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Notice

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

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Information about this document as published in the Federal Register.

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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 Start Printed Page 52234contentions 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 hearingdocket@nrc.gov. 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 OGCMailCenter@nrc.gov. 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 pdr@nrc.gov.

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).

  • Loss-of-Coolant Accident
  • Main Steam Line Break Accident, and
  • 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 Start Printed Page 52235within 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 Start Printed Page 52236thermal 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. Start Printed 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 pdr@nrc.gov.

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 Start Printed Page 52238amendments 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.

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. Start Printed 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. Start Printed 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.

Start Signature

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.

End Signature End Preamble

Footnotes

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.

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[FR Doc. 03-22106 Filed 8-29-03; 8:45 am]

BILLING CODE 7590-01-P