[Federal Register Volume 69, Number 138 (Tuesday, July 20, 2004)]
[Notices]
[Pages 43457-43465]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-16157]


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NUCLEAR REGULATORY COMMISSION


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

Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission to publish notice of any amendments 
issued, or proposed to be issued and 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, June 25, 2004, through July 8, 2004. The 
last biweekly notice was published on July 6, 2004 (69 FRN 40668).

[[Page 43458]]

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. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place 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 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, 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 2.309, 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 within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; 2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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

[[Page 43459]]

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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. 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 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 requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    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].

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: May 25, 2004.
    Description of amendment request: The proposed amendments would 
revise the licensing basis in the Updated Final Safety Analysis Report 
to support installation of a passive low-pressure injection (LPI) cross 
connect inside containment for Unit 3. The proposed changes would 
revise the licensing basis for selected portions of the core flood and 
LPI piping to allow exclusion of the dynamic effects associated with a 
postulated rupture of that piping by application of leak-before-break 
technology. Similar amendments were approved for Unit 1 by NRC letter 
dated September 29, 2003, and for Unit 2 by NRC letter dated February 
5, 2004.
    The proposed amendments would also delete technical specifications 
(TSs) which will no longer apply when the LPI cross connect 
modification has been implemented.
    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 proposed 
License Amendment Request (LAR) modifies the Unit 3 licensing basis 
to allow the dynamic effects associated with postulated pipe rupture 
of selected portions of the Unit 3 Low Pressure Injection (LPI)/Core 
Flood (CF) piping to be excluded from the design basis. The proposed 
LAR also removes Technical Specifications that are no longer 
applicable due to the completion of the LPI cross connect 
modification on all three Oconee Units. The proposed design 
allowances for these selected portions of piping continue to allow 
the LPI system design to meet General Design Criteria (GDC) 4 
requirements related to environmental and dynamic effects. The 
proposed LAR will continue to ensure that ONS [Oconee Nuclear 
Station] can meet design basis requirements associated with the LPI 
safety function. The addition of the crossover line will enhance the 
ability of the control room operator to mitigate the consequences of 
specific events for which LPI is credited. Therefore, the proposed 
LAR 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: The 
proposed LAR modifies the Unit 3 licensing basis to allow the 
dynamic effects associated with postulated pipe rupture of selected 
portions of Unit 3 LPI/CF piping to be excluded from the design 
basis and removes TS requirements that are no longer applicable due 
to the completion of the LPI cross connect modification on all three 
Oconee Units. The proposed design allowances for these selected 
portions of piping continue to allow the LPI system design to meet 
GDC 4 requirements related to environmental and dynamic effects. The 
systems affected by the changes are used to mitigate the 
consequences of an accident that has already occurred. The proposed 
licensing basis change does not affect the mitigating function of 
these systems. Consequently, these changes do not alter the nature 
of events postulated in the Safety Analysis Report nor do they 
introduce any unique precursor mechanisms. Therefore, the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety: The 
proposed licensing basis and TS changes do not unfavorably affect 
any plant safety limits, set points, or design parameters. The 
changes also do not unfavorably affect the fuel, fuel cladding, RCS 
[Reactor Coolant System], or containment integrity. Therefore, the 
proposed changes, which add new design allowances associated with 
the passive LPI cross connect modification and remove obsolete TS 
requirements, 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: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Stephanie M. Coffin (Acting).

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of amendment request: June 4, 2004.
    Description of amendment request: The proposed amendment would 
revise the safety limit values in Technical Specification (TS) 2.1.1.2 
for the minimum critical power ratio (MCPR) for both single and two 
recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 43460]]


    1. The operation of JAFNPP in accordance with the proposed 
amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The basis of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to 
occur if the limit is not violated. The new SLMCPR values preserve 
the existing margin to transition boiling and probability of fuel 
damage is not increased. The derivation of the revised SLMCPR for 
JAFNPP for incorporation into the Technical Specifications, and its 
use to determine plant and cycle-specific thermal limits, have been 
performed using NRC approved methods. These plant-specific 
calculations are performed each operating cycle and if necessary, 
will require future changes to these values based upon revised core 
designs. The revised SLMCPR values do not change the method of 
operating the plant and have no effect on the probability of an 
accident initiating event or transient.
    Based on the above, JAFNPP has concluded that the proposed 
change will not result in a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The operation of JAFNPP in accordance with the proposed 
amendment, will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes result only from a specific analysis for 
the JAFNPP core reload design. These changes do not involve any new 
or different methods for operating the facility. No new initiating 
events or transients result from these changes.
    Based on the above, JAFNPP has concluded that the proposed 
change will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. The operation of JAFNPP in accordance with the proposed 
amendment, will not involve a significant reduction in a margin of 
safety.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity. The operating MCPR limit is set appropriately above the 
safety limit value to ensure adequate margin when the cycle specific 
transients are evaluated. Accordingly, the margin of safety is 
maintained with the revised values.
    As a result, JAFNPP has determined that the proposed change will 
not result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: June 17, 2004.
    Description of amendment request: The amendment will (1) modify 
Technical Specifications (TSs) 5.3.1, Fuel Assemblies, to allow a 
limited number of lead test assemblies (LTAs) and limited substitutions 
of zirconium alloy or stainless steel filler rods for fuel rods, (2) 
include ZIRLOTM as an acceptable fuel rod cladding which is 
consistent with 10 CFR 50.46, (3) relocate some of the information in 
TS 5.3.1 to TS 5.6.1, (4) change TS 6.9.1.11.1 to allow the use of the 
Westinghouse Nuclear Physics code package and to incorporate the 
methodology used to support ZIRLOTM cladding material, and 
(5) delete the Index from the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

    The proposed change allows the use of a limited number of lead 
test assemblies; the use of limited substitutions of zirconium alloy 
or stainless steel filler rods for fuel rods; and the use of methods 
required for the implementation of ZIRLOTM clad fuel 
rods. Inasmuch as the revision identifies codes previously approved 
by the NRC [Nuclear Regulatory Commission] for CE [Combustion 
Engineering] cores, the amendment is administrative in nature and 
has no impact on any plant configuration or system performance 
relied upon to mitigate the consequences of an accident.
    The proposed change in part represents a relocation of a portion 
of the information previously located in the TSs design features 
section to the FSAR [Final Safety Analysis Report], which is 
controlled under 10 CFR 50.59, ``Changes, Tests, and Experiments.'' 
This change is administrative in nature because the design 
requirements for the facility remain the same.
    The proposed change does not remove or modify any of the design 
requirements for the facility or affect any accident initiators, 
conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

    The proposed amendment identifies a change in the nuclear 
physics codes used to confirm the values of selected cycle-specific 
reactor physics parameter limits and includes minor editorial 
changes which do not alter the intent of stated requirements. The 
proposed change also allows the use of methods required for the 
implementation of ZIRLOTM clad fuel rods. Inasmuch as the 
proposed change identifies codes previously approved by the NRC for 
CE cores, the amendment is administrative in nature and has no 
impact on any plant configuration or system performance relied upon 
to mitigate the consequences of an accident. Parameter limits 
specified in the site specific COLR are not changed from the values 
presently required by TSs. Future changes to the calculated values 
of such limits may only be made using NRC approved methodologies, 
must be consistent with all applicable safety analysis limits, and 
are controlled by the 10 CFR 50.59 process. Assumptions used for 
accident initiators and/or safety analysis acceptance criteria are 
not changed by this change.

Index

    The proposed change is administrative in nature and does not 
affect any system or component functional requirements. This change 
does not affect the operation of the plant or affect any component 
that is used to mitigate the consequences of any accident.
    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.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

    The proposed change allows the use of methods required for the 
implementation of ZIRLOTM clad fuel rods. Inasmuch as the 
revision identifies codes previously approved by the NRC for CE 
cores, the amendment is administrative in nature and has no impact 
on any plant configuration or system performance relied upon to 
mitigate the consequences of an accident.
    In addition, the proposed change allows the use of a limited 
number of lead test assemblies. The proposed change is 
administrative in nature. Prior to the use of lead test assemblies, 
fuel designs will be analyzed with applicable NRC staff approved 
codes and methods and shown by tests or analyses to comply with all 
fuel safety design bases to assure no new or different kind of 
accident from any accident previously evaluated will be created.
    And finally the proposed change allows the relocation of a 
portion of the information previously located in the TSs design 
features section to the FSAR. This change is administrative in 
nature and does not create a new or different type of accident than 
previously evaluated because the design requirements for the 
facility remain the same.

[[Page 43461]]

    The proposed change does not remove or modify any of the design 
requirements for the facility or affect any accident initiators, 
conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

    The proposed change identifies a change in the Nuclear Physics 
codes used to confirm the values of selected cycle-specific reactor 
physics parameter limits contained in the COLR. The proposed change 
also allows the use of methodologies required for the implementation 
of ZIRLOTM clad fuel rods. Neither of these changes 
results in a change [to] the physical plant or the modes of 
operation defined in the facility license.

Index

    The proposed change is administrative in nature and does not 
affect any system or component functional requirements. This change 
does not affect the operation of the plant or affect any component 
that is used to mitigate the consequences of any accident.
    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.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

    The proposed change allows the use of methods required for the 
implementation of ZIRLOTM clad fuel rods. Inasmuch as the 
revision identifies codes previously approved by the NRC for CE 
cores, the amendment is administrative in nature and has no impact 
on any plant configuration or system performance relied upon to 
mitigate the consequences of an accident.
    In addition, the proposed change allows the use of a limited 
number of lead test assemblies. The proposed change is 
administrative in nature. Prior to the use of lead test assemblies, 
fuel designs will be analyzed with applicable NRC staff approved 
codes and methods and shown by tests or analyses to ensure 
compliance with any safety analysis acceptance criteria.
    And finally the proposed change allows the relocation of a 
portion of the information previously located in the TSs design 
features section to the FSAR. This change is administrative in 
nature and does not create a new or different type of accident than 
previously evaluated because the design requirements for the 
facility remain the same.
    The proposed change does not remove or modify any of the design 
requirements for the facility or affect any accident initiators, 
conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

    The individual specifications continue to require operation of 
the plant within the bounds of the limits specified in COLR. 
Benchmarking has shown that uncertainties for the Westinghouse 
Physics code system (ANC/PHOENIX-P) yields are essentially the same 
or less than those obtained for the current ROCS/DIT methodology. 
Future changes to the values of these limits by the licensee may 
only be developed using NRC approved methodologies, remaining 
consistent with all applicable plant safety analysis limits 
addressed in the Safety Analysis Report, which are controlled by the 
10 CFR 50.59 process. The relocation of the supplement numbers, 
revision numbers, and approval dates related to the analytical 
methods listed in the COLR does not affect the margin of safety. The 
analysis will continue to be performed using NRC approved 
methodology. Safety analysis acceptance criteria are not being 
altered by this change.

Index

    The proposed change is administrative in nature and does not 
affect any system or component functional requirements. Safety 
analysis acceptance criteria are not being altered by this change.
    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: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application for amendments: June 15, 2004.
    Description of amendment request: The proposed amendment would 
allow the licensee to conduct the monthly diesel surveillance test, the 
diesel full-load rejection test, the diesel 24-hour run test and the 
diesel hot restart test at the higher load of 2800 kW.
    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 revisions to Technical Specification [TS] 
Surveillance Requirements SR 3.8.1.3 (the monthly diesel 
surveillance test), SR 3.8.1.10 (the diesel full-load rejection 
test), SR 3.8.1.14.b (the diesel 24-hour run test), and SR 3.8.1.15 
(the diesel hot restart test) to permit these tests to be conducted 
at the higher load value of 2800 kW do not involve any physical 
change to any EDG [emergency diesel generator] equipment. The 
Operator using existing EDG load controls will adjust the EDG to 
carry the increased load during surveillance testing.
    The EDGs are designed to provide a reliable source of AC 
electrical power in the event of an accident coincident with a loss 
of offsite power. The failure of an EDG itself is not considered an 
accident evaluated in the UFSAR [Updated Final Safety Analysis 
Report]. This proposed loading change does not affect the current 
accident initiators or precursors that could lead to a previously 
evaluated accident.
    The failure of a single EDG to perform when required to mitigate 
the consequences of an accident has already been considered as a 
subsequent single failure in the current plant safety analyses. The 
proposed change to increase the allowable load range does not alter 
the EDG design features, post-accident operation, or accident 
analysis assumptions which could affect the ability of the EDGs to 
mitigate the consequences of a previously evaluated accident. 
Current EDG testing requirements, e.g., starting, timing, and post 
accident sequencing and loading will continue to ensure reliable EDG 
operation and are not being changed in this request.
    Since the EDG TS surveillance test load is the only parameter 
involved in this request, the proposed changes will not increase the 
likelihood of the malfunction of another system, structure, or 
component that has been assumed as an accident initiator or credited 
in the mitigation of an accident.
    Based on the above discussion, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The EDGs are designed to provide a reliable source of AC 
electrical power in the event of an accident coincident with a loss 
of offsite power. No change in the ability of the EDGs to perform 
their design function is involved. Instrumentation setpoints, 
starting, sequencing, and post-accident loading functions associated 
with the EDGs are not affected by the proposed changes. No 
modifications to the EDGs are required to implement the proposed TS 
changes. Therefore, no new failure mechanism, malfunction, or 
accident initiator is considered credible.
    Additionally, the proposed TS changes do not affect the other 
plant design, hardware, system operation, or procedures. Therefore, 
based on the above discussion, the above TS changes do 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 underlying purpose of the four (4) diesel generators is to 
ensure an available source of onsite power to the ESF [engineered 
safety feature] systems. This

[[Page 43462]]

change does [sic] will not impact this underlying purpose. As 
discussed above, this change may result in a slight increase in 
engine wear due to the ability to operate at the higher load, but 
this increased wear is bounded by the existing 24 month maintenance 
inspection program. The OEM [original equipment manufacturer] has 
stated that the change to increase the allowable load value still 
remains well within the EDG 2000-hour rating, and the increased rate 
of wear is within the acceptable limits of the current maintenance 
program.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has found that, because the EDGs will continue to be 
operated within the bounds of the current maintenance program, there is 
no significant increase in the probability of an EDG failure; 
therefore, there is no significant increase in the probability or 
consequences of an accident previously evaluated. The NRC staff further 
finds that, because there is no significant increase in a failure of an 
EDG to perform its function, the proposed change does not create the 
possibility of an accident not previously evaluated.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review and the staff's own findings above, 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 S. O'Neill, Associate and General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: James W. Clifford.

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 amendment request: June 2, 2004.
    Description of amendment request: The proposed amendment would 
revise the BVPS-1 and 2 Technical Specifications to allow operation 
with atmospheric containment designs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The Beaver Valley Power Station (BVPS) 
containments are designed to withstand the internal pressure and 
temperature resulting from a loss of coolant accident (LOCA), main 
steamline break (MSLB), feedwater line break, and a control rod 
ejection accident (CREA). Each of these accidents has been 
previously analyzed with the results provided in the Updated Final 
Safety Analysis Report (UFSAR) except the feedwater line break. This 
accident is not analyzed because the MSLB is more limiting. The 
affect on containment pressure and temperature due to a CREA is 
bounded by a LOCA, since a CREA is modeled after a small break LOCA. 
The probability of occurrence for these accidents is independent of 
the type of containment. Additionally the supporting plant 
modifications will not increase the probability of an accident 
because they perform an accident mitigation function and are not 
accident initiators. Therefore a change from sub-atmospheric to an 
atmospheric containment will not increase the probability of these 
accidents.
    For accident conditions, the proposed changes will potentially 
impact the reported dose consequences of the LOCA and CREA for both 
BVPS units. The radiological consequences of these and the remaining 
design basis accidents are not adversely impacted by the proposed 
changes because they are within the current BVPS licensing and 
design basis.
    From a containment integrity viewpoint, the limiting DBA 
[design-basis accident] presently is the MSLB for Unit 1 and the 
LOCA for Unit 2. Following the conversion to an atmospheric 
containment the limiting DBA will be the LOCA for both units. The 
revised containment integrity analysis demonstrates that with the 
installation of the supporting plant modifications that the 
pressures and temperatures associated with the applicable design 
basis accidents identified above are within the existing containment 
design limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The design basis accidents, which could be 
adversely affected by the proposed changes, have been reanalyzed. 
These [re]analyses demonstrate that all acceptance criteria have 
been satisfied. The revised containment integrity analysis 
demonstrates that the containment will not be subjected to 
temperatures or pressures that are beyond its design limits. 
Converting to an atmospheric containment will not result in any new 
or different kind of accidents because no new accident initiators 
will be introduced.
    The affects of the supporting plant modifications and the 
proposed Technical Specification changes on plant structures, 
systems and components (SSC) have been evaluated and it has been 
verified that the capability of the SSCs to perform their design 
functions will be retained following approval of the proposed 
Technical Specification changes and installation of the supporting 
plant modifications.
    Changes to instrumentation setpoints, surveillance requirements, 
installation of the supporting plant modifications, and the 
elimination of certain operability requirements will not create the 
possibility of a new or different type of accident since these 
changes would not result in significant changes to the manner in 
which the affected equipment is operated during normal plant 
operations.
    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 attributed to the containment 
involves both the pressures and temperatures the containment is 
subjected to following a DBA, and the on-site and offsite dose 
consequences associated with normal and post DBA operations.
    The revised containment analyses demonstrates that, following a 
DBA; containment peak pressure and temperature will not exceed the 
containment's design limits and that the containment pressure will 
not decrease to below 8 psia following the intentional or 
inadvertent actuation of the quench spray system. Since the 
containment design limits are not exceeded, the existing margin of 
safety between these limits and the containment failure limits is 
not reduced.
    Since the current radiological analyses impacted by the 
containment conversion are conservatively based on atmospheric 
operation, it is concluded that the existing dose consequence margin 
of safety will not be impacted when the BVPS units are operated with 
an atmospheric containment.
    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.

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

    Date of amendment request: April 26, 2004.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Limiting Conditions for Operation (LCO) 
3.7.9, ``Ultimate Heat Sink (UHS)'' to allow the UHS to remain OPERABLE 
with three of four fans operating under certain environmental 
conditions.

[[Page 43463]]

    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 consequence of an accident previously evaluated?
    No. The revised requirements will maintain OPERABILITY while 
allowing maintenance on one fan when ambient wet-bulb temperature is 
63 [deg]F or lower. Modifying the condition when one NSCW [nuclear 
service cooling water] tower is impacted is more restrictive. The 
UHS is not an initiator to any analyzed accident sequence. Operation 
in accordance with the proposed TS will continue to ensure that the 
UHS remains capable of performing its safety function and that all 
analyzed accidents will continue to be mitigated as previously 
analyzed. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not introduce any new equipment, 
create new failure modes for existing equipment, or create any new 
limiting single failures. Plant operation will not be altered, and 
all safety functions previously addressed in accident analyses will 
continue to be performed. Therefore, the proposed changes do 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?
    No. The proposed changes will not adversely affect operation of 
plant equipment-principally the UHS and the equipment supported by 
it. Modifying the condition where one NSCW tower is impacted is more 
restrictive and enhances the margin of safety. Therefore, the 
proposed changes do not involve a significant reduction in any 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Stephanie M. Coffin (Acting).

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 21, 2004.
    Description of amendment request: The proposed one-time (per unit) 
change revises the steam generator (SG) inservice inspection frequency 
requirements in Technical Specification (TS) 4.4.5.3a for Unit 1 
immediately after the tenth refueling outage for Unit 1 (1RE10) and for 
Unit 2 immediately after refueling outage 2RE10. The change would allow 
a 78-month inspection interval after one inspection resulting in C-1 
classification, rather than a 40-month interval after two consecutive 
inspections resulting in C-1 classification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    There is no direct increase in SG leakage because the proposed 
change does not alter the plant design. The scope of inspections 
performed during 1RE10 and 2RE10, the first refueling outage 
following SG replacement, exceeded the combined TS requirements for 
the first two refueling outages after replacement. That is, more 
tubes were inspected than were required by TS. Currently, neither 
Unit 1 nor Unit 2 has an active SG damage mechanism and will meet 
the current industry examination guidelines without performing 
inspections during the next 78 months. The Condition Monitoring 
Assessment after 1RE10 and 2RE10 demonstrated that all performance 
criteria were met during these outages. The Operational Assessment 
shows that all performance criteria will be met over the proposed 
operating period.
    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 will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections performed during 1RE10 and 2RE10, the first 
refueling outage for each unit following SG replacement, 
significantly exceeded the combined TS requirements for the scope of 
the first two refueling outages after SG replacement. The 
inspections already performed exceed those required by the current 
TS over the proposed 78-month period.
    The proposed change does not affect the design of the SGs, the 
method of operation, or reactor coolant chemistry controls. No new 
equipment is being introduced and installed and equipment is not 
being operated in a new or different manner. The proposed change 
involves a one-time extension of the SG tube inservice inspection 
interval, and therefore will not give rise to new failure modes. In 
addition, the proposed change does not impact any other plant system 
or components.
    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.
    Steam generator tube integrity is a function of design, 
environment, and current physical condition. Extending the SG tube 
inservice inspection interval to 78 months will not alter the 
function or design of the SGs. Inspections conducted prior to 
placing the SGs into service (pre-service inspections) and 
inspection during the first refueling outages following SG 
replacement demonstrate that the SGs do not have fabrication damage 
or an active damage mechanism. The scope of those inspections 
significantly exceeded those required by the TS. These inspection 
results were comparable to similar inspection results for the same 
model of RSGs [replacement steam generators] installed at other 
plants, and subsequent inspections at those plants yielded results 
that support this extension request. The improved design of the RSGs 
also provides reasonable assurance that significant tube degradation 
is not likely to occur over the proposed operating period.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

[[Page 43464]]

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 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].

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power 
Station, Unit No. 1, New London County, Connecticut

    Date of amendment request: September 18, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification 4.2, ``Fuel Storage,'' to eliminate all credit for 
Boraflex as a neutron absorber, reduce the number of fuel assemblies 
allowed to be stored in the spent fuel pool (SFP), change the required 
SFPkeff and eliminate design features requirements of new 
fuel storage.
    Date of issuance: June 29, 2004.
    Effective date: June 29, 2004, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 113.
    Facility Operating License No. DPR-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68659). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 29, 2004.
    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: October 15, 2003.
    Brief description of amendments: The amendments added a new 
Technical Specification (TS) 3.9.7, ``Unborated Water Source isolation 
Valves,'' and revised TS 3.9.2, ``Nuclear Instrumentation,'' to delete 
the requirement for Boron Dilution Mitigation System automatic valve 
actuations and makeup water pump trip during Mode 6 and to agree with 
the wording of NUREG-1431, ``Standard Technical Specifications 
Westinghouse Plants,'' Revision 2. The licensee proposed these changes 
to provide configuration control of the dilution valves during Mode 6 
to preclude the possibility of a boron dilution event and to provide an 
opportunity to conduct maintenance on the volume control tank valves, 
refueling water storage tank valves, and their respective power 
supplies.
    Date of issuance: June 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 215 and 209.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12366).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 2004.
    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: June 25, 2003.
    Brief description of amendments: The amendments are administrative 
in nature and incorporate several editorial changes.
    Date of issuance: June 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 222 and 204.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19565).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 2004.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point 
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

    Date of application for amendment: March 3, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications administrative controls requirements regarding 
the reactor coolant pump flywheel inspection program to increase the 
inspection interval from 10 years to 20 years.
    Date of issuance: July 2, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 240 and 221.
    Facility Operating License Nos. DPR-26 and DPR-64: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19566).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 2, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 27, 2003, as 
supplemented December 15, 2003, and February 27, 2004.
    Brief description of amendments: The amendments modify Technical 
Specifications requirements to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change TSTF-359, ``Increase 
Flexibility in Mode Restraints.''
    Date of issuance: June 25, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 281 and 265.

[[Page 43465]]

    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59217).
    The supplemental letters dated December 15, 2003, and February 27, 
2004, provided clarifying information that did not change the scope of 
the original Federal Register notice or the original no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of application for amendment: June 3, 2003, as supplemented by 
letters dated October 6, 2003, January 15, and February 13, 2004.
    Brief description of amendment: The amendment revises the operating 
license and technical specifications to increase the licensed rated 
power by 1.4 percent from 2530 megawatts thermal (MWt) to 2565.4 MWt 
using measurement uncertainty recapture.
    Date of issuance: June 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40714).
    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 June 23, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 20, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specification requirements for Shift Technical Advisor 
coverage.
    Date of issuance: June 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 132 and 111.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19574).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 6, 2003, as supplemented by letter 
dated February 24, 2004.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) adopting the TS Task Force (TSTF) 
Standard TS Change Traveler TSTF-360, Revision 1, ``DC Electrical 
Rewrite.'' Specifically, the amendments revise the TS 3.8.4, ``DC 
Sources-Operating,'' TS 3.8.5, ``DC Sources-Shutdown,'' TS 3.8.6, 
``Battery Cell Parameters,'' and TS 5.5.19, ``Battery Monitoring and 
Maintenance Program.''
    Date of issuance: July 1, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 113 and 113.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40721). The February 24, 2004, 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 amendments is contained 
in a Safety Evaluation dated July 1, 2004.
    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 12th day of July 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-16157 Filed 7-19-04; 8:45 am]
BILLING CODE 7590-01-P