[Federal Register Volume 68, Number 218 (Wednesday, November 12, 2003)]
[Notices]
[Pages 64133-64145]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-28065]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from October 17, 2003, through October 30, 2003. 
The last biweekly notice was published on October 28, 2003 (68 FR 
59212).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed

[[Page 64134]]

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 December 12, 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 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

[[Page 64135]]

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].

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

    Date of amendment request: July 1, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed changes will 
delete one and add two references to the list of analytical methods in 
TS 5.6.5, ``Core Operating Limits Report (COLR),'' that can be used to 
determine core operating limits. The deleted reference is to an 
analytical method that is no longer applicable to LaSalle County 
Station (LSCS). The new references will allow LSCS to use General 
Electric Company (GE) methods for the determination of fuel assembly 
critical power of Framatome Advanced Nuclear Fuel, Inc. (Framatome) 
Atrium-9B and Atrium-10 fuel. The proposed changes are the result of a 
LSCS decision to insert GE14 fuel during the upcoming refueling outage 
at LSCS Unit 1 in January 2004. GE's safety analysis methodologies have 
been previously used at LSCS and GE14 fuel is currently in use at other 
Exelon Generation Company, LLC (Exelon), stations.
    The first added reference, ``GEXL96 Correlation for Atrium-9B 
Fuel,'' will list a method that was previously approved by the NRC for 
use by licensees. The second added reference, ``GEXL97 Correlation for 
Atrium-10 Fuel,'' will list a GE method for determining the critical 
power for Atrium-10 fuel. This correlation has not been previously 
reviewed and approved by the NRC for use by licensees. Additionally, 
editorial changes will be made to existing references.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes will delete one and add two additional 
references to the list of administratively controlled analytical 
methods in TS 5.6.5, ``Core Operating Limits Report (COLR),'' that 
can be used to determine core operating limits and make minor 
editorial changes to the existing references. TS 5.6.5 lists NRC 
approved analytical methods used at LaSalle County Station (LSCS) to 
determine core operating limits. [LSCS Unit 1 is scheduled to load 
GE fuel during its upcoming outage in January 2004.]
    The proposed changes to TS Section 5.6.5 will add the fuel 
analytical methods that support the initial insertion of GE14 fuel 
to the list of methods used to determine the core operating limits. 
The deletion or addition of approved methods to TS Section 5.6.5 and 
minor editorial changes to the existing references has no effect on 
any accident initiator or precursor previously evaluated and does 
not change the manner in which the core is operated. The methods 
have been reviewed to ensure that the output accurately models 
predicted core behavior, have no effect on the type or amount of 
radiation released, and have no effect on predicted offsite doses in 
the event of an accident. Thus, the proposed changes do not have any 
effect on the probability of an accident previously evaluated.
    The proposed changes in the administratively controlled 
analytical methods does [do] not affect the ability of LSCS to 
successfully respond to previously evaluated accidents and does [do] 
not affect radiological assumptions used in the evaluations. Thus, 
the radiological consequences of any accident previously evaluated 
are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to TS Section 5.6.5 do not affect the 
performance of any LSCS structure, system, or component credited 
with mitigating any accident previously evaluated. The insertion of 
a new generation of fuel which has been analyzed with NRC approved 
methodologies will not affect the control parameters governing unit 
operation or the response of plant equipment to transient 
conditions. The proposed changes do not introduce any new modes of 
system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes will delete one and add two additional 
references to the list of administratively controlled analytical 
methods in TS 5.6.5 that can be used to determine core operating 
limits and make minor editorial changes to the titles of existing 
references. The proposed changes do not modify the safety limits or 
setpoints at which protective actions are initiated, and do not 
change the requirements governing operation or availability of 
safety equipment assumed to operate to preserve the margin of 
safety. Therefore, LSCS has determined that the proposed changes 
provide an equivalent level of protection as that currently 
provided.
    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.

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

    Date of amendment request: September 11, 2003.
    Description of amendment request: Revise the dose model for the 
containment activated concrete, rebar (hereafter referred to as 
activated concrete) and liner, by incorporating more realistic 
radionuclide release rates and to change the associated derived 
concentration guideline limit (DCGL) for activated concrete.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 64136]]

issue of no significant hazards consideration, which is presented 
below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested license amendment does not authorize any plant 
activities beyond those allowed by 10 CFR Chapter I or beyond those 
considered in the DSAR. The bounding accident described in the 
Defueled Safety Analysis Report (DSAR) for potential airborne 
activity is the postulated resin cask drop accident in the Low Level 
Radioactive Waste Storage Building. This accident is expected to 
contain more potential airborne activity than can be released from 
other decommissioning events. The radionuclide distribution assumed 
for the spent resin cask has a greater inventory of transuranic 
radionuclides (the major dose contributor) than the distribution of 
plant derived radionuclides in the components involved in other 
decommissioning accidents. The other accidents considered in the 
DSAR include: (1) Explosion of liquid petroleum gas (LPG) leaked 
from a front end loader or forklift; (2) Explosion of oxyacetylene 
during segmenting of the reactor vessel shell; (3) Release of 
radioactivity from the RCS decontamination ion exchange resins; (4) 
Gross leak during in-situ decontamination; (5) Segmentation of RCS 
piping with unremoved contamination; (6) Fire involving contaminated 
clothing or combustible waste; (7) Loss of local airborne 
contamination control during blasting or jackhammer operations; (8) 
Temporary Loss of Services; (9) Dropping of Contaminated Concrete 
Rubble; (10) Natural phenomena; and (11) Transportation accidents. 
The probabilities and consequences for these accidents are estimated 
in the basis documentation for DSAR Section 7. No systems, 
structures, or components that could initiate or be required to 
mitigate the consequences of an accident are affected by the 
proposed change in any way not previously evaluated in the DSAR. 
Since Maine Yankee does not exceed the salient parameters associated 
with the plant referenced in the basis documentation in any material 
respects, it is concluded that these probabilities and consequences 
are not increased. Therefore, the proposed change to the Maine 
Yankee license does not involve any increase in the probability or 
consequences of any 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 requested license amendment does not authorize any plant 
activities that could precipitate or result in any accidents beyond 
those considered in the DSAR. The accidents previously evaluated in 
the DSAR are described above. These accidents are described in the 
basis documentation for DSAR Section 7. The proposed change does not 
affect plant systems, structures, or components in any way not 
previously evaluated in the DSAR. Since Maine Yankee does not exceed 
the salient parameters associated with the plant referenced in the 
basis documentation in any material respects, it is concluded that 
these accidents appropriately bound the kinds of accidents possible 
during decommissioning. Therefore, the proposed change to the Maine 
Yankee license would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety defined in Maine Yankee's license basis for 
the consequences of decommissioning accidents has been established 
as the margin between the bounding decommissioning accident and the 
dose limits associated with the need for emergency plan offsite 
protection, namely the Environmental Protection Agency Protective 
Action Guidelines EPA-PAGs. As described above, the bounding 
decommissioning accident is the postulated resin cask drop accident 
in the Low Level Radioactive Waste Storage Building. Since the 
bounding decommissioning accident is expected to contain more 
potential airborne activity than can be released from other 
decommissioning events and since the radionuclide distribution 
assumed for the spent resin cask has more transuranics (the major 
dose contributor) than the distribution in the components involved 
in other decommissioning accidents, the margin of safety associated 
with the consequences of decommissioning accidents cannot be 
reduced. The margin of safety defined in the statements of 
consideration for the final rule on the Radiological Criteria for 
License Termination is described as the margin between the 100 mrem/
yr public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use. 
This margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Since the 
license termination plan (LTP) was designed to comply with the 
radiological criteria for license termination for unrestricted use, 
the margin of safety cannot be reduced. Therefore, the proposed 
changes to the Maine Yankee license would not involve a significant 
reduction in any margin of safety.

Conclusion

    Based on the above, Maine Yankee concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power 
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: April 11, 2003.
    Description of amendment request: The proposed amendment would make 
various administrative, editorial, and typographical changes to 
Technical Specification (TS) Section 5.0, ``Administrative Controls.'' 
Specifically, the proposed changes would:
    (1) Correct TS 5.4.1.a by adding ``Appendix A'' after the reference 
to ``Regulatory Guide 1.33, Revision 2,'' and deleting ``of'' before 
this reference.
    (2) Change TS 5.5.2.e by deleting the phrase ``(approximately 44 
psig)'' which is an invalid reference to the normal hydrostatic head 
from the safety injection refueling water tank for the test conditions 
required for maximum allowable leakage from recirculation heat removal 
systems' components.
    (3) Make several editorial changes to TS 5.6.1 to be consistent 
with the wording of NUREG-1432, ``Standard Technical Specifications-
Combustion Engineering Plants,'' Revision 2 (STS), and the changes to 
the STS in Technical Specification Task Force (TSTF) Traveler TSTF-152. 
The editorial changes include (a) adding the word ``collective'' to 
describe the associated collective deep dose equivalent, (b) adding 
``thermoluminescence dosimeter'' to define its acronym ``(TLD),'' (c) 
changing ``stations'' to ``station,'' (d) adding the words ``received 
from'' when describing the 80 percent of total deep dose equivalent 
received from external sources, and (e) making punctuation changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The following evaluation supports the finding that operation of the 
facility in accordance with the proposed change would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed license amendment provides changes to Technical 
Specification (TS) Administrative Controls sections 5.4.1.a, 
5.5.2.e, and 5.6.1. The proposed corrections to TS 5.4.1.a are 
editorial in nature. The proposed correction to TS 5.5.2.e, which

[[Page 64137]]

deletes an erroneous approximate value from the description of test 
conditions for maximum allowable leakage from recirculation heat 
removal system components, is consistent with the existing plant 
design as described in the Palisades Final Safety Analysis Report. 
The proposed correction to TS 5.6.1 is editorial in nature and is 
consistent with the Nuclear Regulatory Commission approved standard 
technical specifications. The proposed amendment does not involve 
operation of the required structures, systems or components (SSCs) 
in a manner or configuration different from those previously 
recognized or evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 amendment does not involve a physical alteration of 
any SSC or a change in the way any SSC is operated. The proposed 
amendment does not involve operation of any required SSCs in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by the 
changes being requested.
    Thus, this change does 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 amendment does not affect any margin of safety. The 
proposed amendment does not involve any physical changes to the 
plant or manner in which the plant is operated.
    Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: September 19, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Limiting Conditions for Operation 
(LCO) 3.8.4, ``DC Sources--Operating,'' for the remainder of operating 
cycle 19. Specifically, the proposed TS change would increase the 
Completion Time for the 1B Auxiliary Building DC electrical power 
system inoperability due to an inoperable battery to allow for on-line 
replacement of individual cells. Cycle 19 is presently scheduled to end 
on October 2, 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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to LCO 3.8.4 creates an extended Completion 
Time for an inoperable 1B Auxiliary Building DC electrical power 
subsystem due to an inoperable battery on Unit 1 only for the 
remainder of operating cycle 19. The Auxiliary Building battery is 
not a direct initiator of any analyzed accident sequence. The 
radiological consequences of any associated accidents are not 
impacted by the proposed amendment. Therefore, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change involves no change to the physical plant. It 
allows additional time for corrective maintenance on the 1B 
Auxiliary Building battery on Unit 1. The proposed amendment 
involves an extension of a previously determined acceptable mode of 
operation. The proposed amendment does not introduce any new 
equipment, create new failure modes for existing equipment, or 
create any new limiting single failures. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The physical plant is unaffected by these changes. The proposed 
changes do not impact accident offsite dose, containment pressure or 
temperature, emergency core cooling system (ECCS) or reactor 
protection system (RPS) settings or any other parameter that could 
affect a margin of safety. Under the proposed amendment, the unit 
will continue to be operated in a condition that will ensure that 
emergency power will be available as needed. The extended Completion 
Time for an inoperable battery has been shown to have a very small 
impact on plant risk using the criteria of Regulatory Guides 1.174, 
An Approach for Using Probabilistic Risk Assessments in Risk-
Informed Decision-making and 1.177, An Approach for Plant-Specific. 
Risk-Informed Decisionmaking: Technical Specifications and is 
acceptable. Therefore, the proposed amendment does not involve a 
significant reduction in a margin to 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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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 amendment request: August 29, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Limiting Condition of Operation 3.9.3, 
``Containment Penetrations.'' The proposed changes would allow the 
equipment hatch to be open during core alterations and/or during 
movement of irradiated fuel assemblies within containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes will allow the equipment hatch to be 
open during core alterations and movement of irradiated fuel 
assemblies inside containment. The proposed changes will not alter 
the manner in which fuel is handled or core alterations are 
performed. The equipment hatch is not an initiator of any accident. 
The status of the equipment hatch during refueling operations has no 
effect on the probability of the occurrence of any accident 
previously evaluated. The radiological consequences of a fuel 
handling accident inside containment have been determined to be well 
within the limits of 10 CFR 100 and they meet the acceptance 
criteria of General Design Criterion (GDC) 19. Therefore the 
proposed changes do not involve a significant increase in the 
probability or consequences of [any] 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 create any new failure modes for 
any system or component, nor do they adversely affect plant 
operation. No new equipment will be added and no new limiting single 
failures

[[Page 64138]]

will be created. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident [from any 
accident] previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The dose consequences were determined to be well within the 
limits of 10 CFR 100 and they meet the acceptance criteria of GDC 
19. Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 3, 2003.
    Description of amendment request: The proposed amendments would add 
a Limiting Condition for Operation (LCO) for the Linear Heat Generation 
Rate. The new LCO will be included in Section 3.2, Power Distribution 
Limits. The proposed amendments would also change the recirculation 
loop LCO, Section 5.6.5, and the appropriate Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed addition of LCO 3.2.3 and supporting Bases are 
being made to support new modeling improvements in core monitoring. 
This change is administrative in nature in that it does not involve, 
require, or result from any physical change to the plant, including 
the reactor core or its fuel. The addition of LCO 3.2.3 and Bases B 
3.2.3 is consistent with Revision 2 of Volumes 1 and 2 of NUREG-
1433. Changes being proposed for Bases section B 3.2.1 and TS 
Section 5.6.5 are simply supportive in nature to the relocation of 
LHGR [linear heat generation rate] from the APLHGR [averageplant 
linear heat generation rate] Section Bases B 3.2.1 to the new 
section LHGR B 3.2.3.
    Also, no changes are being proposed to any plant system, 
structure, or component designed to prevent or mitigate the 
consequences of a previously evaluated event.
    Therefore, because the physical characteristics and performance 
requirements of the plant systems, structures, and components 
(including the reactor core and fuel) will not be altered, the 
proposed license amendment does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No plant systems, structures, or components (including the 
reactor core and fuel) will be altered by the proposed change to the 
LCO or supporting Bases.
    Additionally, this TS [technical specification] change request 
does not propose changes in the operation of any plant system. 
Consequently, new and unanalyzed modes of operation are not 
introduced.
    As a result, the possibility of a new or different kind of 
accident from any previously evaluated is not introduced.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Previously, the LHGR was included in the monitoring of the 
APLHGR. Now, SNC [Southern Nuclear Company] proposes to monitor LHGR 
on its own while continuing to monitor APLHGR. This proposed TS 
change adds an LCO for LHGR and a corresponding requirement for the 
COLR [core operating limits report].
    The margin of safety is not reduced since the LHGR and APLHGR 
will continue to be monitored.

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

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: May 13, 2003.
    Description of amendment request: The proposed license amendment 
would allow use of a revised methodology, for performance of certain 
accident analyses, described in Westinghouse Electric Corp. (W) report 
WCAP-14882-S1-P, Revision 0 (Proprietary), ``RETRAN-02, Modeling and 
Qualification for Westinghouse Pressurized Water Reactors Non-LOCA 
Safety Analyses, Supplement 1--Thick Metal Mass Heat Transfer Model and 
NOTRUMP-Based Steam Generator Mass Calculation Method,'' dated December 
2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the 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 methodology uses more realistic 
computer models with unnecessary conservatism removed. The 
methodology used to analyze the consequences of a postulated 
accident is not an initiator that can affect the probability or 
consequences of that accident. The change does not alter assumptions 
previously made in the radiological consequences of the 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. The proposed methodology uses more realistic 
computer models with unnecessary conservatism removed. The 
methodology used to analyze the consequences of a postulated 
accident is not an initiator that can cause an accident to occur. 
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 proposed methodology uses more realistic 
computer models with unnecessary conservatism removed. Using the 
methodology of WCAP-14882-S1-P results in additional margin to 
pressurizer overfill for a postulated loss of normal feedwater/ loss 
of offsite power at STP. 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.

[[Page 64139]]

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

    Date of amendment request: September 22, 2003.
    Description of amendment request: The proposed amendments would 
change the requirements for the Engineered Safety Feature sequencer, 
and the Surveillance Requirements that are applicable in Mode 5 and 6 
to provide needed clarification. In addition, the proposed amendment 
would correct a typographical error in that requirement ``c.'' in 
Technical Specification 3.2.4 should actually be requirement ``b.''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator.
    Therefore, STPNOC concludes that there is no 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 changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator.
    Therefore, STPNOC concludes 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.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, STPNOC concludes the 
proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. With regard to the licensee's proposed correction of a 
typographical error in TS 3.2.4, the NRC staff notes the following:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Correction of a typographical error does not change the plant 
design basis, system configuration or operation, and does not add or 
affect any accident initiator. Therefore, there is no 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.
    Correction of a typographical error does not change the plant 
design basis, system configuration or operation, and does not add or 
affect any accident initiator. 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.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the above, the NRC staff concludes 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.

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

    Date of amendment request: September 22, 2003.
    Description of amendment request: The proposed amendments would 
change the Technical Specification 3.3.2 requirements for Loss of Power 
Instrumentation (Functional Unit 8).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator.
    Therefore, STPNOC concludes that there is no 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 changes do not change the plant design basis, 
system configuration or operation, and do not add or affect any 
accident initiator.
    Therefore, STPNOC concludes 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.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, STPNOC concludes the 
proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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.

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: October 16, 2003.
    Description of amendment request: The proposed amendments request a 
one-time change to Technical Specification (TS) 4.4.5.3a to extend the 
40-month steam generator inspection interval to 44 months for Unit 1 
only.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not alter the plant design. The scope 
of inspections

[[Page 64140]]

performed during 1RE10 [Refueling Outage 10 for Unit 1], the first 
refueling outage following SG [steam generator] replacement, 
exceeded the TS requirements for the first two refueling outages 
after replacement combined. That is, more tubes were inspected than 
were required by TS. Currently, South Texas Project Unit 1 does not 
have an active SG damage mechanism and will meet the current 
industry examination guidelines without performing inspections 
during the next refueling outage. The results of the Condition 
Monitoring Assessment after 1RE10 demonstrated that all performance 
criteria were met during 1RE10. The results of the 1RE10 Operational 
Assessment show 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 does not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections performed during 1RE10, the first refueling 
outage following SG replacement, significantly exceeded the TS 
requirements for the scope of the first two refueling outages after 
SG replacement combined.
    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 equipment is not being 
operated in a new or different manner. The proposed change involves 
a one-time extension to 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 frequency [interval] by four months does not 
alter the function or design of the SGs. Inspections conducted prior 
to placing the SGs into service (preservice inspections) and 
inspection during the first refueling outage 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 
replacement SGs also provides reasonable assurance that significant 
tube degradation is not likely to occur over the proposed operating 
period.
    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 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.

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: September 23, 2003.
    Brief description of amendments: The proposed change would revise 
Technical Specification (TS) 3.6.3 entitled, ``Containment Isolation 
Valves,'' to extend the frequency of Surveillance Requirement 3.6.3.7 
for containment and hydrogen purge valves and containment pressure 
relief valves with resilient seats.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Operability and leakage control effectiveness of the containment 
purge, hydrogen purge and containment pressure relief system 
isolation valves have no effect on whether or not an accident 
occurs. Consequently, increasing the interval between surveillances 
of isolation valve leakrate does not involve a significant increase 
in the probability of an accident previously evaluated. The 
consequences of a non-isolated reactor containment building at the 
time of a fuel-handling accident or LOCA [loss-of-coolant accident] 
is release of radionuclides to the environment. Analyses have 
conservatively assumed that a containment pressure relief system 
line is open at the time of an accident, and release to the 
environment continues until the isolation valves are closed. In 
addition, LOCA analyses assume containment leakage of 0.1% of the 
containment volume per day for the first 24 hours and 0.05% per day 
for the duration of the accident. Consequently, increasing the 
interval between surveillances of isolation valve leakrate does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. The functions of the 
containment purge, hydrogen purge and containment pressure relief 
systems are not altered by this change. Therefore, this proposed 
change does not create the possibility of an accident of a different 
kind than previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This proposed change only increases the interval between 
surveillance tests of the containment purge, hydrogen purge and 
containment pressure relief system valves. Analyses have 
conservatively assumed that the containment purge valves are open at 
the time of a fuel handling accident, and that the containment 
pressure relief valve is open at the time of a loss-of-coolant 
accident. In addition, LOCA analyses assume containment leakage of 
0.1% of the containment volume per day for the first 24 hours and 
0.05% per day for the duration of the accident. The radiological 
consequences of both an fuel handling accident and a LOCA are 
unchanged and remain within the 10 CFR 100 limits. 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: October 17, 2003.
    Description of amendment request: The licensee is proposing to 
revise Technical Specification (TS) Section 5.5.6, ``Containment Tendon 
Surveillance Program,'' for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The proposed 
revision to

[[Page 64141]]

TS 5.5.6 is to indicate that the Containment Tendon Surveillance 
Program, inspection frequencies, and acceptance criteria shall be in 
accordance with Section XI, Subsection IWL of the ASME Boiler and 
Pressure Vessel Code and the applicable addenda as required by 10 CFR 
50.55a, except where an exemption or relief has been authorized by the 
NRC. The licensee has also proposed to delete the provisions of 
Surveillance Requirement (SR) 3.0.2 from this specification. In 
addition, the licensee is proposing to revise TS 5.5.16, ``Containment 
Leakage Rate Testing Program,'' to add exceptions to Regulatory Guide 
1.163, ``Performance-Based Containment Leak-Testing Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The revised 
requirements do not affect the function of the containment post-
tensioning system components. The post-tensioning systems are 
passive components whose failure modes could not act as accident 
initiators or precursors.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Containment Leakage Rate Testing Program. In 
addition, the proposed change allows those examinations to be 
performed during power operation[,] as opposed to during a refueling 
outage. The frequency of visual examinations of the concrete 
surfaces of the containment and the mode of operation during which 
those examinations are performed has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations[,] that are performed pursuant to NRC approved ASME 
Section XI Code requirements (except where relief has been granted 
by the NRC)[,] to meet the intent of visual examinations [as] 
required by Regulatory Guide 1.163, without requiring additional 
visual examinations pursuant to the Regulatory Guide. The intent of 
early detection of deterioration will continue to be met by the more 
rigorous requirements of the Code[-]required visual examinations. As 
such, the safety function of the containment as a fission product 
barrier is maintained.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The 
function of the containment post-tensioning system components are 
not altered by this change. The change affects the frequency of 
visual examinations that will be performed for the concrete surfaces 
containments. In addition, the proposed change allows those 
examinations to be performed during power operation[,] as opposed to 
during a refueling outage. The proposed change does not involve a 
modification to the physical configuration of the plant (i.e., no 
new equipment will be installed) or change in the methods governing 
normal plant operation. The proposed change will not impose any new 
or different requirements or introduce a new accident initiator, 
accident precursor, or malfunction mechanism. Additionally, there is 
no change in the types or increases in the amounts of any 
effluent[s] that may be released off-site and there is no increase 
in individual or cumulative occupational exposure.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class CC. The 
function of the containment post-tensioning system components are 
not altered by this change. The change affects the frequency of 
visual examinations that will be performed for the concrete surfaces 
containments. In addition, the proposed change allows those 
examinations to be performed during power operation[,] as opposed to 
during a refueling outage. The safety function of the containment as 
a fission product barrier will be 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

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

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, (IP2) Westchester County, New York

    Date of application for amendment: March 27, 2002, as supplemented 
May 30, 2002, July 10, 2002, October 10, 2002, October 28, 2002, 
November 26, 2002, December 18, 2002, January 6, 2003, January 27, 
2003, February 26, 2003, April 8, 2003, May 19, 2003, June 23, 2003, 
June 26, 2003, July 15, 2003, August 6, 2003, September 11, 2003, 
October 8, 2003, and October 14, 2003.
    Brief description of amendment: The licensee proposed to convert 
the current Technical Specifications (TSs) for IP2, to a set of 
improved TSs based on NUREG-1431, ``Standard Technical Specifications 
for Westinghouse Plants,'' Revision 2, dated April 2001.
    Date of publication of individual notice in Federal Register: 
September 26, 2003 (68 FR 55660).
    Expiration date of individual notice: October 27, 2003.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

[[Page 64142]]

Commission's rules and regulations. The Commission has made appropriate 
findings as required by the Act and the Commission's rules and 
regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: January 16, 2003, as 
supplemented June 11, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specifications to incorporate changes associated with Cycle 15 core 
reload design analysis. The Cycle 15 core reload design implements the 
Framatome ANP Statistical Core Design methodology . This amendment 
permits the licensee to determine the minimum departure from nucleate 
boiling ratio using an NRC-approved methodology based on statistical 
analysis of operational and design uncertainties.
    Date of issuance: October 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 247.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12948). The supplement dated June 11, 2003, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 20, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 7, 2002, as 
supplemented by letters dated April 25, July 10, July 30, August 13, 
September 18, and October 1, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.2.4, ``Departure From Nucleate Boiling Ratio 
(DNBR),'' TS 3.3.1, ``Reactor Protective System (RPS) Instrumentation--
Operating,'' TS 3.3.3, ``Control Element Assembly Calculators 
(CEACs),'' and TS 5.4.1, ``Administrative Controls--Procedures.'' The 
revisions are to Limiting Conditions for Operations (LCOs), LCO 
Actions, LCO Surveillance Requirements, and the procedures used to 
modify the core protection calculator addressable constants.
    Date of issuance: October 24, 2003.
    Effective date: October 24, 2003, and shall be implemented for Unit 
1 no later than prior to entry of Unit 1 into Mode 4 during the restart 
from the Unit 1 spring 2004 refueling outage; for Unit 2 within 90 days 
of the date of issuance, but no later than prior to entry of Unit 2 
into Mode 4 during the restart from the Unit 2 fall 2003 refueling 
outage; and for Unit 3 no later than prior to entry of Unit 3 into Mode 
4 during the restart from the Unit 3 fall 2004 refueling outage.
    Amendment Nos.: Unit 1-150, Unit 2-150, Unit 3-150.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revise the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75868) with a later notice on August 18, 2003 (68 FR 49527).
    The August 13, September 18 and October 1, 2003, supplemental 
letters provided clarifying information that was within the scope of 
the Federal Register Notice (68 FR 49257) and did not change the no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments are contained 
in a Safety Evaluation dated October 24, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 16, 2002, as supplemented by 
letter dated September 11, 2003.
    Brief description of amendment: The amendment revises the current 
main steam isolation valve (MSIV) Technical Specification (TS) 3/4 
7.1.5 to more closely reflect TS 3.7.2 contained in NUREG-1432, 
Revision 2. In addition, this change removes the MSIVs from the scope 
of containment isolation valve TS 3/4 6.3 such that only TS 3/4.7.1.5 
will apply to the MSIVs.
    Date of issuance: October 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 190.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5671).
    The licensee attached a revised no significant hazards 
consideration (NSHC) determination with the supplement dated September 
11, 2003. This revised NSHC determination contained minor wording 
changes as compared with the NSHC determination sent with the original 
application dated December 16, 2002, changes made to reflect the new TS 
changes, and provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the conclusions of the original NSHC 
determination.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 64143]]

Safety Evaluation dated October 21, 2003.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: May 12, 2003, as supplemented by 
letter dated August 7, 2003.
    Brief description of amendment: The amendment would revise the 
Technical Specifications to remove the MODE restrictions for 
performance of Surveillance Requirements 3.8.4.7 and 3.8.4.8 for the 
Division 3 direct current electrical power subsystem.
    Date of issuance: October 27, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 159.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34665). The August 7, 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 October 27, 2003.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: December 20, 2002, as 
supplemented August 15, 2003.
    Brief description of amendments: The amendments provide editorial 
and administrative changes to the Technical Specifications. The changes 
correct typographical, spelling, numbering syntax, page break, and font 
consistency errors as well as removing blank pages and associated 
references. There are no substantive changes made in the proposed 
amendment.
    Date of issuance: October 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 224 and 219.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5677). The supplemental letter provided clarifying information that was 
within the scope of the initial notice and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in an Environmental Assessment dated October 17, 2003, and a Safety 
Evaluation dated October 21, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 21, 2003.
    Brief description of amendment request: The amendment revises the 
technical specification (TS) administrative controls to make the Three 
Mile Island (TMI) Unit 2 radioactive effluent control program 
consistent with the program for the TMI Unit 1 operating reactor TS. 
The proposed change adopts the TMI Unit 1 liquid discharge limits since 
both Units 1 and 2 use the same liquid discharge monitor and have a 
common discharge pathway. The gaseous discharge limits will also be 
updated to reflect the current 10 CFR 20 nomenclature along with some 
minor editorial changes. Additionally, the definition of a member of 
the public will be made consistent with the definition in 10 CFR 20.
    Date of issuance: October 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 60.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 18, 2003 (68 
FR 54750).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated October 20, 2003.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1, Oswego County, New York

    Date of application for amendment: November 15, 2002, as 
supplemented by letters dated January 15, July 31, and September 15, 
2003.
    Brief description of amendment: The amendment revised the reactor 
coolant system pressure-temperature limit curves and tables in Section 
3/4.2.2, ``Minimum Reactor Vessel Temperature for Pressurization,'' of 
the Technical Specifications. The revised curves and tables are 
effective up to 28 effective full-power years.
    Date of issuance: October 27, 2003.
    Effective date: October 27, 2003, to be implemented within 60 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75882).
    The supplemental letters of January 15, July 31, and September 15, 
2003, provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 27, 2003.
    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: April 30, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification Section 5.3, ``Plant Staff Qualifications,'' to 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 10 CFR Part 55, ``Operators'' 
Licenses.''
    Date of issuance: October 24, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 212.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34670).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: October 17, 2002.
    Brief description of amendments: The amendment revises Technical

[[Page 64144]]

Specification 3.7.9, ``Control Room Emergency Filtration System 
(CREFS),'' by deleting the one-time extension to the allowed outage 
time (AOT) for CREFS and the exception requirements of Limiting 
Condition for Operation 3.04 and Surveillance Requirement 3.04 that 
were allowed during the AOT.
    Date of issuance: October 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 210 and 215.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7818).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 16, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: April 30, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification Section 5.3, ``Plant Staff Qualifications,'' to 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 10 CFR Part 55, ``Operators'' 
Licenses.''
    Date of issuance: October 24, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 211 and 216.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34670).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 24, 2003.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit 2, San Luis Obispo County, California

    Date of application for amendment: June 26, 2003, as supplemented 
by letters dated September 3 and September 30, 2003.
    Brief description of amendments: The amendment authorizes revisions 
to the Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report 
(FSAR) Update to incorporate the NRC approval of a revised steam 
generator (SG) voltage-based repair criteria probability of detection 
(POD) method for DCPP Unit No. 2. The revised POD, based on the 
probability of prior cycle detection method, is approved to determine 
the beginning of cycle voltage distribution for DCPP Unit 2 Cycle 12 
operational assessment.
    Date of issuance: October 21, 2003.
    Effective date: October 21, 2003, and shall be implemented within 
30 days of the date of issuance. The implementation of the amendment 
includes the incorporation into the FSAR Update the changes discussed 
above, as described in the licensee's application dated June 26, 2003, 
and supplements dated September 3 and September 30, 2003, and evaluated 
in the staff's Safety Evaluation attached to the amendment.
    Amendment No.: 164.
    Facility Operating License No. DPR-82: The amendment authorized 
revision of the FSAR Update.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43392). The supplemental letters dated September 3 and September 30, 
2003, provided additional clarifying information, did not expand the 
scope of the application as originally noticed, and did not change the 
NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 21, 2003.
    No significant hazards consideration comments received: No.

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 application for amendments: April 11, 2003, as supplemented 
on August 28 and September 22, 2003.
    Brief description of amendments: The amendments modify the Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications 
(TS) Surveillance Requirements (SRs) 4.3.1.1.3 and 4.3.2.1.3, and TS 
Bases Sections B 3/4.3.1 and B 3/4.3.2 relating to response time 
testing of the Engineered Safety Features Actuation System and the 
Reactor Trip System. In addition, the amendment for Salem, Unit No. 1, 
deletes a footnote associated with SR 4.3.2.1.3, regarding a one-time 
extension to the SR, that is no longer required.
    Date of issuance: October 28, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 260 and 241.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34672).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 28, 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: January 14, 2003, as 
supplemented by letters dated July 1, 2003, and August 20, 2003.
    Brief description of amendment: This amendment revises Technical 
Specification 4.4.5.3.a, maximum inspection interval from 40 calendar 
months to 58 calendar months after two consecutive inspections which 
were classified as C-1.
    Date of issuance: October 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2003 (68 FR 
10280).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 29, 2003.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: March 25, 2003.
    Brief description of amendments: The proposed changes would revise 
Technical Specification 3.5.2, ``Emergency Core Cooling Systems 
(ECCS)--Operating,'' Surveillance Requirement 3.5.2.5. Specifically, 
the changes replace the requirement to verify specific surveillance 
test values for the ECCS pumps with the requirement to verify the 
developed head for each ECCS pump in accordance with the inservice 
testing Program.

[[Page 64145]]

These changes are requested to implement recommendations of the 
Standard Technical Specifications for Combustion Engineering Plants, 
NUREG-1432, Revision 2.
    Date of issuance: October 24, 2003.
    Effective date: October 24, 2003, to be implemented within 60 days 
of issuance.
    Amendment Nos.: Unit 2-190; Unit 3-181.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18285).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 24, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: January 14, 2003 (TS 02-08).
    Brief description of amendment: The proposed amendments revised 
applicability requirements for Technical Specification (TS) 3.3.9.4, 
``Containment Building Penetrations.'' This modified the applicability 
requirement associated with movement of ``irradiated fuel'' by adding a 
new applicability statement for the containment building equipment 
door. The requested also modified the current licensing basis to 
replace the current accident source term used in the design basis fuel 
handling accident radiological analyses with alternate source term.
    Date of issuance: October 28, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 288 and 278.
    Facility Operating License No. DPR-77: Amendments revised the TSs.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7822).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 28, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 3rd day of November 2003.

    For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-28065 Filed 11-10-03; 8:45 am]
BILLING CODE 7590-01-P