[Federal Register Volume 68, Number 169 (Tuesday, September 2, 2003)]
[Notices]
[Pages 52233-52240]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-22106]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, August 8, 2003, through August 21, 2003. 
The last biweekly notice was published on August 19, 2003, (68 FR 
49812).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By October 2, 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

[[Page 52234]]

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 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].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: April 3, 2003.
    Description of amendment request: The proposed amendment would 
permit application of an alternative source term (AST) methodology, 
according to Section 50.67, ``Accident source term,'' of title 10 of 
the Code of Federal Regulations (10 CFR) with the exception that 
Technical Information Document (TID) 14844, ``Calculation of Distance 
Factors for Power and Test Reactor Sites,'' will continue to be used as 
the radiation dose basis for equipment qualification. The proposed 
amendment would include Technical Specifications (TS) and associated 
Bases revisions to reflect implementation of AST assumptions; TS and 
associated Bases revisions to increase main steam isolation valve 
allowable leakage; TS and associated Bases revisions to decrease 
allowed feedwater isolation valve leakage to allow margin to be used 
for other release paths; TS and associated Bases revisions to delete 
requirements for the main steam isolation valve leakage control system; 
TS and associated Bases revisions to reflect requirements for 
availability of Standby Liquid Control (SLC) System in Mode 3 and use 
of the SLC System to buffer suppression pool pH to prevent iodine re-
evolution during a postulated radiological release; TS and associated 
Bases revisions to reflect higher allowed charcoal adsorber 
penetrations in laboratory testing; TS Bases revision to reflect an 
increased allowed secondary containment drawdown time; TS Bases 
revision to identify additional containment leakage exclusions from 
La and exclusions from secondary containment bypass 
allowances; additional allowance for filtered and unfiltered inleakage 
into the control room envelope; and development of new offsite and 
control room atmospheric dispersion factors calculated using site-
specific meteorology data collected between 2000 and 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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment implements alternative source term (AST) 
assumptions in revisions to the analyses of the following limiting 
design basis accidents at Clinton Power Station (CPS).

[sbull] Loss-of-Coolant Accident
[sbull] Main Steam Line Break Accident, and
[sbull] Control Rod Drop Accident

    The AST does not require modification of the facility; rather, 
once the occurrence of an accident has been postulated the new 
source term is an input to evaluate the potential consequences. The 
implementation of the AST has been evaluated in revisions to the 
analyses of the limiting design basis accidents at CPS. Based upon 
the results of these analyses, it has been demonstrated that, with 
the requested changes, the dose consequences of these limiting 
events is

[[Page 52235]]

within the regulatory guidance provided by the NRC for use with the 
AST. This guidance is presented in 10 CFR 50.67 and associated 
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1.
    The equipment affected by the revised operational conditions is 
not considered an initiator to any previously analyzed accident and 
therefore, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident. The radiological 
consequences of the above design basis accidents have been evaluated 
with applications of AST assumptions. The results conclude that the 
radiological consequences remain within applicable regulatory 
limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The application of AST does not affect the design, functional 
performance or operation of the facility. Similarly, it does not 
affect the design or operation of any structures, systems or 
components involved in the mitigation of any accidents, nor does it 
affect the design or operation of any component in the facility such 
that new equipment failure modes are created.
    As such the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Approval of the basis change from the original source term 
developed in accordance with Technical Information Document (TID) 
14844 to a new AST, as described in Regulatory Guide 1.183, is 
requested. The results of the accident analyses revised in support 
of the proposed changes, and the requested Technical Specification 
changes, are subject to revised acceptance criteria. These analyses 
have been performed using conservative methodologies as specified in 
Regulatory Guide 1.183.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analyses adequately bound postulated event scenarios. The dose 
consequences due to design basis accidents comply with the 
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 
1.183.
    The margin of safety is considered to be that provided by 
meeting the applicable regulatory limits. Relaxation of these 
Technical Specification requirements results in an increase in dose 
following certain design basis accidents. However, since the doses 
following these design basis accidents remain within the regulatory 
limits, there is not a significant reduction in a margin of safety. 
The changes continue to ensure that the doses at the exclusion area 
and low population zone boundaries, as well as the control room, are 
within the corresponding regulatory limits.
    Therefore, operation of CPS in accordance with the proposed 
changes will 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: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 15, 2003.
    Description of amendment request: The licensee proposed to revise 
the reactor coolant system pressure-temperature (P-T) limit curves 
specified in Section 3.4.11, ``RCS [Reactor Coolant System] Pressure 
and Temperature (P/T) Limits,'' of the Technical Specifications (TSs). 
The proposed P-T limit curves will be based, in part, on an alternative 
methodology and will be valid for 22 effective full-power years. The 
alternative methodology, identified as American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code Case N-640, has been 
previously approved for generic use by the Nuclear Regulatory 
Commission (NRC).
    The associated licensee-controlled TSs Bases pages would also be 
changed to reflect the above TS 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. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed changes, if approved by the NRC, will be made 
in a manner such that conservatism is maintained through compliance 
with applicable NRC regulations and guidance. No hardware design change 
is involved with the proposed amendment, thus there will be no adverse 
effect on the functional performance of any plant structure, system, or 
component (SSC). All SSCs will continue to perform their design 
functions with no decrease in their capabilities to mitigate the 
consequences of postulated accidents. P-T limit curves were not 
previously factored into the probability of accidents, nor were they 
factored into scenarios of previously analyzed accidents. Accordingly, 
the revised P-T limit curves will lead to no increase in the 
consequences of an accident previously evaluated, and no increase of 
the probability of an accident previously evaluated.
    The second standard requires that operation of the unit 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 amendment is not the result of a hardware 
design change, nor does it lead to the need for a hardware design 
change. There is no change in the methods the unit is operated. As a 
result, all SSCs will continue to perform as previously analyzed by the 
licensee, and previously evaluated and accepted by the NRC staff. 
Therefore, the proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, the proposed amendment 
will not affect in any way the performance characteristics and intended 
functions of any SSC. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 17, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 3.3.1-2 by modifying a constant in 
the variable

[[Page 52236]]

thermal margin/low pressure (TM/LP) trip equation. The proposed change 
would reduce calculated values for the variable TM/LP trip equation. 
The proposed equation constant value change results from improvements 
in plant equipment used to establish the TM/LP trip setpoint. 
Ultrasonic feedwater flow measurement devices, recently installed at 
the Palisades Plant, result in less uncertainty applied in the 
methodology used for determining core power level. Additionally, the 
devices used to calculate the TM/LP trip setpoint have previously been 
replaced with devices having less uncertainty. These reduced 
uncertainties, when combined using the NRC-endorsed methodology 
described in ANSI/ISA-S67.04-1994, ``Setpoints for Nuclear Safety-
Related Instrumentation,'' result in a reduction in the constant (bias 
term) used to calculate the TM/LP trip setpoint.
    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 amendment does not involve operation of any 
required structures, systems or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated. The methodology that was used in determining the 
recommended change in the constant follows Nuclear Regulatory 
Commission endorsed standard ANSI/ISA-S67.04-1994, ``Setpoints for 
Nuclear Safety-Related Instrumentation.'' The probability of an 
accident previously evaluated will not be increased since the 
proposed change to the constant value in the Thermal Margin/Low 
Pressure (TM/LP) trip equation maintains all necessary 
considerations in the development of uncertainties.
    The consequences of an accident previously evaluated will not be 
increased since the reactor is still protected from violating the 
TM/LP trip setpoint used in the safety analysis for the Palisades 
Nuclear Plant.
    Therefore, operation of the facility in accordance with the 
proposed change to the Technical Specifications 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 change to the constant value for the TM/LP trip 
equation in the Technical Specifications would not change or add a 
system function. 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 change being requested.
    Therefore, 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 change to the constant value for the TM/LP trip 
equation in the Technical Specifications accounts for all 
uncertainties that affect the TM/LP trip setpoint. The revised TM/LP 
trip setpoint will continue to assure that the acceptance criteria 
established in the safety analysis will be met.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: July 24, 2003.
    Description of amendment requests: The proposed change will revise 
Technical Specification (TS) Section 3.8.4, ``DC Sources--Operating''; 
TS Section 3.8.5, ``DC Sources--Shutdown''; and TS Section 3.8.6, 
``Battery Cell Parameters.'' The proposed change will also add a new 
section to TS 5.5, ``Programs and Manuals'' for the maintenance and 
monitoring of the station safety-related batteries that is based on the 
recommendations of the Institute of Electrical and Electronics 
Engineers (IEEE) Standard 450-1995.
    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 affects Technical Specification (TS) 
sections 3.8.4 ``DC Sources--Operating,'' TS 3.8.5 ``DC Sources--
Shutdown,'' TS 3.8.6 ``Battery Cell Parameters,'' and TS 
Administrative Controls section 5.5.
    The proposed change restructures the TS for the direct current 
(DC) electrical power subsystem and adds new Conditions and Required 
Actions with increased Completion Times to address battery charger 
inoperability. Neither the DC electrical power subsystem nor 
associated battery chargers are initiators of any accident sequence 
analyzed in the Final Safety Analysis Report Update (FSARU). 
Operation in accordance with the proposed TS ensures that the DC 
electrical power subsystem is capable of performing its function as 
described in the FSARU. Therefore the mitigating functions supported 
by the DC electrical power subsystem will continue to provide the 
protection assumed by the analysis.
    The relocation of preventive maintenance surveillances, and 
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.17, ``Battery Monitoring and Maintenance 
Program,'' will not challenge the ability of the DC electrical power 
subsystem to perform its design function. The maintenance and 
monitoring required by current TS, which are based on industry 
standards, will continue to be performed. In addition, the DC 
electrical power subsystem is within the scope of 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' which will ensure the control of maintenance 
activities associated with the DC electrical power subsystem.
    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 does not involve any physical alteration of 
the units. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints at which protective or mitigating actions are 
initiated that are affected by the proposed changes. The operability 
of the DC electrical power subsystems in accordance with the 
proposed TS is consistent with the initial assumptions of the 
accident analyses and is based upon meeting the design basis of the 
plant. The proposed change will not alter the manner in which 
equipment operation is initiated, nor will the functional demands on 
credited equipment be changed. No alteration in the operating 
procedures, which ensure the unit remains within analyzed limits, is 
proposed, and no change is being made to procedures relied upon to 
respond to an off-normal event. As such, no new failure modes are 
being introduced. The proposed change does not alter assumptions 
made in the safety analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.

[[Page 52237]]

    The proposed change will not adversely affect operation of plant 
equipment and will not result in a change to the setpoints at which 
protective actions are initiated. Sufficient DC capacity to support 
operation of mitigation equipment is ensured. The changes associated 
with the new battery maintenance and monitoring program will ensure 
that the station batteries are maintained in a highly reliable 
manner. The equipment fed by the DC electrical system will continue 
to provide adequate power to safety-related loads in accordance with 
analysis assumptions.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (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 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. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: December 20, 2002, as 
supplemented by letter dated May 30, 2003.
    Brief description of amendment: The amendment approves changes to 
the Clinton facility as described in the Updated Safety Analysis 
Report. The amendment modifies the basis for compliance with the 
requirements of Appendix H to title 10 of the Code of Federal 
Regulations part 50 (appendix H to 10 CFR part 50), ``Reactor Vessel 
Material Surveillance Program Requirements,'' by approving 
implementation of the Boiling-Water Reactor Vessel and Internals 
Project reactor pressure vessel integrated surveillance program.
    Date of issuance: August 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 157.
    Facility Operating License No. NPF-62: The amendment approved 
revisions to the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669). The supplemental letter 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 August 12, 2003.
    No significant hazards consideration comments received: No.

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: September 30, 2002, as 
supplemented by letter dated March 19, 2003.
    Brief description of amendment: The amendment revised Technical 
Specification Section 6.8.5, ``Reactor Building Leakage Rate Testing 
Program,'' to reflect a one-time deferral of the scheduled performance 
of the next Type A Containment Integrated Leak Rate Test from October, 
2003, to no later than September 2008.
    Date of issuance: August 14, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 244.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68730). The supplement 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 August 14, 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: April 25, 2003.
    Brief description of amendments: The amendments revise Section 5.3, 
``Unit Staff Qualifications,'' of the Technical Specifications to state 
new education and experience eligibility requirements for operator 
license applicants. As stated in the letter dated April 25, 2003, the 
new requirements are outlined by the National Academy for Nuclear 
Training in its ``Guidelines for Initial Training and Qualification of 
Licensed Operators,'' which were issued January 2000.
    Date of issuance: August 13, 2003.
    Effective date: August 13, 2003, and shall be implemented within 90 
days of the date of issuance.
    Amendment Nos.: Unit 1-148, Unit 2-148, Unit 3-148.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

[[Page 52238]]

amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34662). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 13, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-003, Indian Point 
Nuclear Generating Station, Unit 1

    Date of amendment request: May 30, 2002.
    Brief description of amendment: It would revise the Indian Point 
Nuclear Generating Station, Unit 1 (IP1) Technical Specifications (TSs) 
to facilitate the Indian Point Generating Station, Unit 2 (IP2) 
transition to the Improved TSs. The amendment also revises the 
requirements of the ``Order Approving Decommissioning Plan and 
Authorizing Decommissioning of Facility'' \1\ to ensure compliance with 
the current requirements of 10 CFR 50.59 and 10 CFR 50.83. It also 
revises the expiration date of Provisional Operating License No. DPR-5 
for IP1 to be current with the expiration date for the Facility 
Operating License No. DPR-26 for IP2.
---------------------------------------------------------------------------

    \1\ NRC letter to Consolidated Edison, ``Order to Authorize 
Decommissioning and Amendment No. 45 to License No. DPR-5 for Indian 
Point Unit 1 (TAC No. M59664),'' dated January 31, 1996.
---------------------------------------------------------------------------

    Date of issuance: August 11, 2003.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No: 52.
    Provisional Operating License No. DPR-5: The amendment revised the 
Technical Specifications, and made changes to and revised the 
expiration date for IP1 Provisional Operating License DPR-5.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45564).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2003.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: September 19, 2002, as 
supplemented by letters dated January 8, May 22, and July 1, 2003.
    Brief description of amendment: The amendment extends the allowable 
outage time for the emergency diesel generators from 72 hours to a 
maximum of 14 days.
    Date of issuance: August 8, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 249.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68733). The January 8, May 22, and July 1, 2003, supplemental 
letters provided clarifying information that did not change the scope 
of the original Federal Register notice or the original no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 8, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: August 7, 2002, as supplemented 
by your letters dated February 28, and May 27, 2003.
    Brief description of amendments: The amendments revised the 
limiting condition for operation, the associated Conditions and 
Required Actions of Technical Specification (TS) 3.7.1, ``Main Steam 
Safety Valves (MSSVs),'' and the values in Table 3.7.1-1, ``Operable 
Main Steam Safety Valves versus Applicable Power in Percent of Rated 
Thermal Power,'' by requiring five MSSVs per steam generator to be 
operable consistent with the accident analyses assumptions. The 
amendments also modify the associated Required Actions of TS 3.7.1 by 
adding a requirement to reduce the Power Range Neutron Flux-High 
reactor trip setpoint when one or more steam generators with one or 
more MSSVs are inoperable.
    Date of issuance: August 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 133/133, 128/128.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61681). 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 amendments is contained 
in a Safety Evaluation dated August 12, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 20, 2002, as 
supplemented by letters dated May 30, and June 27, 2003.
    Brief description of amendments: The amendments approve changes to 
the LaSalle County Station facility as described in the Updated Final 
Safety Analysis Report. The amendments modify the basis for compliance 
with the requirements of appendix H to title 10 of the Code of Federal 
Regulations part 50 (appendix H to 10 CFR part 50), ``Reactor Vessel 
Material Surveillance Program Requirements,'' by approving 
implementation of the Boiling-Water Reactor Vessel and Internals 
Project reactor pressure vessel integrated surveillance program.
    Date of issuance: August 13, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 160/146.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
approve revisions to the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669). 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 amendments is contained 
in a Safety Evaluation dated August 13, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: November 30, 2001.
    Brief description of amendment: This amendment revises Technical 
Specification 3/4.4.4, ``Reactor Coolant System--Pressurizer,'' to 
adopt a new pressurizer high-level limit and to revise the required 
action when the pressurizer is inoperable.
    Date of issuance: August 12, 2003.

[[Page 52239]]

    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 255.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37578). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: May 21, 2003.
    Brief description of amendment: This amendment relocates to the 
Technical Requirements Manual the Technical Specification surveillance 
requirement pertaining to flow balance testing of the emergency core 
cooling system (ECCS) high pressure injection and low pressure 
injection subsystems following system modifications that alter 
subsystem flow characteristics. Also, the amendment adds an ECCS pump 
operability requirement to the Technical Specifications.
    Date of issuance: August 12, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 256.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34669). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: May 30, 2003.
    Brief description of amendment: The amendment deletes technical 
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminates the requirements to have and maintain the post accident 
sampling system (PASS) at the Duane Arnold Energy. The amendment also 
addresses related changes to TS 5.5.2, ``Primary Coolant Sources 
Outside Containment.''
    Date of issuance: August 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 252.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40713).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 8, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 8, 2002.
    Brief description of amendment: The October 8, 2002, submittal 
proposed the following: (1) The use of a pressure temperature limits 
report (PTLR), (2) change the minimum boltup temperature, (3) revise 
the low temperature overpressure protection (LTOP) methodology and 
analysis, (4) perform the LTOP analyses ``in-house,'' (5) change the 
LTOP enable temperature, (6) modify TS 2.10.1 to exactly specify the 
reactor coolant system (RCS) temperature at which the reactor can be 
made critical, and (7) add a TS for a maximum pressure value for the 
safety injection tanks. This amendment approves the use of a PTLR for 
the Fort Calhoun Station. As such TS Figure 2-1 (RCS Pressure--
Temperature Limits for Heatup, Cooldown, and In-service Test) will be 
relocated into Figure 5-1 of the PTLR. In addition, the following TSs 
were either modified or added for the implementation of the PTLR: 
define the PTLR in Definitions; TS 2.1.1(8); TS 2.1.1(11); TS 2.1.2 and 
2.1.2 References; TS 2.1.6(4); TS 2.3(1)(c); TS 2.3(3); TS 2.3 
References; TS 2.10.1; Table 3-5, item 23, TS 3.3(1)(c); and TS 5.9.6. 
The following TS Bases sections were modified to reflect the 
implementation of the PTLR: TS 2.1.1, TS 2.1.2, and TS 2.10.1.
    Date of issuance: August 15, 2003.
    Effective date: August 15, 2003. The amendment shall be implemented 
within 30 days from the date of issuance, including submitting the 
first Pressure Temperature Limits Report to the NRC Document Control 
Desk with copies to the Region IV Regional Administration and Resident 
Inspector.
    Amendment No.: 221.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37579). The April 10, June 4, July 31, and August 5, 2003, supplemental 
letters 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 August 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 8, 2002, as supplemented by 
letters dated April 11 and May 21, 2003.
    Brief description of amendment: The amendment grants a one-time 
five-year extension to the current ten-year test interval for the 
containment integrated leak rate testing.
    Date of issuance: August 15, 2003.
    Effective date: August 15, 2003, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 220.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68742). The April 11 and May 21, 2003, supplemental letters 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 August 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 5, 2003.
    Brief description of amendments: The amendments extend from 1 hour 
to 24 hours the completion time for Condition B of Technical 
Specification 3.5.1, which defines requirements for the restoration of 
an emergency core cooling system accumulator when it has been declared 
inoperable for a reason other than boron concentration.
    Date of issuance: August 15, 2003.
    Effective date: August 15, 2003, and shall be implemented within 60 
days from the date of issuance.
    Amendment Nos.: Unit 1--160; Unit 2--161.

[[Page 52240]]

    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40716). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 2003.
    No significant hazards consideration comments received: No.

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: March 31, 2003.
    Brief description of amendments: The amendments replace ``Central 
Power and Light Company (CPL)'' with ``AEP Texas Central Company'' 
throughout the Operating License of each unit.
    Date of issuance: August 11, 2003.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--155; Unit 2--143.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34673). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 11, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 13, 2002, as 
supplemented May 19 and July 11, 2003.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.7.2.12, ``Steam Generator (SG) Tube Surveillance 
Program.'' The revised TS allows the use of Westinghouse leak-limiting 
Alloy 800 sleeves to repair defective SG tubes as an alternative to 
plugging the tube.
    Date of issuance: August 15, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 44.
    Facility Operating License No. NPF-90: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 
(68FR12958). The supplemental letters provided clarifying information 
that did not expand the scope of the original request and 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 August 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 5, 2002.
    Brief Description of amendments: These amendments delete the 
requirement to perform a 15-minute degassed beta and gamma activity 
test of the secondary coolant and require that the dose equivalent I-
131 analysis be performed on a more conservative monthly basis.
    Date of issuance: August 15, 2003.
    Effective date: August 15, 2003.
    Amendment Nos.: 234 and 233.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78525). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 2003.
    No significant hazards consideration comments received: No.

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

    For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-22106 Filed 8-29-03; 8:45 am]
BILLING CODE 7590-01-P