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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, September 19, 2003, through October 2, 2003. The last biweekly notice was published on September 30, 2003 (68 FR 56340).

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 Start Printed Page 59213take 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 November 13, 2003, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to 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, Start Printed Page 59214Public 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.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: July 14, 2003.

Description of amendment request: The proposed change involves the extension from 1 hour to 24 hours of the completion time (CT) for Condition B of Technical Specification (TS) 3.5.1, which defines requirements for accumulators. Accumulators are part of the emergency core cooling system and consist of tanks partially filled with borated water and pressurized with nitrogen gas. The contents of the tank are discharged to the reactor coolant system if, as during a loss-of-coolant accident, the coolant pressure decreases to below the accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore an accumulator to operable status when it has been declared inoperable for a reason other than the boron concentration of the water in the accumulator not being within the required range. This change was proposed by the Westinghouse Owners Group participants in the Technical Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is supported by NRC-approved topical report WCAP-15049-A, “Risk-Informed Evaluation of an Extension to Accumulator Completion Times,” submitted on May 18, 1999. The NRC staff issued a notice of opportunity for comment in the Federal Register on July 15, 2002 (67 FR 46542), on possible amendments concerning TSTF-370, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on March 12, 2003 (68 FR 11880). The licensee affirmed the applicability of the following NSHC determination in its application dated July 14, 2003.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The basis for the accumulator limiting condition for operation (LCO), as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-A, the proposed change will allow plant operation with an inoperable accumulator for up to 24 hours, instead of 1 hour, before the plant would be required to begin shutting down. The impact of the increase in the accumulator CT on core damage frequency for all the cases evaluated in WCAP-15049-A is within the acceptance limit of 1.0E-06/yr for a total plant core damage frequency (CDF) less than 1.0E-03/yr. The incremental conditional core damage probabilities calculated in WCAP-15049-A for the accumulator CT increase meet the criterion of 5E-07 in Regulatory Guides (RG) 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” and 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” for all cases except those that are based on design basis success criteria. As indicated in WCAP-15049-A, design basis accumulator success criteria are not considered necessary to mitigate large break loss-of-coolant accident (LOCA) events, and were only included in the WCAP-15049-A evaluation as a worst case data point. In addition, WCAP-15049-A states that the NRC has indicated that an incremental conditional core damage frequency (ICCDP) greater than 5E-07 does not necessarily mean the change is unacceptable.

The proposed technical specification change does not involve any hardware changes nor does it affect the probability of any event initiators. There will be no change to normal plant operating parameters, engineered safety feature (ESF) actuation setpoints, accident mitigation capabilities, accident analysis assumptions or inputs.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. As described in Section 9.1 of the WCAP-15049-A evaluation, the plant design will not be changed with this proposed technical specification CT increase. All safety systems still function in the same manner and there is no additional reliance on additional systems or procedures. The proposed accumulator CT increase has a very small impact on core damage frequency. The WCAP-15049-A evaluation demonstrates that the small increase in risk due to increasing the CT for an inoperable accumulator is within the acceptance criteria provided in RGs 1.174 and 1.177. No new accidents or transients can be introduced with the requested change and the likelihood of an accident or transient is not impacted.

The malfunction of safety related equipment, assumed to be operable in the accident analyses, would not be caused as a result of the proposed technical specification change. No new failure mode has been created and no new equipment performance burdens are imposed.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change does not involve a significant reduction in a margin of safety. There will be no change to the departure from nucleate boiling ratio (DNBR) correlation limit, the design DNBR limits, or the safety analysis DNBR limits.

The basis for the accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-A, the proposed change will allow plant operation with an inoperable accumulator for up to 24 hours, instead of 1 hour, before the plant would be required to begin shutting down. The impact of this on plant risk was evaluated and found to be very small. That is, increasing the time the accumulators will be unavailable to respond to a large LOCA event, assuming accumulators are needed to mitigate the design basis event, has a very small impact on plant risk. Since the frequency of a design basis large LOCA (a large LOCA with loss of offsite power) would be significantly lower than the large LOCA frequency of the WCAP-15049-A evaluation, the impact of increasing the accumulator CT from 1 hour to 24 hours on plant risk due to a design basis large LOCA would be significantly less than the plant risk increase presented in the WCAP-15049-A evaluation.

Therefore, this change does not involve a significant reduction in a margin of safety.

Start Printed Page 59215

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: October 16, 2001; as supplemented by letters dated May 20, September 12, and November 21, 2002; and January 27, and September 22, 2003.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) to incorporate changes resulting from the use of an alternate source term (AST) and the implementation of several plant modifications. Publication of the Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing for the October 16, 2001, submittal appeared in the Federal Register on January 22, 2002, (67 FR 2922). The September 22, 2003, submittal contained a revised No Significant Hazards Consideration Determination. The September 22, 2003, submittal includes (1) Implementing the AST for accident analysis as described in Regulatory Guide 1.183; (2) relaxing the TS for the penetration room ventilation system (PRVS) and the spent fuel pool ventilation system (SFPVS) because these systems are no longer credited for control room and offsite doses; (3) revising the control room ventilation system (CRVS) to allow for a one-time completion extension to support implementation of the control room intake/booster fan modification; (4) lowering the reactor building leakage rate from 0.25 weight percent per day to 0.20 weight percent per day; (5) revising the ventilation filter testing program radioactive methyl iodide removal acceptance criterion for PRVS, SFPVS, and CRVS booster fan trains; and (6) adoption of TS Task Force (TSTF)-51.

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 amendment will not involve a significant increase in the probability of consequences of an accident previously evaluated.

The AST [alternate source term] and those plant systems affected by implementing the proposed changes to the TS [technical specifications] are not assumed to initiate design basis accidents. The AST does not affect the design or operations of the facility. Rather, the AST is used to evaluate the consequences of a postulated accident. The implementation of the AST has been evaluated in the revisions to the analysis of the design basis accident for ONS [Oconee Nuclear Station]. Based on the results of these analyses, it has been demonstrated that, with the requested changes, the dose consequences of these events meet the acceptance criteria of 10 CFR 50.67 and RG [Regulatory Guide] 1.183. Therefore, the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The AST and those plant systems affected by implementing the proposed changes to the TS are not assumed to initiate design basis accidents. The systems affected by the changes are used to mitigate the consequences of an accident that has already occurred. The proposed TS changes and modifications do not significantly affect the mitigative function of these systems. Consequently, these systems do not alter the nature of events postulated in the Safety Analysis Report nor do they introduce any unique precursor mechanisms. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) The proposed amendment will not involve a significant reduction in the margin of safety.

The implementation of the AST, proposed changes to the TS and implementation of the proposed modifications have been evaluated in the revisions to the analysis of the consequences of the design basis accidents for the ONS. Based on the results of these analyses, it has been demonstrated that with the requested changes the dose consequences of these events meet the acceptance criteria of 10 CFR 50.67 and following the provisions of RG 1.183. Thus, the proposed amendment will not involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 17th Street, NW., Washington, DC 20005.

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: July 29, 2003.

Description of amendment request: The proposed amendments would allow the licensee to modify technical specifications (TS) to be consistent with Technical Specification Task Force (TSTF) Traveler TSTF-360, Revision 1, “DC Electrical Rewrite,” and to implement new actions for inoperable battery chargers, modify certain actions and surveillance requirements, relocate certain surveillance requirements to a licensee controlled program, and create an administrative program for battery monitoring and maintenance to be referenced in the TS.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes restructure the Technical Specifications (TS) for the direct current (DC) electrical power system. The proposed changes add actions to specifically address battery charger inoperability. This change will rely upon the capability of providing the battery charger function by an alternate means (e.g., a 125 volts direct current (VDC) portable battery charger or a 250 VDC portable battery charger) to justify the proposed Completion Times. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the Updated Final Safety Analysis Report (UFSAR). Operation in accordance with the proposed TS ensures that the DC electrical power system is capable of performing its function as described in the UFSAR. Therefore, the mitigative functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis.

The relocation of preventive maintenance surveillance, and certain operating limits and actions, to a newly-created licensee controlled Battery Monitoring and Maintenance Program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC electrical power system 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 system. The integrity of fission product barriers, plant configuration, Start Printed Page 59216and operating procedures as described in the UFSAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes involve restructuring the TS for the DC electrical power system. This change will rely upon the capability of providing the battery charger function by an alternate means (e.g., a swing charger or a portable battery charger) to justify the proposed Completion Times when a normal battery charger is inoperable. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the UFSAR. Rather, the DC electrical power system is used to supply equipment used to mitigate an accident.

The 125 VDC portable battery charger will be utilized as a common spare to feed the Division I or Division 2 125 VDC bus of Unit 2 or Unit 3. For the 250 VDC system, a full capacity swing charger is available for use between the units, and can be aligned to any one of the 250 VDC batteries. In addition, the 250 VDC portable battery charger can be utilized as a common spare to feed the 250 VDC safety related batteries of Unit 2 or Unit 3. This portable charger is identical to the existing chargers and is non-safety related. The output of the portable charger will be capable of being connected to any one of the Class IE DC buses for Division I or Division 2 of Unit 2 or Unit 3. Allowing the use of a portable spare and swing battery chargers will increase the reliability of the DC electrical power system. The mitigative functions supported by the DC electrical power system will continue to provide the protection assumed by the safety analyses described in the UFSAR. Therefore, there are no new types of failures that could be created by a failure of the portable battery charger. As such, no new or different kind of accident or transient is expected by these changes.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed changes do not involve a significant reduction in a margin of safety.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment. These changes 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 use of a portable battery charger will increase the reliability of the DC system during periods of normal battery charger inoperability. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.

Attorney for licensee: Senior Counsel, Nuclear; Exelon Generation Company LLC; 4300 Winfield Road; Warrenville, IL 60555.

NRC Section Chief: Anthony J. Mendiola.

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

Date of application for amendments: July 14, 2003.

Description of amendment request: The proposed change is requested to support application of an alternative source term methodology, with the exception that Technical Information Document 14844, “Calculation of Distance Factors for Power and Test Reactor Sites,” will continue to be used as the radiation dose basis for equipment qualification.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The implementation of alternative source term (AST) assumptions has been evaluated in revisions to the analyses of the following limiting design basis accidents (DBAs) at Peach Bottom Atomic Power Station (PBAPS):

  • Loss-of-Coolant Accident,
  • Main Steam Line Break Accident,
  • Fuel Handling Accident, and
  • Control Rod Drop Accident.

Based upon the results of these analyses, it has been demonstrated that, with the requested changes, the dose consequences of these limiting events are 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 Alternative Source Term is an input to calculations used to evaluate the consequences of an accident, and does not by itself affect the plant response, or the actual pathway of the radiation released from the fuel. It does however, better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. Therefore, the consequences of an accident previously evaluated are not significantly increased.

The equipment affected by the proposed changes is mitigative in nature, and relied upon after an accident has been initiated. Application of the Alternative Source Term (AST) does not involve any physical changes to the plant design. While the operation of various systems do change as a result of these proposed changes, these systems are not accident initiators. Application of the AST is not an initiator of a design basis accident. The proposed changes to the Technical Specifications (TS), while they revise certain performance requirements, do not involve any physical modifications to the plant. As a result, the proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of any accidents. As such, removal of operability requirements during the specified conditions will not significantly increase the probability of occurrence for an accident previously analyzed. Since design basis accident initiators are not being altered by adoption of the Alternative Source Term analyses, the probability of an accident previously evaluated is not affected.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed amendment does not involve a physical alteration of the plant (no new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed changes). Similarly, it does not physically change any structures, systems or components involved in the mitigation of any accidents, thus, no new initiators or precursors of a new or different kind of accident are created. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed amendment.

As such the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

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 proposed amendment is associated with the implementation of a new licensing basis for PBAPS Design Basis Accidents (DBAs). Approval of the change from the Start Printed Page 59217original source term to a new source term taken from Regulatory Guide 1.183 is being requested. The results of the accident analyses, revised in support of the proposed license amendment, are subject to revised acceptance criteria. The analyses have been performed using conservative methodologies, as specified in Regulatory Guide 1.183. Safety margins have been evaluated and analytical conservatism has been utilized to ensure that the analyses adequately bound the postulated limiting event scenario. The dose consequences of these DBAs remain within the acceptance criteria presented in 10 CFR 50.67, “Accident Source Term”, and Regulatory Guide 1.183.

The proposed changes continue to ensure that the doses at the exclusion area boundary (EAB) and low population zone boundary (LPZ), as well as the Control Room, are within corresponding regulatory limits.

Therefore, operation of PBAPS 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: Mr. Edward Cullen, Vice President and General Counsel, Exelon Generation Company, LLC, 2301 Market Street, S23-1, Philadelphia, PA 19101.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: September 12, 2003.

Description of amendment request: The proposed amendment is for relaxation of the heater acceptance criteria contained in Surveillance Requirement (SR) 4.6.6.1d.5, SR 4.7.6.1d.3, and SR 4.7.7d.4 for the shield building ventilation, control room ventilation, and controlled ventilation area systems, respectively. These SRs are performed to verify that heat dissipated by the heaters is within a given band. The requested change is to increase the upper limit of the acceptance criteria from rated capacity plus 5 percent (%) to rated capacity plus 10%. No change is proposed for the lower limit of the band of rated capacity minus 10%.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The relaxation of the SR acceptance criteria to increase the operating band does not alter the way plant equipment is designed or operated. The ESF [engineered safety feature] filtration unit heating coils will continue to reduce the humidity of the incoming air to 70% relative humidity or below. In addition, the air temperature will continue to be controlled such that additional iodine will not be released into the environment. Thus, the charcoal adsorber will continue to meet its design basis and its efficiency will not be adversely affected. The effect of the higher heat dissipation has also been evaluated and the ignition temperature of the charcoal adsorbers is not approached with flow through the systems. In addition, the impact of the new acceptance criterion was determined not to impact the loading or fuel consumption of the emergency diesel generators.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The relaxation of the SR acceptance criteria to increase the operating band does not alter the way plant equipment is designed, operated, or tested. No possibility for a new or different accident or failure mode is introduced by modifying the SR acceptance criteria. The proposed change does not affect the functional capability of safety-related equipment.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The ESF filtration unit heating coils will continue to reduce the humidity of the incoming air to 70% relative humidity or below. Thus, the efficiency of the charcoal adsorber will not be adversely affected. In addition, the impact of the new acceptance criterion was determined not to impact the loading or fuel consumption of the emergency diesel generators. Therefore, the systems have the same capabilities to mitigate accidents as they had prior to the SR acceptance criteria change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

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

Date of amendment requests: August 27, 2003.

Brief description of amendments: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated August 27, 2003.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed Start Printed Page 59218by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: L. Raghavan.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: August 25, 2003.

Description of amendment request: The proposed license amendment request would revise Technical Specification (TS) 3.5.1 to incorporate TS Task Force 318 for one Low Pressure Coolant Injection (LPCI) pump inoperable in each of the two Emergency Core Cooling Systems (ECCS) divisions.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change does not affect the LPCI subsystem design or function. The change to TS 3.5.1 Condition A with one LPCI pump inoperable in both subsystems is more reliable than the current configuration allowed by Condition A. The current TS actions require entry into shutdown LCO [Limiting Condition for Operation] 3.0.3 for this condition. In addition, for an event that does not impact LPCI availability the change provides for more injection flow than the current TS 3.5.1 Condition A LPCI pump configuration. Review of Updated Safety Analysis Report Section XIV-6.0 “Analysis of Design Basis Accidents” confirms that the LPCI mode of the Residual Heat Removal system is not assumed to be the initiator of any previously analyzed event.

Based on the above, NPPD concludes that the proposed TS change to TS 3.5.1 Condition A does not significantly increase the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical change to the plant, add any new equipment or require any existing equipment to be operated in a manner different from the present system design.

Based on the above, NPPD concludes that the proposed TS change to TS 3.5.1 Condition A does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?

The proposed TS change will not reduce the margin of safety. The proposed configuration of one LPCI pump in each LPCI subsystem represents a more reliable configuration. The current TS actions require entry into shutdown LCO 3.0.3 for this condition. In addition, for an event that does not impact LPCI availability the change provides for more injection flow than the current [LCO] requirement which only allows two LPCI pumps in one ECCS subsystem to be inoperable for seven days.

Based on the above, NPPD concludes that the proposed TS change to TS 3.5.1 Condition A does not involve a significant reduction in the margin of safety.

From the above discussions, NPPD concludes that the proposed amendment involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: September 18, 2003.

Description of amendment request: The proposed amendment would revise the limiting condition for operation (LCO) and the associated surveillance requirements of Technical Specification 3.4.1, “[Primary Coolant System] PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to reflect relocation of the DNB limits from the TSs to the Core Operating Limits Report (COLR). These DNB limits are for pressurizer pressure, PCS cold leg temperature, and PCS total flow rate. The proposed amendment would also revise paragraph a of TS 5.6.5, “Core Operating Limits Report (COLR),” to reflect the addition of “DNB Limits” to the COLR. In addition, LCO 3.4.1 would be added to items 16 and 17 in TS 5.6.5b, which lists the documents approved by the NRC for the analytical methods for which the licensee is to use the latest revisions to determine the core operating limits.

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? Start Printed Page 59219

Response: No

The proposed amendment relocates the primary coolant system (PCS) departure from nucleate boiling (DNB) limits to the core operating limits report (COLR) and does not involve any change to the PCS DNB limits themselves. 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 Nuclear Regulatory Commission (NRC) has approved all the analytical methods described in Technical Specification (TS) section 5.6.5, “Core Operating Limits Report (COLR).” Relocation of the PCS DNB limits to the COLR will maintain existing operating fuel cycle analysis requirements. Any future revisions to the safety analyses that require prior NRC approval are identified per the 10 CFR 50.59 review process.

Therefore, the probability of an accident previously evaluated will not be increased by the proposed change.

The consequences of an accident previously evaluated will not be increased since the reactor is still protected from violating the PCS DNB parameters used in the safety analysis for Palisades Nuclear Plant.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment to relocate the PCS DNB limits to the COLR 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 proposed change.

Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment to relocate the PCS DNB limits to the COLR will continue to assure that the acceptance criteria established in the safety analysis will be met. The safety analyses of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on DNB related parameters ensure that these parameters, when appropriate measurement uncertainties are applied, will not be less conservative than those assumed in the safety analyses and thereby provide assurance that the minimum departure from nucleate boiling ratio (DNBR) will meet the required criteria for each of the analyzed transients. The proposed amendment does not change the existing PCS DNB limits. Any future revisions to the safety analyses that require prior NRC approval are identified per the 10 CFR 50.59 review process.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Section Chief: L. Raghavan.

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

Date of amendment requests: September 15, 2003.

Description of amendment requests: In Technical Specification (TS) 2.0, “Safety Limits (SLs),” Reactor Core SL 2.1.1.2, the proposed change would replace the peak linear heat rate SL with a peak fuel centerline temperature SL. This change is requested so SL 2.1.1.2 adequately conforms to 10 CFR 50.36(c)(1)(ii)(A), which requires that Limiting Safety System Settings prevent a Safety Limit from being exceeded.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change does not require any physical change to any plant systems, structures, or components nor does it require any change in systems or plant operations. The proposed change does not require any change in safety analysis methods or results. The change to establish the PFCT [Peak Fuel Centerline Temperature] as the SL is consistent with the Standard Review Plan (SRP) and the SONGS Units 2 and 3 licensing basis for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The SONGS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analysis for Anticipated Operational Occurrences (AOOs) where the peak linear heat rate may exceed the existing Safety Limit of 21 KW/ft is the Control Element Assembly (CEA) Withdrawal at subcritical and low power startup conditions.

The accident analyses indicate that the peak linear heat rate may exceed the Limiting Safety System Setpoint of 21 KW/ft during Control Element Assembly Withdrawal Events at Subcritical and Hot Zero Power conditions. The analyses for these AOOs indicate that the PFCT is not approached or exceeded. The existing analyses remain unchanged and do not affect any accident initiators that would create a new accident.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not require any change in accident analysis methods or results. Therefore, by changing the SL from PLHR [Peak Linear Heat Rate] to Peak Fuel Centerline Temperature, the margin as established in the current license basis remains unchanged.

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: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), Unit No. 1, Fairfield County, South Carolina

Date of amendment request: July 29, 2003.

Description of amendment request: The proposed change will revise Surveillance Requirement 4.0.5 to reflect the deletion of Subsections IWP and IWV from Section XI of the 2000 Addenda of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. This change will also result in revising the Technical Specification (TS) Bases for 4.0.5, 3/4.4.2 and 3/4.4.6 to reflect the applicability of the Code for Operation and Maintenance of Nuclear Power Plants (OM Code) to inservice testing activities. TS 4.0.5 is also being revised as recommended by NUREG-1492, “Guidelines for Inservice Testing at Nuclear Power Plants,” April 1995. Start Printed Page 59220

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change to TS 4.0.5 reflects NRC approval of the ASME Code [2000 Adenda], in 10CFR50.55a, for the conduct of Inservice Testing (IST). The current TS references use of ASME Section XI for this testing, which will no longer be applicable for the third IST interval. The adoption of an NRC approved test code, as required by 10CFR50.55a(f)(4)(ii) will not increase the probability of an accident previously evaluated. Testing is performed to ensure the operational readiness of pumps and valves to perform their safety functions.

The probability or consequences of accidents previously evaluated in the VCSNS FSAR [Final Safety Analysis Report] are unaffected by this proposed change because there is no change to any equipment response or accident mitigation scenario. There are no additional challenges to fission product barrier integrity. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change involves the adoption of an NRC approved Inservice Testing Code for the conduct of Operating License mandated testing. The adoption of the new Code is required to satisfy 10CFR50.55a(f)(4)(ii). The new Code enhances plant safety by requiring the bi-directional testing of check valves and comprehensive pump testing. These changes were incorporated to better monitor pumps and check valves for degradation. The adoption of the new Code does not create the possibility of a new or different kind of accident or malfunction.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does this change involve a significant reduction in margin of safety?

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. A change to the surveillance requirement is proposed, but the ASME OM Code is an NRC approved standard incorporating inservice testing enhancements not contained in ASME Section XI.

Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed Technical Specifications change poses no significant hazard as delineated by 10 CFR 50.92.

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: Thomas G. Eppink, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: John A. Nakoski.

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

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

Date of amendment request: September 2, 2003.

Description of amendment request: The proposed change involves the extension from 1 hour to 24 hours of the completion time (CT) for Condition B of Technical Specification (TS) 3.5.1, which defines requirements for accumulators. Accumulators are part of the emergency core cooling system and consist of tanks partially filled with borated water and pressurized with nitrogen gas. The contents of the tank are discharged to the reactor coolant system if, as during a loss-of-coolant accident, the coolant pressure decreases to below the accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore an accumulator to operable status when it has been declared inoperable for a reason other than the boron concentration of the water in the accumulator not being within the required range. This change was proposed by the Westinghouse Owners Group participants in the Technical Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is supported by NRC-approved topical report WCAP-15049-A, “Risk-Informed Evaluation of an Extension to Accumulator Completion Times,” submitted on May 18, 1999. The NRC staff issued a notice of opportunity for comment in the Federal Register on July 15, 2002 (67 FR 46542), on possible amendments concerning TSTF-370, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on March 12, 2003 (68 FR 11880). The licensee affirmed the applicability of the following NSHC determination in its application dated September 2, 2003.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The basis for the accumulator limiting condition for operation (LCO), as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-A, the proposed change will allow plant operation with an inoperable accumulator for up to 24 hours, instead of 1 hour, before the plant would be required to begin shutting down. The impact of the increase in the accumulator CT on core damage frequency for all the cases evaluated in WCAP-15049-A is within the acceptance limit of 1.0E-06/yr for a total plant core damage frequency (CDF) less than 1.0E-03/yr. The incremental conditional core damage probabilities calculated in WCAP-15049-A for the accumulator CT increase meet the criterion of 5E-07 in Regulatory Guides (RG) 1.174, “An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,” and 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” for all cases except those that are based on design basis success criteria. As indicated in WCAP-15049-A, design basis accumulator success criteria are not considered necessary to mitigate large break loss-of-coolant accident (LOCA) events, and were only included in the WCAP-15049-A evaluation as a worst case data point. In addition, WCAP-15049-A states that the NRC has indicated that an incremental conditional core damage frequency (ICCDP) greater than 5E-07 does not necessarily mean the change is unacceptable.

The proposed technical specification change does not involve any hardware changes nor does it affect the probability of any event initiators. There will be no change to normal plant operating parameters, engineered safety feature (ESF) actuation setpoints, accident mitigation capabilities, accident analysis assumptions or inputs.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Start Printed Page 59221

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. As described in Section 9.1 of the WCAP-15049-A evaluation, the plant design will not be changed with this proposed technical specification CT increase. All safety systems still function in the same manner and there is no additional reliance on additional systems or procedures. The proposed accumulator CT increase has a very small impact on core damage frequency. The WCAP-15049-A evaluation demonstrates that the small increase in risk due to increasing the CT for an inoperable accumulator is within the acceptance criteria provided in RGs 1.174 and 1.177. No new accidents or transients can be introduced with the requested change and the likelihood of an accident or transient is not impacted.

The malfunction of safety related equipment, assumed to be operable in the accident analyses, would not be caused as a result of the proposed technical specification change. No new failure mode has been created and no new equipment performance burdens are imposed.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change does not involve a significant reduction in a margin of safety. There will be no change to the departure from nucleate boiling ratio (DNBR) correlation limit, the design DNBR limits, or the safety analysis DNBR limits.

The basis for the accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-A, the proposed change will allow plant operation with an inoperable accumulator for up to 24 hours, instead of 1 hour, before the plant would be required to begin shutting down. The impact of this on plant risk was evaluated and found to be very small. That is, increasing the time the accumulators will be unavailable to respond to a large LOCA event, assuming accumulators are needed to mitigate the design basis event, has a very small impact on plant risk. Since the frequency of a design basis large LOCA (a large LOCA with loss of offsite power) would be significantly lower than the large LOCA frequency of the WCAP-15049-A evaluation, the impact of increasing the accumulator CT from 1 hour to 24 hours on plant risk due to a design basis large LOCA would be significantly less than the plant risk increase presented in the WCAP-15049-A evaluation.

Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorneys for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201; Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: May 22, 2003.

Description of amendment request: The proposed amendment revises Technical Specification 3.3.2 governing radiation monitoring instrumentation to relax restrictions on containment purge valve operation.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The radiation monitors affected by the proposed amendment are not potential accident initiators. Adequate measures are available to compensate for instrumentation that is out of service. The proposed amendment does not affect how the affected instrumentation normally functions or its role in the response of an operator to an accident or transient. The core damage frequency in the STP [South Texas Project] PRA [probabilistic risk assessment] is not impacted by the proposed changes. Therefore, STPNOC [South Texas Project Nuclear Operating Company] concludes that there is no significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The instrumentation affected by the proposed amendment is not credited for the prevention of any accident not evaluated in the safety analysis. The proposed amendment involves no changes in the way the plant is operated or controlled. It involves no change in the design configuration of the plant. No new operating environments are created. Therefore, STPNOC concludes the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change has no significant effect on functions that are supported by the affected instrumentation. There will be no significant effect on the availability and reliability of the affected instrumentation. Adequate measures are available to compensate for instrumentation that is out of service. Therefore, STPNOC concludes the proposed 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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

Date of amendment request: August 7, 2003.

Brief description of amendments: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated August 7, 2003. Start Printed Page 59222

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: July 18, 2003.

Brief description of amendments: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, and Surveillance Requirement 3.0.4 revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated July 18, 2003.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the Start Printed Page 59223TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

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

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

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

Date of amendment request: August 29, 2003.

Brief description of amendment request: The proposed amendment would revise the Updated Final Safety Analysis Report to use the reactor building crane for heavy loads up to a total of 117 tons for removal and reinstallation activities for the reactor shield blocks prior to and during the Units 2 outage D2R18.

Date of publication of individual notice in Federal Register: September 10, 2003.

Expiration date of Individual notice: October 10, 2003.

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, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

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

Date of application for amendment: April 2, 2001, as supplemented by letters dated January 15, and August 23, 2002, March 28, and August 19, 2003.

Brief description of amendment: The amendment identifies the conditions under which the inclined fuel transfer system blind flange may be removed when primary containment integrity is required (i.e., during Modes 1, 2, and 3) and restricts this configuration to no more than 40 days per operating cycle. These changes are reflected by (1) adding Note 5 for the Actions of Technical Specification (TS) 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” (2) deleting Note 3 of TS Surveillance Requirement 3.6.1.3.3, (3) adding a conditional note to TS 3.6.1.1, “Primary Containment—Operating,” and (4) associated TS Bases changes.

Date of issuance: September 17, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment No.: 158.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 13, 2003 (68 FR 25650). The supplemental letter of August 19, 2003, 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 September 17, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 31, 2002, and supplemented by letters dated March 7 and August 28, 2003.

Brief description of amendment: The amendment revises Appendix A, Technical Specifications (TSs), of the Operating License by adding a Surveillance Requirement (SR) to TS 3.2.2, “Minimum Critical Power Ratio (MCPR),” that requires determination of the MCPR limits following completion of control rod scram time testing. The new SR provides for the required evaluation necessary to apply faster scram times to provide for improved MCPR operating limits.

Date of issuance: September 29, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment No.: 159.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58637). The supplemental letters Start Printed Page 59224contained 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 September 29, 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: January 14, 2003.

Brief description of amendment: The amendment revised technical specification sections 3.8.9, 3.15.2, 4.12.2, and associated Bases to delete the requirements for the reactor building purge air treatment system.

Date of issuance: September 23, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment No.: 245.

Facility Operating License No. DPR-50: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: (68 FR 10278) March 4, 2003. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 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 20, 2002.

Brief description of amendment: The amendment revises the Technical Specification (TS) definition of containment integrity to ensure that all power-operated valves, relief valves, and check valves are included and clarifies the handling of operability and reportability issues related to Type III containment isolation valves. The amendment also includes minor administrative and editorial changes to improve the consistency and clarity of the TSs.

Date of issuance: September 30, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days.

Amendment No.: 246.

Facility Operating License No. DPR-50: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 12, 2002 (67 FR 68729). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 30, 2003.

No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo Verde Nuclear Generating Station, Unit No. 2, Maricopa County, Arizona

Date of application for amendment: December 21, 2001, as supplemented by letters dated March 13, August 27, August 29, September 4, September 6, October 11, November 21, December 10, December 23, 2002, and March 11, June 10, July 25, and August 22, 2003.

Brief description of amendment: The amendment changes the Unit 2 Technical Specifications and operating license to support (1) replacement of the steam generators and (2) the subsequent operation at an increased maximum power level of 3990 MWt, which is a 2.94 percent increase from the current 3876 MWt.

Date of issuance: September 29, 2003.

Effective date: This license amendment is effective as of the date of issuance, and shall be implemented prior to entry into Mode 4 during the restart from the Fall 2003 refueling outage.

Amendment No.: Unit 2-149.

Facility Operating License No. NPF-51: The amendment revised the Technical Specifications and Operating License.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7412). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: August 14, 2002, as supplemented on March 11, May 16, and May 23, 2003.

Brief description of amendment: The amendment revises the Technical Specifications (TSs) related to reactivity control systems, power distribution limits, and special test exceptions.

Date of issuance: September 25, 2003.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: 280.

Facility Operating License No. DPR-65: This amendment revised the TSs.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58640). The supplements dated March 11, May 16, and May 23, 2003, provided additional information which 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 September 30, 2003.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: August 7, 2002, as supplemented on October 23, 2002.

Brief description of amendment: The amendment revises Technical Specification (TS) 6.9.1.8, “Core Operating Limits Report,” to update the list of documents that describe the analytical methods used to determine the core operating limits.

Date of issuance: September 25, 2003.

Effective date: As of the date of issuance and shall be implemented prior to Mode 4 operation of Cycle 16.

Amendment No.: 281.

Facility Operating License No. DPR-65: This amendment revised the TSs.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58639). The supplement dated October 23, 2002, provided additional information which 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 September 25, 2003.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: May 7, 2002, as supplemented on January 16, May 27, July 1, and August 21, 2003.

Brief description of amendment: The amendment revises Technical Specifications (TSs) 2.2, “Limiting Safety System Settings” and 3/4.3, “Instrumentation” to more accurately reflect the existing plant design for the Reactor Protection System, the Engineered Safety Features Actuation System, and the Radiation Monitoring System instrumentation and to provide Start Printed Page 59225consistency within the associated TS Tables.

Date of issuance: September 25, 2003.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment No.: 282.

Facility Operating License No. DPR-65: This amendment revises the TSs.

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42819). The supplements dated January 16, May 27, July 1, and August 21, 2003, provided additional information which 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 September 25, 2003.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: August 14, 2002, as supplemented December 19, 2002.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) related to Containment Systems. Specifically, the revisions: (1) Added clarification to TS 1.7, “Definitions—Containment Integrity;” (2) added clarifying information, as well as revised a portion of Surveillance Requirement 4.6.1.1 associated with the affected section of TS 3.6.1.1, “Containment Integrity;” (3) revised TS 3.6.3, “Containment Isolation Valves,” that made editorial changes, added clarifying information, and added an Action item that increased the allowed outage time from 4 hours to 72 hours for Containment Isolation Valves in closed systems; and (4) made other changes that were clarifying and/or administrative in nature. In addition, the TS Bases were revised to address these changes, as appropriate.

Date of issuance: September 29, 2003.

Effective date: As of the date of issuance, and shall be implemented within 90 days from the date of issuance.

Amendment No.: 216.

Facility Operating License No. NPF-49: This amendment revised the TSs.

Date of initial notice in Federal Register: October 1, 2002 (67 FR 61678). The December 19, 2002, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the amendment beyond the scope of the initial notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: December 20, 2001, as supplemented by letters dated March 4, 2002, September 12, 2002, November 20, 2002, and August 28, 2003.

Brief description of amendments: The amendments revised the Technical Specifications (TS) 3.3.2, Engineered Safety Features Actuation System Instrumentation.

Date of issuance: September 10, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 208 and 202.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 19, 2002 (67 FR 12601). The supplements dated March 4, 2002, September 12, 2002, November 20, 2002, and August 28, 2003, provided clarifying information that did not change the scope of the December 20, 2001, application or the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 10, 2003.

No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: January 31, 2003, as supplemented by letters dated June 12, and September 2, 2003.

Brief description of amendments: The amendments revise the Technical Specifications to incorporate revised means of determining the mass of ice in the ice condenser containment.

Date of issuance: September 29, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: 209 and 203.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 15, 2003, (68 FR 18274). The supplements dated June 12, and September 2, 2003, provided clarifying information that did not change the scope of the January 31, 2003, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: January 31, 2003, as supplemented by letters dated June 12, and September 2, 2003.

Brief description of amendments: The amendments revise the Technical Specifications to incorporate revised means of determining the mass of ice in the ice condenser containment.

Date of issuance: September 29, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: 217 and 199.

Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 15, 2003, (68 FR 18274). The supplements dated June 12, and September 2, 2003, provided clarifying information that did not change the scope of the January 31, 2003, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: March 20, 2003, supplemented by letters dated July 22, and August 5, 2003.

Brief description of amendments: The amendments revised the Technical Specifications and the licensing basis in the Updated Safety Analysis Report to support installation of a passive low-pressure injection cross connect inside containment.

Date of Issuance: September 29, 2003.

Effective date: As of the date of issuance and shall be implemented Start Printed Page 59226within 90 days from the date of issuance.

Amendment Nos.: 335, 335, and 336.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 15, 2003 (68 FR 22745). The supplement dated July 22 and August 5, 2003, provided clarifying information that did not change the scope of the March 20, 2003, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: August 16, 2002, as supplemented June 6, 2003.

Brief description of amendment: The amendment adds a new Technical Specification (TS) requirement to the Pilgrim Nuclear Power Station (Pilgrim) TSs consistent with Technical Specification Task Force (TSTF)-358. TSTF-358 addresses modifications to requirements for missed surveillances consistent with NUREG 1433, Revision 2, “Standard Technical Specification, General Electric Plants, BWR/4” (STS) surveillance requirement 3.0.3. The amendment to the Pilgrim TSs is added as TS 4.0.3.

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400), on possible amendments concerning missed surveillances, including a model safety evaluation (SE) and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 28, 2001 (66 FR 49714). The licensee affirmed the applicability of the model NSHC determination in its application dated August 16, 2002, as supplemented on June 6, 2003.

In addition, the following statement was added to the TS definition of Limiting Condition for Operation (LCO): “Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO.” The amendment also made administrative changes to add new TS Sections 3.0, “Limiting Condition for Operation (LCO) Applicability,” and 4.0, “Surveillance Requirement (SR) Applicability,” into the Pilgrim TSs. New TSs 3.0, 4.0.1, and 4.0.2 are identified as “Not Used.” These changes rectify the differences in the format and terminology of the current Pilgrim TSs compared to the STS. The associated Bases are also implemented.

Date of issuance: September 30, 2003.

Effective date: As of the date of issuance, and shall be implemented within 60 days.

Amendment No.: 203.

Facility Operating License No. DPR-35: The amendment revised the TSs.

Date of initial notice in Federal Register: July 22, 2003 (68 FR 43390). The Commission's related evaluation of the amendment is contained in a SE dated September 30, 2003.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 11, 2003.

Brief description of amendment: The amendment revises and relocates Surveillance Requirement (SR) 4.0.5 and SR 4.4.9 to the administrative section of the Technical Specifications (TS) under sections 6.5.8 and 6.5.7, respectively. The amendment also relocates TS 3.4.9, “Reactor Coolant System Structural Integrity” and its Bases to the Technical Requirements Manual. Additionally, the amendment extends the Waterford 3 flywheel volumetric examination interval to ten years.

Date of issuance: September 22, 2003.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 189.

Facility Operating License No. NPF-38: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28851). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 22, 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

Date of application for amendments: April 19, 2002, and as supplemented September 9, 2002, January 3, and July 13, 2003.

Brief description of amendments: The amendments would revise the surveillance frequency of the containment spray system nozzles from 10 years to “Following maintenance that could result in nozzle blockage, OR Following fluid flow through the nozzles.”

Date of issuance: September 22, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment Nos.: 134 and 134.

Facility Operating License Nos. NPF-37 and NPF-66: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 11, 2002 (68 FR 40023). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 22, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 28, 2002.

Brief description of amendments: The amendments authorize changes to the Updated Final Safety Analysis Report to describe the use of cast iron materials in the containment cooling service water and diesel generator cooling water systems.

Date of issuance: September 17, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment Nos.: 201 and 193.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments revise the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: December 10, 2002 (67 FR 75875). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 17, 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: December 4, 2001.

Brief description of amendment: The amendment revises the Davis-Besse Nuclear Power Station Operating License, Appendix A, Technical Specifications (TS) Section 6.9, “Administrative Controls—Reporting Requirements,” to eliminate the requirement to submit startup test reports to the NRC. Start Printed Page 59227

Date of issuance: September 25, 2003.

Effective date: As of the date of issuance and shall be implemented within 90 days.

Amendment No.: 258.

Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 22, 2003 (68 FR 43391). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 25, 2003.

No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of application for amendments: July 18, 2002.

Brief description of amendments: These amendments revise the Technical Specifications regarding the time period that inoperable channels of the engineered safety feature actuation system can be in the bypassed or tripped condition.

Date of Issuance: September 30, 2003.

Effective Date: As of the date of issuance and shall be implemented within 60 days of issuance.

Amendment Nos.: 188 and 132.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: August 20, 2002 (67 FR 53987). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2003.

No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 11, 2002, as supplemented by letters dated April 21, and July 29, 2003.

Description of amendment request: The amendment revises the Technical Specifications (TS) to eliminate the Power Range Neutron Flux High Negative Rate Reactor Trip function from TS 3/4.3.1, “Reactor Trip System Instrumentation,” TS 2.2.1, “Reactor Trip System Instrumentation Setpoints,” and their associated Bases. The amendment also revises TS 3/4.10.3, “Physics Tests,” TS 3/4.10.4, “Reactor Coolant Loops,” and TS Table 4.3-1, “Reactor Trip System Instrumentation Surveillance Requirements,” that are associated with certain testing activities required during STARTUP operations. The revision also rewords the time interval for the Analog Channel Operational Test (ACOT) in surveillance requirement (SR) 4.10.3.2. In correlation with the revision to extend the ACOT interval in SR 4.10.3.2, Table 4.3-1 Note 1 is revised. This revision also extends the ACOT interval for those Functional Units that reference TS Table 4.3-1 Note 1. The revision to TS 3/4.10.4 will delete TS 3/4.10.4 in its entirety. Additionally, as a result of deleting TS 3/4.10.4, the footnote which references TS 3/4.10.4 in TS 3/4.4.1.1 is deleted as well.

Date of issuance: October 1, 2003.

Effective date: As of its date of issuance, and shall be implemented within 90 days.

Amendment No.: 91.

Facility Operating License No. NPF-86: Amendment revises the TSs.

Date of initial notice in Federal Register: November 26, 2002 (67 FR 70767). The April 21, 2003 and July 16, 2003, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the amendment beyond the scope of the initial notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 30, 2002, as supplemented July 24 and September 25, 2003.

Brief description of amendment: The amendment authorizes changes to the Updated Safety Analysis Report (USAR) to allow the use of an upgraded computer code for design-basis accident containment integrity analyses called Generation of Thermal-Hydraulic Information for Containment (GOTHIC) version 7.0p2 (GOTHIC 7) with noted conditions.

Date of issuance: September 29, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment No.: 169.

Facility Operating License No. DPR-43: The amendment authorizes changes to the Updated Safety Analysis Report.

Date of initial notice in Federal Register: October 29, 2002 (67 FR 66011). The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 30, 2003.

Brief description of amendment: The amendment revises Kewaunee Nuclear Power Plant, Technical Specification Section 6.3, “Plant Staff Qualifications.” The amendment updates requirements that have been outdated based on licensed operator training programs being accredited by the National Academy for Nuclear Training and promulgation of the revised 10 CFR 55, “Operators’ Licenses.”

Date of issuance: October 2, 2003.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment No.: 170.

Facility Operating License No. DPR-43: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 10, 2003 (68 FR 34670). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 11, 2003, as supplemented July 16, 2003.

Brief description of amendments: The amendments revise Technical Specification (TS) 3.1.4, “Rod Group Alignment Limits,” and TS 3.1.7, “Rod Position Indication,” to add a 1-hour soak time to both TSs to allow the control rod drive mechanisms additional time following substantial rod motion to reach thermal equilibrium.

Date of issuance: October 1, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment Nos.: 160 and 151.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the TSs.

Date of initial notice in Federal Register: April 15, 2003 (68 FR 18280). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2003. Start Printed Page 59228

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 15, 2002, as supplemented by letters dated September 27, 2002, February 28, 2003, April 25, 2003, June 24, 2003, and September 12, 2003.

Brief description of amendments: The amendments authorize changes to the Final Safety Analysis Report (FSAR) Update, together with other analyses, design, and procedure changes, to implement the Diablo Canyon Power Plant NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants” program that is required to implement a dry cask Independent Spent Fuel Storage Installation (ISFSI).

Date of issuance: September 26, 2003.

Effective date: September 26, 2003, and shall be implemented following the implementation of the ISFSI. The implementation of the amendments include the incorporation into the FSAR Update the changes discussed above, as described in the licensee's application dated April 15, 2002; its supplements dated September 27, 2002, February 28, 2003, April 25, 2003, June 24, 2003, and September 12, 2003; and evaluated in the staff's safety evaluation attached to the amendment.

Amendment Nos.: 162 and 163.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments authorized revision of the FSAR Update.

Date of initial notice in Federal Register: June 11, 2002 (67 FR 40025) The supplemental letters dated September 27, 2002, February 28, 2003, April 25, 2003, June 24, 2003, and September 12, 2003, provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 26, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 10, 2003.

Brief description of amendments: The amendments revise Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), Technical Specifications (TSs) Table 3.3-1, “Reactor Trip System Instrumentation,” by modifying the “Condition and Setpoint” description of permissive interlock “P-7.” The phrase “Turbine impulse chamber pressure,” contained in the “Condition and Setpoint” description for permissive P-7, is replaced with the phrase “Turbine steam line inlet pressure” in order to support planned modifications to Salem's high pressure turbines.

Date of issuance: October 1, 2003.

Effective date: As of the date of issuance, and shall be implemented within 30 days.

Amendment Nos.: 259 and 240.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the TSs.

Date of initial notice in Federal Register: June 10, 2003 (68 FR 34672). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2003.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: September 24, 2002, supplemented by letters dated April 8 and May 21, 2003.

Brief description of amendment: This amendment revises the Action Statement and surveillance requirements for the emergency diesel generators (EDGs). The proposed changes would revise TS Section 3.8.1.1, Action b.2 and Action c.2, and TS Section 4.8.1.1, “ AC Sources” and associated Bases Section related to the EDG.

Date of issuance: September 26, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

Amendment No.: 164.

Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: November 12, 2002 (67 FR 68742). The April 8 and May 21, 2003, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 26, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 19, 2002, as supplemented by letters dated April 7, May 21, May 30, June 4, September 4, and September 12, 2003.

Brief description of amendments: The amendments revise the licensed power level for Hatch, Units 1 and 2 by 1.5 percent from 2763 megawatts thermal (MWt) to 2804 MWt. The change is based on the installation of the Advanced Measurement Analysis Group, Inc. (AMAG)/Westinghouse Crossflow ultrasonic flow measurement instrumentation, resulting in improved feedwater flow measurement accuracy. The amendment changes the Renewed Facility Operating License (RFOL) and the Technical Specifications (TSs) to reflect the increased licensed power level.

Date of issuance: September 23, 2003.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: 238 and 180.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revise the RFOL and the TSs.

Date of initial notice in Federal Register: February 18, 2003 (68 FR 7821). The supplements dated April 7, May 21, May 30, June 4, September 4, and September 12, 2003, provided clarifying information that did not change the scope of the December 19, 2002, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 23, 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: February 14, 2003, as supplemented by letters dated June 5 and August 21, 2003.

Brief description of amendment: The amendment consists of changes to Technical Specification (TS) 5.9.5, “Core Operating Limits Report (COLR).” The revised TS modifies TS 5.9.5 to add three additional methodologies in support of the Westinghouse 17×17 Start Printed Page 59229Robust Fuel Assembly (RFA)-2 fuel design with Intermediate Flow Mixers.

Date of issuance: September 30, 2003.

Effective date: As of the date of issuance and shall be implemented no later than MODE 6 entry following the next refueling outage in the fall of 2003.

Amendment No.: 46.

Facility Operating License No. NPF-90: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 1, 2003 (68 FR 15765). The supplemental letters provided clarifying information that did not expand the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 30, 2003.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 6, 2003, as supplemented by letters dated July 25, August 29, and September 16, 2003.

Brief description of amendments: The amendments revise the Final Safety Analysis Report (FSAR) and Technical Specification (TS) Bases reflecting approval of elimination of response time testing for selected Reactor Trip System and Engineered Safety Features Actuation System protection channel equipment.

Date of issuance: September 25, 2003.

Effective date: As of the date of issuance. The TS Bases shall be implemented within 60 days from the date of issuance and the FSAR shall be implemented in the next periodic update to the FSAR in accordance with 10 CFR 50.71(e).

Amendment Nos.: 107 and 107.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the FSAR and TS Bases.

Date of initial notice in Federal Register: April 15, 2003 (68 FR 18288). The July 25, August 29, and September 16, 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 as published in the Federal Register on April 15, 2003 (68 FR 18288). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 25, 2003.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 3rd day of October.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

End Signature End Preamble

[FR Doc. 03-25742 Filed 10-10-03; 8:45 am]

BILLING CODE 7590-01-P