube and thus 
prevent pullout. The operating conditions utilized in WCAP-14797, 
Revision 1, bound BVPS Unit 1 operating conditions. Upon 
implementation of the W* inspection scope, operation with potential 
cracking below the W* distance in

[[Page 46244]]

the WESTEX expansion region of the SG tubing meets the margin of 
safety as defined in RG 1.121 and RG 1.83 and the requirements of 
General Design Criteria 14, 15, 16, 31, and 32.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida.

    Date of amendment request: July 14, 2003.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 3.7.9 by adding a note to allow a 
one-time 10-day completion time for restoring an inoperable nuclear 
services seawater system train to operable status. The proposed change 
would allow the refurbishment of one nuclear services seawater system 
emergency pump (RWP-2A or RWP-2B) online. The note would specify that 
the one-time 10-day completion time will expire on December 30, 2004.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Nuclear Services 
Seawater System train to Operable status. The Nuclear Services 
Seawater System is designed to provide cooling for components 
essential to the mitigation of plant transients and accidents. The 
system is not an initiator of design basis accidents. During the 
requested extended time period of ten days, the redundant Emergency 
Nuclear Services Seawater pump will be available and capable of 
providing cooling for containment heat loads and essential equipment 
during emergency conditions. RWP-1 is the CR-3 [Crystal River Unit 
3] normal duty Nuclear Closed Cycle Cooling Water pump. Although 
RWP-1 is non-safety related and its motor is non-seismic, has a 
lower flow capability than either RWP-2B or RWP-2A and is not 
connected to an emergency power source, it will also be available 
and capable of removing emergency heat loads from essential 
equipment from all design basis events. Informal calculations 
performed show that below a Ultimate Heat Sink (UHS) temperature of 
approximately 90[deg]F, RWP-1 can maintain adequate heat removal 
under accident conditions.
    A Probabilistic Safety Assessment (PSA) has been performed to 
assess the risk impact of an increase in Completion Time. Although 
the proposed one-time change results in an increase in Core Damage 
Frequency (CDF) and Large Early Release Frequency (LERF), the value 
of these increases are considered as very small in the current 
regulatory guidance.
    Therefore, granting this LAR [License Amendment Request] does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Nuclear Services 
Seawater System train to Operable status.
    The proposed LAR will not result in changes to the design, 
physical configuration of the plant or the assumptions made in the 
safety analysis. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed license amendment extends, on a one-time basis, the 
Completion Time for restoring an inoperable Nuclear Services 
Seawater System train to Operable status. The proposed change will 
allow online repair of one of the Emergency Nuclear Services 
Seawater pumps to improve its reliability and useful lifetime, thus 
increasing the long term margin of safety of the system.
    The proposed LAR will reduce the probability (and associated 
risk) of a plant shutdown to repair an Emergency Nuclear Services 
Seawater pump. To ensure defense in depth capabilities and the 
assumptions in the risk assessment are maintained during the 
proposed one-time extended Completion Time, CR-3 will continue the 
performance of 10 CFR 50.65(a)(4) assessments before performing 
maintenance or surveillance activities and no maintenance activities 
of other risk sensitive equipment beyond that required for the 
refurbishment activity will be scheduled concurrent with the repair 
activity. Other compensatory actions that may be implemented, 
include: Use of pre-job briefings and periodic operator walkdowns to 
assess status of risk sensitive equipment in the redundant train, 
selection of beneficial Makeup Pump configurations and redundant 
off-site power feeds to the remaining Emergency Nuclear Services 
Seawater System pump, no elective maintenance to be scheduled in the 
switchyard, and the establishment of fire watches in fire areas 
identified in [PSA Risk Assessment of RWP-2A/2B Extended AOT 
[Allowed Outage Time]].
    As described above in Item 1, a PSA has been performed to assess 
the risk impact of an increase in Completion Time. Although the 
proposed one-time change results in an increase in Core Damage 
Frequency (CDF), and Large Early Release Frequency, the value of 
these increases are considered as very small in the current 
regulatory guidance.
    Therefore, granting this LAR does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: July 7, 2003.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specification (TS) 
Section 3.3.e, ``Service Water System,'' to add requirements for the 
turbine building service water header isolation logic.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The service water system and specifically the supply to the 
turbine building, does not initiate any accidents previously 
evaluated. This change will provide an automatic feature to a 
function that was previously available to operators, to ensure 
Emergency Safety Features (ESF) loads will receive adequate service 
water flow. Flow is provided to ESF components that are cooled by 
service water without relying on the operator to identify and take 
action to provide isolation. Diesel loading and sequencing will not 
be adversely affected by this change. The components supplied by the 
service water system will continue to be supplied in a timely 
manner. The valve logic will be properly calibrated and tested 
consistent with other valves associated with safety significant 
structures, systems and components.
    Therefore, the proposed change will not increase the probability 
or consequences of an accident previously evaluated.

[[Page 46245]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change will not affect the service water system function or 
any components that are accident initiators. The ability to isolate 
the turbine building load in the event of a system malfunction has 
been previously evaluated.
    Therefore, any change to the system would not affect the 
probability of an accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    This change will ensure that Engineered Safety Features (ESF) 
components receiving service water-cooling are not negatively 
impacted by turbine building load. There are no components served by 
the turbine building header that are safety systems, structures, or 
components.
    Therefore, NMC concludes that there is not a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: July 1, 2003.
    Description of amendment request: The proposed amendment would 
change the Unit 1 Technical Specifications (TSs) by including the Unit 
1 Cycle 14 (U1C14) Minimum Critical Power Ratio (MCPR) Safety Limits in 
Section 2.1.1.2, changing the references listed in Section 5.6.5.b, and 
changing the design features in Section 4.2.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change to the MCPR Safety Limits does not directly 
or indirectly affect any plant system, equipment, component, or 
change the processes used to operate the plant. Further, the U1C14 
MCPR Safety Limits are generated using NRC approved methodology and 
meet the applicable acceptance criteria. Thus, this proposed 
amendment does not involve a significant increase in the probability 
of occurrence of an accident previously evaluated.
    Prior to the startup of U1C14, licensing analyses are performed 
(using NRC approved methodology referenced in Technical 
Specification Section 5.6.5.b) to determine changes in the critical 
power ratio as a result of anticipated operational occurrences. 
These results are added to the MCPR Safety Limit values proposed 
herein to generate the MCPR operating limits in the U1C14 COLR [Core 
Operating Limits Report]. These limits could be different from those 
specified for the U1C13 COLR. The COLR operating limits thus assure 
that the MCPR Safety Limit will not be exceeded during normal 
operation or anticipated operational occurrences. Postulated 
accidents are also analyzed prior to startup of U1C14 and the 
results shown to be within the NRC approved criteria.
    The U1C14 reload fuel bundles will utilize a small amount of 
depleted uranium in certain fuel rods, in addition to natural and 
slightly enriched uranium. There is no change to the composition of 
the fuel pellets containing depleted uranium material (i.e., 
UO2) except a slight decrease in the amount of Uranium-
235. Therefore, the use of depleted uranium in the fuel rods does 
not affect the mechanical performance of the fuel rods. The depleted 
uranium was modeled in the approved design and licensing 
methodology.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U1C14 core operating limits. The use of this approved methodology 
does not increase the probability of occurrence or consequences of 
an accident previously evaluated.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change to the MCPR Safety Limits does not directly or 
indirectly affect any plant system, equipment, or component and 
therefore does not affect the failure modes of any of these items. 
Thus, the proposed changes do not create the possibility of a 
previously unevaluated operator error or a new single failure.
    The use of depleted uranium in the fuel rods does not affect the 
mechanical performance of the fuel rods.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U1C14 core operating limits. The use of this approved methodology 
does not create the possibility of a new or different kind of 
accident.
    Therefore, the proposed amendment does not involve the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since the proposed changes do not alter any plant system, 
equipment, component, or the processes used to operate the plant, 
the proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by Technical 
Specifications. The proposed MCPR Safety Limits do not involve a 
significant reduction in the margin of safety as currently defined 
in the Bases of the applicable Technical Specification sections, 
because the MCPR Safety Limits calculated for U1C14 preserve the 
required margin of safety.
    The use of depleted uranium in the fuel rods does not affect the 
mechanical performance of the fuel rods.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U1C14 core operating limits. This approved methodology is used to 
demonstrate that all applicable criteria are met, thus, 
demonstrating that there is no reduction in the margin of safety.
    Therefore, these 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: July 9, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.2, to increase the allowed 
outage time (AOT) for one train of the control room emergency 
filtration (CREF) system from 7 days to 30 days. The proposed AOT 
change would only apply when one CREF train is inoperable due to an 
inoperable chiller during Modes 1, 2, or 3.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) section 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?
    Response: No.

[[Page 46246]]

    The proposed TS change does not affect the design, operational 
characteristics, function or reliability of the control room 
emergency filtration (CREF) system. The CREF is not an initiator of 
any previously evaluated accident. The proposed change will increase 
the allowed outage time for the chiller from seven days to 30 days 
for the chiller in OPERATIONAL CONDITIONS 1, 2, AND 3. The 30-day 
AOT is based on the low probability of an event requiring control 
room isolation concurrent with failure of the redundant train. 
Therefore, one train will always be available to remove the normal 
and accident heat loads and provide control room isolation.
    Increasing the AOT will allow for completion of maintenance 
activities requiring extended down time to perform and result in 
significant improvements to the overall reliability of control room 
chillers. Improving reliability will provide additional assurance 
that chillers will be capable of performing their design basis 
accident function.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
radiological consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will increase the AOT for the control room 
chiller from seven to thirty days in modes 1 through 3. During the 
time one chiller is inoperable, the redundant train is capable of 
handling the heat loads during normal operation and accident 
conditions. The proposed change does not involve a change in the 
design, configuration, or method of operation of the plant that 
could create the possibility of a new or different kind of accident. 
The proposed change would not introduce new failure modes or effects 
and would not, in the absence of other unrelated failures, create a 
new or different accident from any accidents previously evaluated.
    Therefore, the proposed changes would not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    The basis for technical specification 3/4.7.2 is to ensure that 
the temperature in the control room does not exceed the maximum 
allowable for the equipment and instrumentation located therein. The 
system also limits radiation exposure to control room personnel 
following an accident to below GDC-19 [General Design Criterion 19] 
limits. Either of the two redundant trains can perform these 
functions. Although one chiller may be inoperable for longer than 
seven days, the redundant train can perform all normal and accident 
functions. The length of time for the chiller AOT is sufficiently 
short to assure that an event requiring control room isolation 
concurrent with the failure of the redundant train is not credible.
    Therefore, these changes do not involve a significant reduction 
in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: June 6, 2003.
    Description of amendment request: The proposed amendment would 
revise the Salem Nuclear Generating Station, Unit Nos. 1 and 2, 
Technical Specification (TS) 3/4.11.2.5, ``Explosive Gas Mixture.'' The 
proposed changes would: (1) Add a footnote to Limiting Condition for 
Operation (LCO) 3.11.2.5, to allow maintenance on the waste gas system; 
(2) revise Surveillance Requirement 4.11.2.5, to delete reference to 
hydrogen which is not limited by the LCO; and (3) incorporate changes 
to the appropriate TS Bases pages.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications (TS) 3/
4.11.2.5, Explosive Gas Mixtures, would correct inconsistencies 
while continuing to preclude the combination of explosive 
concentrations of oxygen and hydrogen in the Salem Generating 
Station (SGS) Unit 1 and 2 waste gas system. The changes eliminate 
the potential for misinterpretation and achieve internal consistency 
between TS sections. No changes to the design of structures, 
systems, or components (SSC) are made and there are no effects on 
accident mitigation.
    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.
    Section 15.3.6 of the SGS Updated Final Safety Analysis Report 
(UFSAR) summarizes the results of a postulated non-mechanistic 
rupture of a waste gas decay tank. This postulated accident scenario 
is not affected by the proposed amendment, nor is any new accident 
scenario introduced by the proposed changes. The proposed 
administrative and editorial changes to the TS do not change the 
design function of or operation of any SSCs. The TS, as amended, 
would continue to limit explosive and flammable gas concentrations 
to prevent an uncontrolled release from the waste gas system.
    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?
    Response: No.
    The proposed changes [ ] do not affect the ability of plant SSCs 
to perform their design basis accident functions. In addition, the 
[proposed TS license amendment] does not change the margin of safety 
since no SSCs are changed and the [current] limits on explosive gas 
mixtures are maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: July 10, 2003.
    Brief description of amendments: The proposed amendments would 
provide for a one-time change for each unit to revise Technical 
Specification 3.7.10, entitled ``Control Room Emergency Filtration/
Pressurization System (CREFS),'' to extend the COMPLETION TIME for 
ACTION B from 24 hours to 14 days.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?

[[Page 46247]]

    Response: No.
    This is a revision to the Technical Specifications for the 
control room emergency/filtration system which is a mitigation 
system designed to minimize in leakage and to filter the control 
room atmosphere to protect the operator following accidents 
previously analyzed. An important part of the system is the control 
room boundary. The control room boundary integrity is not an 
initiator or precursor to any accident previously evaluated. 
Therefore, the probability of any accident previously evaluated is 
not increased. The analysis of the consequences of analyzed accident 
scenarios under the control room breach conditions along with the 
compensatory actions for restoration of control room integrity 
demonstrate that the consequences of any accident previously 
evaluated are not increased. Therefore, it is concluded that this 
change does not significantly increase the probability 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.
    The proposed changes will not impact the accident analysis. The 
changes will not alter the requirements of the control room 
emergency/filtration system or its function during accident 
conditions. The administrative controls and compensatory actions 
will ensure the control room emergency/filtration system will 
perform its safety function. [Sentence deleted] The changes do not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis for an unacceptable period of time without compensatory 
actions and administrative controls. The proposed changes do not 
affect systems that respond to safely shutdown the plant and to 
maintain the plant in a safe shutdown condition. 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of amendment request: September 5, 2002, as supplemented April 
16, June 9, and July 7, 2003.
    Description of amendment request: The proposed technical 
specification (TS) amendment will add provisions to permit inspection 
and related repair of a buried fuel oil storage tank during plant 
operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    This proposed TS change does not alter the assumptions of the 
accident analyses or the TS Basis. The inclusion of provisions to 
permit inspection and related repair of a buried fuel oil storage 
tank during plant operation does not impact the availability of the 
EDGs [emergency diesel generators] to perform their required 
function, which is to provide an emergency source of power to vital 
equipment when a normal power source is not available. Furthermore, 
while a buried tank is out of service, the proposed change includes 
requirements to verify the availability of onsite and offsite fuel 
oil sources to ensure that an adequate supply of fuel oil remains 
available. Therefore, the proposed change does not result in a 
significant increase in either the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    This proposed TS change does not involve a physical change to 
the plant, nor does it alter the assumptions of the accident 
analyses. Inclusion of provisions to permit inspection and related 
repair of a buried fuel oil storage tank does not introduce any new 
failure modes. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from those 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    This proposed TS change alters the method of operation of the 
Fuel Oil System. However, the availability of the EDGs to perform 
their required function is not impacted, and the assumptions of the 
accident analyses are not altered. Furthermore, a plant specific 
risk evaluation of the acceptability of the provisions was 
performed. The risk evaluation concluded that the risk impact is 
acceptable (i.e., is characterized as ``very small'' by Regulatory 
Guide 1.174 criteria and is within the acceptance criteria of 
Regulatory Guide 1.177). Therefore, the proposed change does 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 request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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

[[Page 46248]]

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 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].

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

    Date of application for amendments: August 28, 2002.
    Brief description of amendments: The amendments extend the 
expiration date of the operating licenses from December 31, 2024, to 
June 1, 2025, for Unit 1, December 9, 2025, to April 24, 2026, for Unit 
2, and March 25, 2027, to November 25, 2027, for Unit 3 of Palo Verde 
Nuclear Generating Station.
    Date of Issuance: July 15, 2003.
    Effective date: July 15, 2003, and shall be implemented within 60 
days of the date of issuance.
    Amendment Nos.: Unit 1-147, Unit 2--147, Unit 3--147.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63688). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 11, 2002, as supplemented 
May 2, 2003, and June 23, 2003.
    Brief description of amendments: These amendments revise the CCNPP 
Technical Specification Administrative Controls Section to incorporate 
six changes previously approved for the Improved Standard Technical 
Specifications and one administrative change in renumbering pages.
    Date of issuance: July 16, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 259 and 236.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56318). The May 9, 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 initial proposed no 
significant hazards consideration determination as published in the 
Federal Register on September 3, 2002 (67 FR 56318). The June 23, 2003, 
letter withdrew the requested change dealing with clarifying references 
to 10 CFR part 20 in the Technical Specifications and did not change 
the initial proposed no significant hazards consideration determination 
as published in the Federal Register on September 3, 2002 (67 FR 
56318). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated July 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 13, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.10, ``Technical Specifications (TS) Bases 
Control Program,'' to provide consistency with the changes to 10 CFR 
50.59, which were published in the Federal Register (64 FR 53582) on 
October 4, 1999, and became effective March 13, 2001. Specifically, TS 
5.5.10 has been revised to remove the phrase ``unreviewed safety 
question.''
    Date of issuance: July 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 156.
    Facility Operating License No. NPF-43: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28848). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 22, 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: March 20, 2002, as supplemented by 
letter dated May 28, 2003.
    Brief description of amendment: The amendment revises the reporting 
requirements specified in Section 2.E of the Facility Operating License 
and Technical Specification Section 5.6.4 by eliminating requirements 
that provide the U.S. Nuclear Regulatory Commission with information 
that is not risk significant, and change the reporting time period to 
be consistent with Section 50.73 of Title 10 of the Code of Federal 
Regulations.
    Date of issuance: July 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 135.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21286). The May 28, 2003, supplemental letter withdrew a portion of the 
original amendment request, but did not expand the scope of the 
original Federal Register notice or change the proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 16, 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: August 15, 2002, as supplemented by 
letter dated May 9, 2003.
    Brief description of amendment: The amendment revises the reactor 
vessel surveillance program required by Title 10 of the Code of Federal 
Regulations, part 50, appendix H, section IIIB.3, allowing River Bend 
Station to incorporate the Boiling Water Reactor Vessel Internals 
Project Integrated Surveillance Program into the licensing basis.
    Date of issuance: July 24, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 136.
    Facility Operating License No. NPF-47: The amendment consists of 
NRC staff approval of changes to the Updated Safety Analysis Report.

[[Page 46249]]

    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61679). The May 9, 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 July 24, 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: July 5, 2002, as supplemented 
August 13, 2002.
    Brief description of amendment: The amendment relocates portions of 
Technical Specification (TS) 3/4.6.B, ``Primary System Boundary--
Coolant Chemistry,'' from the TSs to the Updated Final Safety Analysis 
Report (UFSAR). The portions of the TSs relocated to the UFSAR are the 
reactor coolant chemistry requirements for conductivity and chloride 
concentration. Specifically, TSs 3/4.6.B.2, 3/4.6.B.3, and 3.6.B.4 are 
relocated to the UFSAR.
    Date of issuance: July 21, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28850). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 21, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 31, 2001, as 
supplemented by letters dated December 21, 2001, and February 4, May 
31, and December 2, 2002.
    Brief description of amendments: The changes relocated the pressure 
temperature (P/T) limit curves and low temperature overpressure 
protection system limits to the Pressure and Temperature Limits Report 
(PTLR) in the BVPS-1 and 2 Licensing Requirements Manual and the 
reference that report in the affected TS limiting conditions for 
operation and Bases. The changes also included the addition of the PTLR 
to the Definitions Section of the TSs and added a new section to the 
reporting requirements in the Administrative Controls Section of the 
TSs delineating the necessary reports. The proposed changes were based 
on Generic Letter 96-03, ``Relocation of the Pressure Temperature Limit 
Curves and Low Temperature Overpressure Protection System Limits,'' 
dated January 31, 1996, and the Nuclear Regulatory Commission (NRC) 
staff's approval of the BVPS-1 and 2 plant-specific P/T limits 
methodology documented in the letter from Richard J. Laufer, NRC, to 
Mark B. Bezilla, FENOC, dated October 8, 2002.
    Date of issuance: July 15, 2003.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 256 and 138.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66465). The supplements dated December 21, 2001, and February 4, May 
31, and December 2, 2002, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated July 15, 2003.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: October 11, 2002, as 
supplemented March 4, 2003.
    Brief description of amendment: The amendment revises Crystal River 
Unit 3 Improved Technical Specifications (ITS) 3.3.15, ``Reactor 
Building Purge Isolation-High Radiation''; ITS Bases 3.7.15, ``Spent 
Fuel Assembly Storage''; ITS 3.9.3, ``Containment Penetrations''; and 
ITS 3.9.6, ``Refueling Canal Water Level'' to account for the handling 
of irradiated fuel within containment that has not occupied part of a 
critical reactor core within the previous 72 hours.
    Date of issuance: July 14, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented prior to entering Mode 6 for the Cycle 13 refueling outage.
    Amendment No.: 208.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 ( 68 
FR 5676). The March 4, 2003, supplement contained clarifying 
information only and did not change the initial 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 July 14, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 10, 2002.
    Brief description of amendment: The amendment consists of changes 
to the Donald C. Cook Nuclear Plant (D. C. Cook) Unit 1 Technical 
Specifications related to the reactor pressure vessel (RPV) operating 
limits at low temperatures. The amendment approves revised pressure-
temperature limits for the RPV to be applicable for a maximum of 32 
effective full-power years of facility operation. These changes were 
based, in part, on the use of American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code (Code) Case N-641.
    Date of issuance: July 18, 2003.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from Unit 1 refueling outage 19.
    Amendment No.: 278.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15762). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 18, 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: April 24, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specifications to eliminate the requirement for at least one person 
qualified to stand watch being present in the control room when 
irradiated fuel is stored in the fuel storage pool.

[[Page 46250]]

    Date of issuance: July 02, 2003.
    Effective date: Date of issuance to be implemented within [30] days 
from the date of issuance.
    Amendment No.: 169.
    Facility Operating License No. DPR-36: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28854). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 02, 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: October 8, 2002.
    Brief description of amendment: The amendment changes the title of 
Shift Supervisor to Shift Manager. This amendment also replaces plant-
specific titles with generic titles consistent with Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler TSTF-65, Rev. 1.
    Date of issuance: July 15, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 200.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28854). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: March 19, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 5.3, ``Plant Staff Qualifications.'' The 
amendments update requirements that have been outdated based on 
licensed operator training programs being accredited by the National 
Academy for Nuclear Training and promulgation of the revised Title 10 
of the Code of Federal Regulations, part 55, ``Operators' Licenses.''
    Date of issuance: July 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 159 and 150.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18281). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 22, 2003.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: April 10, 2002.
    Brief description of amendments: The license amendments revised 
several required actions in the Diablo Canyon Nuclear Power Plant 
(DCPP) Technical Specifications (TSs) that require suspension of 
operations involving positive reactivity additions or suspension of 
operations involving reactor coolant system (RCS) boron concentration 
reductions. In addition, the amendments revised several Limiting 
Condition for Operation notes that preclude reductions in RCS boron 
concentration when a reactor coolant pump(s) and/or a residual heat 
removal pump(s) are removed from operation. The changes allow small, 
controlled, safe insertions of positive reactivity, but limit the 
introduction of positive reactivity to ensure that compliance with the 
required shutdown margin or refueling boron concentration limits are 
satisfied.
    Date of issuance: July 10, 2003.
    Effective date: July 10, 2003, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1-158; Unit 2-159.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40024). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 10, 2003.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: February 28, 2003, as 
supplemented by letter dated June 26, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 3.5.2, ``ECCS--Operating,'' Action A to 
allow a one-time increase in the allowed outage time for centrifugal 
charging pump (CCP) 1-1, for the purpose of seal replacement during 
Unit 1's Cycle 12 from 72 hours to 7 days. Additionally, the amendments 
delete a similar one-time TS change for Unit 2's CCP 2-1 that has 
expired.
    Date of issuance: July 15, 2003.
    Effective date: July 15, 2003, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1--159; Unit 2--160.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 29, 2003 (68 FR 
22753). The June 26, 2003, supplemental letter provided additional 
clarifying information that 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.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated July 15, 2003.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M. 
Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendments request: February 11, 2003.
    Brief Description of amendment: The amendment modifies Technical 
Specifications (TS) to allow a 40-month inspection interval for Farley, 
Unit 2 after the completion of the first post-replacement in-service 
inspection, rather than the completion of two consecutive inspections 
resulting in a classification of C-1.
    Date of issuance: July 14, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 153.
    Facility Operating License No. NPF-8: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25657). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 14, 2003.
    No significant hazards consideration comments received: No.

[[Page 46251]]

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: March 21, 2003.
    Brief description of amendments: The proposed Technical 
specifications (TS) amendments revise TS Section 5.5.1 ``Offsite Dose 
Calculation Manual (ODCM).'' The proposed change will remove reference 
to the Plant Operations Review Committee review and acceptance of 
licensee initiated changes to the ODCM.
    Date of issuance: July 14, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 160 & 152.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28857).The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 14, 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: September 24, 2002, as 
supplemented April 8, 2003.
    Brief description of amendment: This amendment revises the 
Technical Specification (TS) Surveillance Requirement 4.0.3, to 
incorporate the approved consolidated line item improvement program 
change associated with the TS Task Force traveler TSTF-358, ``Change to 
Surveillance Requirement 3.0.3 Regarding Missed Surveillances.'' 
Additionally, a change to the administrative controls section, Section 
6.8, is included, to incorporate a new TS requirement for a Bases 
control program, consistent with the Bases control program presented in 
Section 5.5 NUREG 1431, ``Improved Standard Technical Specifications 
for Westinghouse Plants,'' Revision 2.
    Date of issuance: July 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 163.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70768). The April 8, 2003, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the application.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated July 11, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of July 2003.

    For The Nuclear Regulatory Commission.

Cornelius F. Holden,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-19487 Filed 8-4-03; 8:45 am]
BILLING CODE 7590-01-P