Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: August 19, 2003 (Volume 68, Number 160)]

[Notices]

[Page 49812-49825]

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

[DOCID:fr19au03-85]

NUCLEAR REGULATORY COMMISSION

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

Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from July 25, 2003, through August 7, 2003. The last biweekly notice was published on August 5, 2003 (68 FR 46239).

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.

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However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By September 18, 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,

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supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/ or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov. Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50- 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: July 14, 2003.

Description of amendments request: The proposed amendment would revise the Technical Specifications to eliminate Surveillance Requirement (SR) 3.6.6.8. This SR is a 10-year flow test to verify that the containment spray nozzles are unobstructed.

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. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

   The proposed change eliminates the surveillance requirement to verify that the Containment Spray System spray nozzles are unobstructed every ten years. The spray nozzles are not initiators of any previously analyzed accident. Therefore, this proposed change does not increase the probability of any accident previously evaluated.

   The spray nozzles are assumed in the accident analysis to mitigate design basis accidents. Calvert Cliffs' system design Foreign Material Exclusion practices during maintenance and material accountability following maintenance, and post-maintenance testing practices ensure that the system is free of foreign material that could significantly reduce its ability to perform its intended function. These controls are considered adequate to ensure continued operability of the spray system. Since the system will be able to perform its accident mitigation function, the consequences of accidents previously evaluated are not increased.

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

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

   The proposed change eliminates the surveillance requirement to verify that the Containment Spray System spray nozzles are unobstructed every ten years. The proposed change does not introduce a new method of plant operation, does not involve a physical modification to the plant, nor does it introduce any accident initiators.

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

3. Would not involve a significant reduction in [a] margin of safety.

   The margin of safety in this case is the assurance of operability of the Containment Spray System. Calvert Cliffs' system design, Foreign Material Exclusion practices during maintenance and material accountability following maintenance, and post-maintenance testing practices ensure that the system is free of foreign material that could significantly reduce its ability to perform its intended function. These requirements, along with the remote physical location and the simple construction of the spray nozzles, provide assurance that the nozzles will remain operable.

   Therefore, this 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 amendments request involves no significant hazards consideration.

   Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, Baltimore, MD 21202.

   NRC Section Chief: Richard J. Laufer.

   Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

   Date of amendments request: May 29, 2003.

   Description of amendments request: The proposed license amendments request approval to remove the current Brunswick Steam Electric Plant (BSEP) reactor material specimen surveillance schedule from the Updated Final Safety Analysis Report and specify that BSEP, Units 1 and 2, will participate in an integrated surveillance program (ISP) developed by the Boiling Water Reactor Vessel and Internals Project.

   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 adopts an integrated surveillance program (ISP) for reactor vessel material specimen surveillances. The ISP ensures that the reactor pressure vessel will continue to meet all applicable fracture toughness requirements. No physical changes to the facilities will result from the proposed change. The initial conditions and methodologies used in accident analyses remain unchanged. The proposed change does not revise the design assumptions for systems or components used to mitigate the consequences of accidents. The accident analyses results are not affected by this proposed change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   Response: No.

   The proposed change adopts an integrated surveillance program (ISP) for reactor vessel material specimen surveillances. The ISP ensures that the reactor pressure vessel will continue to meet all applicable fracture toughness requirements. No physical changes to the facilities will result from the proposed change. The proposed change does not affect the design or operation of any system, structure, or component in the facilities. The safety functions of the related systems, structures, or components are not changed in any manner, nor is the reliability of any system, structure, or component reduced. The change does not affect the manner by which the facilities are operated and does not change any facility, structure, or component. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   Response: No.

   The proposed change has no impact on the margin of safety of any Technical Specification. There is no impact on safety limits or limiting safety system settings. The proposed change does not affect any plant safety parameters or setpoints. No physical or operational changes to the facilities will result from the proposed 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

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   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: Steven R. Carr, Associate General Counsel-- Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

   NRC Section Chief: Allen G. Howe.

   Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

   Date of amendment request: June 24, 2003.

   Description of amendment request: The proposed amendment would revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain Valves,'' to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to Operable status within 7 days.

   The NRC staff issued a notice of opportunity for comment in the Federal Register on February 24, 2003 (68 FR 8637), on possible amendments to revise the action for one or more SDV vent or drain lines with an inoperable valve, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process (CLIIP). The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 15, 2003 (68 FR 18295). The licensee affirmed the applicability of the NSHC determination below in its application dated June 24, 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.

   A change is proposed to allow the affected SDV vent and drain line to be isolated when there are one or more SDV vent or drain lines with one valve inoperable instead or requiring the valve to be restored to operable status within 7 days. With one SDV vent or drain valve inoperable in one or more lines, the isolation function would be maintained since the redundant valve in the affected line would perform its safety function of isolating the SDV. Following the completion of the required action, the isolation function is fulfilled since the associated line is isolated. The ability to vent and drain the SDVs is maintained and controlled through administrative controls. This requirement assures the reactor protection system is not adversely affected by the inoperable valves. With the safety functions of the valves being maintained, the probability or consequences of an accident previously evaluated are not significantly increased.

   Criterion 2--The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

   The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

   The proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by redundant valves and by the required action to isolate the affected line. The ability to vent and drain the SDVs is maintained through administrative controls. In addition, the reactor protection system will prevent filling of an SDV to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety.

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

   Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

   NRC Section Chief: L. Raghavan.

   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: November 14, 2002.

   Description of amendment request: Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Sec. 50.90, Duke Energy Corporation requested an amendment to the McGuire Nuclear Station Facility Operating Licenses and Technical Specifications (TS). The proposed change would revise TS 3.3.2, Engineered Safety Features Actuation System Instrumentation.

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

   The following discussion is a summary of the evaluation of the changes contained in this proposed license amendment against the three required standards of 10 CFR 50.92(c). A no significant hazards consideration is indicated if 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.

   First Standard. Implementation of this amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. Implementation of the changes contained in this amendment will have no effect on accident probabilities or consequences. The proposed changes apply to Technical Specifications 3.3.2, Engineered Safety Features Actuation System, and the equipment referenced in this Technical Specification are not accident initiating equipment. Therefore, there will be no impact on any accident probabilities caused by the NRC approval of this license amendment request. Additionally, since the design of the equipment is not being adversely modified by these proposed changes, there will be no impact on any accident consequences.

   Second Standard. Implementation of this amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms will be created as a result of the NRC approval of this license amendment request. No changes are being made to the plant which will introduce any new accident causal mechanism. This amendment does not impact any plant systems that are accident initiators; therefore, no new accident types are being created.

   Third Standard. Implementation of this amendment would not involve a significant reduction in a margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this amendment. The equipment referenced in the proposed change to Technical Specification 3.3.2 will remain capable of performing as designed. No safety margins will be impacted.

   Conclusion. Based upon the preceding discussion, Duke Energy Corporation has concluded that this proposed license amendment does not involve a significant hazards consideration.

   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.

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   Attorney for licensee: Ms. Lisa F. Vaughn, 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: July 10, 2003.

   Description of amendment request: The proposed amendments would revise the Technical Specifications to eliminate requirements that are no longer applicable due to the completion of the automatic feedwater system modifications.

   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:

   Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. This ensures 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: The proposed change to the Oconee Technical Specifications removes obsolete requirements associated with the Main Steam Line Break (MSLB) detection circuitry that are no longer necessary because of the completion of the Automatic Feedwater Isolation System (AFIS) modification on all three Oconee Units. AFIS replaced the MLSB detection system. As such, the proposed change is administrative. No actual plant equipment, operating practices, or accident analyses are affected by this change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   (2) Create the possibility of a new or different kind of accident from any kind of accident previously evaluated: The proposed change to the Oconee Technical Specifications removes obsolete requirements associated with the MSLB detection circuitry that are no longer necessary because of the completion of the AFIS modification on all three Oconee Units. AFIS replaced the MLSB detection system. As such, the proposed change is administrative. No actual plant equipment, operating practices, or accident analyses are affected by this change. No new accident causal mechanisms are created as a result of this change. The proposed change does not impact any plant systems that are accident initiators; neither does it adversely impact any accident mitigating systems. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

   (3) Involve a significant reduction in a margin of safety: The proposed change does not adversely affect any plant safety limits, set points, or design parameters. The change also does not adversely affect the fuel, fuel cladding, Reactor Coolant System, or containment integrity. The proposed change eliminates obsolete requirements and is administrative in nature. Therefore, the proposed change does not involve a 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: 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, Units 2 and 3, Grundy County, Illinois and Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

   Date of application for amendment request: October 10, 2002, as supplemented March 21 and March 28, 2003.

   Description of amendment request: The proposed amendment revises the licensing bases and Technical Specifications by utilizing an alternative source term in the design-basis radiological analyses in accordance with 10 CFR 50.67, with the exception that Technical Information Document 14844 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 at Dresden Nuclear Power Station (DNPS) and Quad Cities Nuclear Power Station (QCNPS):

   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 is 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.

   Requirements for secondary containment operability, secondary containment isolation valves, the Standby Gas Treatment (SGT) System, the Control Room Emergency Ventilation (CREV) System, and the Control Room Emergency Ventilation Air Conditioning (AC) System during movement of irradiated fuel assemblies that have decayed at least 24-hours and during core alterations are being eliminated. This is acceptable because, with the application of AST, none of these systems are credited in mitigating the consequences of a fuel handling accident after a 24-hour decay period.

   The proposed change also increases the maximum allowable primary containment leakage and the maximum allowable main steam isolation valve leakage limits. This is acceptable due to the new assumptions, used in calculating control room and offsite dose following a design basis loss-of-coolant accident, related to application of AST.

   The proposed changes do not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequence. The radiological consequences of the above design basis accidents have been evaluated with application of AST assumptions. The results conclude that the radiological consequences remain within applicable regulatory limits. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The application of AST does not affect the design, functional performance or operation of the facility. Similarly, it does not affect the design or operation of any structures, systems or components involved in the mitigation of any accidents, nor does it affect the design or operation of any component in the facility such that new equipment failure modes are created.

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

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

   Approval of the basis change from the original source term developed in accordance with Technical Information Document (TID) 14844 to a new AST, as described in Regulatory Guide 1.183, is requested. The results of the accident analyses revised in support of the proposed changes, and the requested Technical Specification changes, are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies.

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   Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed events have been carefully selected and margin has been retained to ensure that the analyses adequately bound postulated event scenarios. The dose consequences due to design basis accidents comply with the requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 1.183.

   The margin of safety is considered to be that provided by meeting the applicable regulatory limits. Relaxation of these Technical Specification requirements results in an increase in dose following certain design basis accidents. However, since the doses following these design basis accidents remain within the regulatory limits, there is not a significant reduction in a margin of safety. The changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits.

   Therefore, operation of DNPS and QCNPS 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 J. Cullen, Deputy General Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.

   NRC Section Chief: Anthony J. Mendiola.

   Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

   Date of amendment request: June 27, 2003.

   Description of amendment request: The proposed amendments revise Technical Specification 4.0.5.f and associated Bases, and Bases Section 3/4.4.8, with regard to the commitment to perform piping inspections in accordance with Generic Letter 88-01, by adding the words ``or in accordance with alternate measures approved by the NRC staff.''

   Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below.

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

   No physical changes to the facilities will result from the proposed changes. The initial conditions and methodologies used in accident analyses remain unchanged. The proposed changes do not revise or alter the design assumptions for systems or components used to mitigate the consequences of accidents. Thus, accident analyses results are not affected by these changes. Therefore, the proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The proposed changes do not affect the design or operation of any system, structure, or component in the plants. No new or different type of equipment will be installed by these proposed changes. Therefore, the proposed amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The changes do not affect any plant safety parameters or setpoints. No physical or operational changes to the facility will result from the proposed changes. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

   Based on the NRC staff's 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: Mr. Edward Cullen, Vice President & General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

   NRC Section Chief: James W. Clifford.

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

   Date of amendment request: February 11, 2003, as supplemented July 16, 2003.

   Description of amendment request: The proposed amendments would revise Technical Specification (TS) 5.5.9, ``Ventilation Filter Testing Program (VFTP),'' by (1) incorporating filter test face velocity limits for the control room special ventilation system, auxiliary building special ventilation system, spent fuel pool special and inservice purge ventilation system, and shield building ventilation system; and (2) making editorial changes. The proposed amendments would also delete the additional conditions in Appendix B of the Operating Licenses which require the licensee to complete an evaluation of the maximum test face velocity for the ventilation systems in TS 5.5.9. The additional conditions would also require the licensee to submit a license amendment request for a TS amendment to specify the maximum test face velocity if the maximum actual face velocity is greater than 110 percent of 40 feet per minute.

   In its July 16, 2003, supplemental letter, NMC withdrew the portion of its original request to revise the penetration and system bypass limit from 0.05 percent to 0.5 percent for the ventilation systems. The proposed amendments were previously noticed in the Federal Register on April 15, 2003 (68 FR 18279).

   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:

   Addition of Filter Test Face velocities

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

   This license amendment request proposes to add filter test face velocity minimum values for the control room special ventilation system, auxiliary building special ventilation system, spent fuel pool special and inservice purge ventilation system and shield building ventilation system. These ventilation systems are included in the plant design to mitigate accident consequences and are not assumed accident initiators, thus, this change does not involve a significant increase in the probability of an accident. This change will assure that the subject ventilation systems will perform within their intended design ranges thus, this change assures that the consequences of an accident are not increased.

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

   This proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment. 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 the possibility of a new

   [[Page 49818]]

   or different kind of accident from those previously analyzed has not been created.

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

   This license amendment request proposes to add filter test face velocity minimum values for the control room special ventilation system, auxiliary building special ventilation system, spent fuel pool special and inservice purge ventilation system and shield building ventilation system. These additional Technical Specification limits on system performance assures these ventilation systems are tested and maintained within their designed function limits and may increase the margin of safety for these systems. Therefore this change does not involve a significant reduction in the margin of safety.

   Editorial and administrative changes

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

   This license amendment request proposes editorial changes to Technical Specification Section 5.5.9, including replacement of ventilation system names with abbreviations and miscellaneous changes associated with addition of a new paragraph to this section, and proposes an administrative change to delete the Operating License Additional Condition for each unit that relates to NRC Generic Letter 99-02. Since these changes are editorial or administrative, they do not change any plant operating limits or technical requirements. Therefore these changes do 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 analyzed.

   This proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment. 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, the possibility of a new or different kind of accident from those previously analyzed has not been created.

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

   This license amendment request proposes editorial changes to Technical Specification Section 5.5.9, including replacement of ventilation system names with abbreviations and miscellaneous changes associated with addition of a new paragraph to this section, and proposes an administrative change to delete the Operating License Additional Condition for each unit that relates to NRC Generic Letter 99-02. Since these changes are editorial or administrative, they do not change any plant operating limits or technical requirements. Therefore these changes do not involve a significant reduction in the margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

   NRC Section Chief: L. Raghavan.

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

   Date of amendment request: July 25, 2003.

   Description of amendment request: The proposed amendment modifies Technical Specification 2.1.4, ``Reactor Coolant System (RCS) Leakage Limits.'' The proposed amendment will: (1) Add a requirement for no RCS pressure boundary leakage, (2) combine the existing RCS leakage limits into a format similar to the Improved Standard Technical Specification (ISTS), and (3) replace the existing basis associated with this specification with a basis similar in format and content of the ISTS. The proposed changes will assure that the design criteria of no RCS pressure boundary leakage is maintained and bring the Fort Calhoun Station, Unit 1 (FCS) RCS leakage specifications into alignment with the Improved Standard Technical Specifications. This amendment is modeled after the Improved Standard Technical Specifications.

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

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

   The proposed changes to Technical Specifications 2.1.4 establish a limit on reactor coolant system pressure boundary leakage and provide an allowed outage time and actions required for restoring operability. The proposed Technical Specifications address the regulatory requirements for equipment required for FCS Design Criterion 16 (similar to 10 CFR 50 GDC [General Design Criterion] 30). The change will ensure that proper Limiting Conditions for Operation are entered for equipment or functional inoperability. There are no physical alterations being made to the reactor coolant system or related systems. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The proposed changes will not result in any physical alterations to the reactor coolant system, any plant configuration, systems, equipment, or operational characteristics. There will be no changes in operating modes, or safety limits, or instrument limits. With the proposed changes in place, Technical Specifications will retain requirements for the reactor coolant system. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

   The proposed changes clarify the regulatory requirements for the reactor coolant system as defined by FCS Design Criterion 16 (similar to 10 CFR 50 GDC 30). The times established are within those invoked by the present Technical Specifications or equal to those previously reviewed and approved for use by the NRC. The proposed changes will not alter any physical or operational characteristics of the reactor coolant system and associated systems and equipment. Therefore, the proposed changes do not involve a 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: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

   NRC Section Chief: Stephen Dembek.

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

   Date of amendment request: July 25, 2003.

   Description of amendment request: The proposed amendment modifies Technical Specifications (TS) 3.0.2, Table 3-2, Table 3-5, 3.6, 3.7, 3.8, and the Definitions Section. This proposed change provides a risk- informed alternative to the existing surveillance interval for the integrated engineered safety features (ESF) and loss-of-offsite power (LOOP) testing required to be performed on each ESF equipment train each outage. The proposed change modifies the surveillance interval requirement for these refueling interval surveillance requirements to go to a staggered test-basis scheme. Using a staggered test basis, only one train would be tested each refueling outage. This amendment is modeled after the

   [[Page 49819]]

   Improved Standard Technical Specifications (ISTS) and is based on a study conducted by the Westinghouse Electric Company, on behalf of the Combustion Engineering Owners Group (CEOG) in Topical Report WCAP- 15830-P, ``Staggered Integrated ESF Testing,'' and Technical Specification Task Force (TSTF) 450.

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

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

   The proposed change affects only the Frequency at which integrated ESF testing should be performed. This testing provides assurance that the integrated ESF response will occur as assumed in the accident analyses. Testing of the components will continue to be performed as currently specified in the Technical Specifications. The only change will be for the integrated test. This test will continue to be performed on each train of ESF equipment, however, it will be performed on a Staggered Test Basis. This means that the testing will be less frequent than currently required. However, testing seldom shows failure of the equipment to perform its safety function. Because of the complexity of performing the test, the test is most likely to be repeated for some discrepancy in the set up of the test. The detailed risk review and assessment of a longer test interval shows that the change in risk is low or unchanged for equipment covered by the topical report. Licensees will provide acceptable risk reviews for plant specific equipment.

   This test does not increase the probability of an accident previously evaluated because it is not a precursor to an accident. In addition, the test is performed in a shutdown mode, where these types of accidents are not assumed to occur. The proposed change also does not increase the consequences of an accident previously evaluated because the equipment is still demonstrated to perform its safety function in an integrated manner. One complete train of equipment will be tested every refueling interval for each train. Successful completion of the test is still required.

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

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

   The proposed change affects only the Frequency at which integrated ESF testing should be performed. All more frequently performed testing is unaffected by this proposed change. No changes are being made to the equipment or to the method of equipment operation as a result of this change. No changes are being made to the tests addressed by this proposed change except the frequency.

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

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

   The proposed change affects only the surveillance interval at which integrated ESF testing should be performed. It does not impact safety system design criteria; safety system setpoint calculations or assumptions made in the safety analyses. All of the affected systems will continue to perform their safety functions as designed.

   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: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

   NRC Section Chief: Stephen Dembek.

   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: July 28, 2003.

   Description of amendment requests: The proposed amendments would revise Technical Specification 3.3.1, ``RPS Instrumentation-- Operating,'' and 3.3.5, ``ESFAS Instrumentation.'' Specifically, the proposed changes would replace the requirement for the Steam Generator Pressure--Low allowable value from its current value of 729 psia to a revised value of 717 psia.

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

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

   Response: No.

   Accidents evaluated in Chapter 15 of the Updated Final Safety Analysis Report use an analytical value of 675 psia for Steam Generator Pressure--Low and for the Main Steam Isolation Signal/ Emergency Feedwater Actuation Signal, which is the basis for the proposed change to the allowable value. The current and proposed Allowable Values are 729 psia and 717 psia respectively, which means that a 12 psi reduction in margin between the Allowable Value and the Analytical Value is being proposed. Since the Trip Setpoint may not be below 717 psia (it would be at 725 psia as required by the supporting calculation), the proposed reduction in margin between the Allowable Value and the Analytical Value does not involve a significant increase in the consequences of an accident previously evaluated.

   The proposed amendment has no effect on the probability of occurrence of accidents evaluated in Chapter 15 of the Updated Final Safety Analysis Report.

   Therefore, the proposed change does 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.

   There will be no change to the design basis of the plant. There are no new anticipated operational occurrences, or design basis accidents. No changes to any other analytical limits are being made. The current Analytical Value for Steam Generator Pressure--Low is being retained, and no changes to any of the assumptions in the accident analyses are being proposed.

   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 amendment involve a significant reduction in a margin of safety?

   Response: No.

   The change in allowable value will not adversely affect the design analysis. The plant would trip on Steam Generator Pressure-- Low at values at, or above, the analysis limit. The proposed change in the Allowable Value does not involve any change to the Analytical Value, so that the design bases limit is maintained.

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

   Based on the above, Southern California Edison concludes that the proposed amendments present 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 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.

   [[Page 49820]]

   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 21, 2003.

   Brief description of amendments: The proposed change would revise Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance Program'', to allow the use of Westinghouse Electric LLC (Westinghouse) leak limiting Alloy 800 sleeves to repair defective SG tubes as an alternative to plugging these tube.

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

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

   Response: No.

   The Westinghouse Alloy 800 leak limiting repair sleeves are designed using the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [Code] and, therefore, meet the design objectives of the original steam generator tubing. The applied stresses and fatigue usage for the repair sleeves are bounded by the limits established in the ASME Code. Mechanical testing has shown that the structural strength of repair sleeves under normal, upset, emergency, and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting (three times normal operating pressure differential) burst margin recommended by NRC's

   [U.S. Nuclear Regulatory Commission] Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR [Pressurized Water Reactor] Steam Generator Tubes.'' Burst testing of sleeve/tube assemblies has confirmed the analytical results and demonstrated that no unacceptable levels of primary-to-secondary leakage are expected during any plant condition.

   The Alloy 800 repair sleeve depth-based structural limit is determined using NRC guidance and the pressure stress equation of ASME Code, Section III, with additional margin added to account for configuration of long axial cracks. A bounding detection threshold value has been conservatively identified and statistically established to account for growth and determine the repair sleeve/ tube assembly plugging limit. A sleeved tube is plugged on detection of degradation in the sleeve/tube assembly.

   Evaluation of the repaired steam generator tube testing and analysis indicates no detrimental effects on the sleeve or sleeved tube assembly from reactor system flow, primary or secondary coolant chemistries, thermal conditions or transients, or pressure conditions as may be experienced at Comanche Peak Steam Electric Station (CPSES), Unit 1 and Unit 2. Corrosion testing and historical performance of sleeve/tube assemblies indicates no evidence of sleeve or tube corrosion considered detrimental under anticipated service conditions.

   The implementation of the proposed amendment has no significant effect on either the configuration of the plant or the manner in which it is operated. The consequences of a hypothetical failure of the sleeve/tube assembly is bounded by the current steam generator tube rupture (SGTR) analysis described in CPSES Updated Final Safety Analysis Report. Due to the slight reduction in the inside diameter caused by the sleeve wall thickness, primary coolant release rates would be slightly less than assumed for the steam generator tube rupture analysis and, therefore, would result in a lower total primary fluid mass release to the secondary system. A main steam line break or feedwater line break will not cause a SGTR since the sleeves are analyzed for a maximum accident differential pressure greater than that predicted in the CPSES safety analysis. The minimal repair sleeve/tube assembly leakage that could occur during plant operation is well within the Technical Specification leakage limits when grouped with current alternate plugging criteria calculated leakage values.

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

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

   Response: No.

   The Alloy 800 leak limiting repair sleeves are designed using the applicable ASME Code as guidance; therefore, it meets the objectives of the original steam generator tubing. As a result, the functions of the steam generators will not be significantly affected by the installation of the proposed sleeve. The proposed repair sleeves do not interact with any other plant systems. Any accident as a result of potential tube or sleeve degradation in the repaired portion of the tube is bounded by the existing SGTR accident analysis. The continued integrity of the installed sleeve/tube assembly is periodically verified by the Technical Specification requirements and the sleeved tube will be plugged on detection of degradation.

   The implementation of the proposed amendment has no significant effect on either the configuration of the plant, or the manner in which it is operated. Therefore, TXU Energy concludes that this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

   Response: No.

   The repair of degraded steam generator tubes with Alloy 800 leak limiting repair sleeves restores the structural integrity of the degraded tube under normal operating and postulated accident conditions and thereby maintains current core cooling margin as opposed to plugging the tube and taking it out of service. The design safety factors utilized for the repair sleeves are consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in the original steam generator design. The portions of the installed sleeve/tube assembly that represent the reactor coolant pressure boundary can be monitored for the initiation of sleeve/tube wall degradation and the affected tube plugged on detection of degradation. Use of the previously identified design criteria and design verification testing assures that the margin of safety is not significantly different from the original steam generator tubes.

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

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

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

   NRC Section Chief: Robert A. Gramm.

   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.

   [[Page 49821]]

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

   Date of application for amendment: August 12, 2002, as supplemented on February 28, 2003.

   Brief description of amendment: The amendment changes the surveillance requirements for the emergency diesel generators (EDGs) in Technical Specification (TS) 3/4.8.1.1, ``Electrical Power Systems-- A.C. Sources--Operating'' and TS 3/4.8.1.2, ``Electrical Power Systems--Shutdown.'' In addition, TS Section 6.0, ``Administrative Controls,'' has been revised to add a new TS to define the program requirements for testing the EDG fuel oil.

   Date of issuance: July 25, 2003.

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

   Amendment No.: 277.

   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 February 28, 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 July 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: August 14, 2002, as supplemented on April 7, 2003.

   Brief description of amendment: The amendment revises the Technical Specifications (TSs) related to Containment Systems. Specifically, the amendment: (1) Adds a new requirement for a Containment Tendon Surveillance Program to TS Section 6.0, ``Administrative Controls;'' (2) deletes TS 3/4.6.1.6, ``Containment Structural Integrity;'' (3) revises TS 3/4.6.1.1, ``Containment Integrity,'' to add a new surveillance requirement that requires that containment structural integrity be verified in accordance with the Containment Tendon Surveillance Program; (4) revises TS 3/4.6.3.1, ``Containment Isolation Valves,'' to add a new action statement that increases the allowed outage time from 4 hours to 72 hours for Containment Isolation Valves (CIVs) in closed systems; (5) makes other changes to the TSs for Containment Integrity and CIVs to provide clarity to the TSs; and (6) makes other administrative changes. In addition, the TS Bases have been revised to address the proposed changes.

   Date of issuance: July 25, 2003.

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

   Amendment No.: 278.

   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 supplement dated April 7, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 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: August 12, 2002, as supplemented on October 21, 2002, and January 15, 2003.

   Brief description of amendment: The amendment revises Technical Specification (TS) 3.8.2.3, ``Electrical Power Systems, D.C. Distribution--Operating;'' TS 3.8.2.4, ``Electrical Power Systems, D.C. Distribution--Shutdown;'' and TS 3.8.2.5, ``Electrical Power Systems, D.C. Distribution Systems (Turbine Battery)--Operating'' to use standard TS terminology in order to provide enhanced readability and usability. The amendment also provides additional criteria for determining battery operability upon restoration from a recharge or equalizing charge.

   Date of issuance: July 29, 2003.

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

   Amendment No.: 279.

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

   Date of initial notice in Federal Register: October 1, 2002 (67 FR 61677). The supplements dated October 21, 2002, and January 15, 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 July 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: November 20, 2002, as supplemented by letters dated January 21 and June 4, 2003.

   Brief description of amendments: The amendments revised the Technical Specifications Required Actions requiring suspension of operations involving positive reactivity additions and various Notes that preclude reduction in boron concentration.

   Date of issuance: July 29, 2003.

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

   Amendment Nos.: 207 & 201.

   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 18273). The supplement dated June 4, 2003, provided clarifying information that did not change the scope of the November 20, 2002, application and its supplement dated January 21, 2003, nor

   [[Page 49822]]

   the initial proposed no significant hazards consideration determination.

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 29, 2003.

   No significant hazards consideration comments received: No.

   Duke Energy Corporation, Docket No. 50-370, McGuire Nuclear Station, Unit 2, Mecklenburg County, North Carolina

   Date of application for amendments: January 31, as supplemented by letter dated May 1, 2003.

   Brief description of amendments: The amendment authorizes a revision to the Updated Final Safety Analysis Report to allow the degassing and straightening of a bent Mark-BW irradiated fuel rod in the McGuire spent fuel pool.

   Date of issuance: August 4, 2003.

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

   Amendment No.: 198.

   Facility Operating License Nos. NPF-17: Amendment authorized revision of the Updated Final Safety Analysis Report.

   Date of initial notice in Federal Register: April 15, 2003 (68 FR 18274). The supplement dated May 1, 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 August 4, 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: November 20, 2002, as supplemented by letters dated January 21 and June 4, 2003.

   Brief description of amendments: The amendments revise the Technical Specifications Required Actions requiring suspension of operations involving positive reactivity additions and various Notes that preclude reduction in boron concentration.

   Date of issuance: July 29, 2003.

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

   Amendment Nos.: 216 & 197.

   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 18273). The supplement dated June 4, 2003, provided clarifying information that did not change the scope of the November 20, 2002, application and its supplement dated January 21, 2003, nor the initial proposed no significant hazards consideration determination.

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 29, 2003.

   No significant hazards consideration comments received: No.

   Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

   Date of application for amendment: June 24, 2002, as supplemented on June 23, 2003.

   Brief description of amendment: The amendment revises Technical Specification Surveillance Requirement (SR) 3.7.7.2 to require all city water header isolation valves be open rather than only the one header supply isolation valve. On June 23, 2003, the licensee withdrew its request for changes to SR 3.7.7.1 pertaining to the city water tank volume.

   Date of issuance: August 4, 2003.

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

   Amendment No.: 218.

   Facility Operating License No. DPR-64: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 6, 2002 (67 FR 50952). The June 23, 2003, letter provided clarifying information that did not enlarge the scope of the original Federal Register notice or change the initial proposed no significant hazards consideration determination.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 4, 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 amendment: January 31, 2003.

   Brief description of amendment: The amendment revises the Technical Specification allowable values for two isolation condenser system isolation functions, namely the Steam Flow--High and Return Flow--High, for Units 2 and 3.

   Date of issuance: July 30, 2003.

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

   Amendment No.: 200.

   Facility Operating License No. DPR-19: The amendment revised the Technical Specifications.

   Amendment No.: 192.

   Facility Operating License No. DPR-25: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: April 15, 2003.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 30, 2003.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: January 27, 2003.

   Brief description of amendment: The amendment revised Technical Specifications (TS) 2.1.6, 3.2 (Table 3-5), and 5.9.1c as follows:

   (1) TS 2.1.6(1), the ``as-found'' pressurizer safety valve (PSV) lift setting tolerance band of +/-1% is increased to +1%/-3% to allow for normal setpoint variance for Modes 1 and 2. The Basis of TS 2.1.6 is revised to clarify that the PSVs are still operable and capable of performing their safety function with the wider tolerance band. The other revisions to TS 2.1.6 are administrative in nature to change defined terms to upper case text.

   (2) TS 3.2, Table 3-5, Item 3 is revised to require an ``as-left'' PSV lift setting tolerance band of +/-1%.

   (3) TS 5.9.1c is revised to remove the requirement to provide a statement in the Monthly Operating Report (MOR) concerning failures or challenges to power operated relief valves or safety valves. Generic Letter 97-02, ``Revised Contents of the Monthly Operating Report,'' does not require the MOR to provide this information.

   Date of issuance: July 25, 2003.

   Effective date: July 25, 2003, and shall be implemented within 30 days from the date of issuance.

   Amendment No.: 129.

   Facility Operating License No. DPR-40: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: March 18, 2003 (68 FR 12956).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 25, 2003.

   No significant hazards consideration comments received: No.

   [[Page 49823]]

   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 23, 2002.

   Brief description of amendment: The proposed amendment revises the Unit 2 Operating License and several sections of Technical Specifications to delete information differentiating between Unit 1 and Unit 2 specific to Model E steam generators.

   Date of issuance: July 21, 2003.

   Effective date: As of the date of issuance and shall be implemented 120 days from the date of issuance.

   Amendment Nos.: Unit 1-154; Unit 2-142.

   Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: June 25, 2002 (67 FR 42831). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 21, 2003.

   No significant hazards consideration comments received: No.

   Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant (SQN), Unit 2, Hamilton County, Tennessee

   Date of application for amendment: June 5, 2003.

   Description of amendment: The amendment revised the reactor coolant system heatup and cooldown curves (pressure-temperature (P-T) limits). The revision replaced the P-T limits that were analyzed for 14.5 Effective Full Power Years (EFPYs) with new limits analyzed for 32 EFPYs. In addition, the amendment included corresponding changes to the Technical Specification (TS) figure associated with the Low Temperature Over Pressure Protection and the TS Bases.

   Date of issuance: July 31, 2003.

   Effective date: As of the date of issuance and shall be implemented within 15 days of issuance.

   Amendment No.: 277.

   Facility Operating License No. DPR-79: Amendment revises the TSs.

   Date of initial notice in Federal Register: June 24, 2003 (68 FR 37583).

   No significant hazards consideration comments received: No.

   Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

   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 for the amendment 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.

   Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual 30-day Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

   For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

   In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

   Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

   The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 application for amendment, (2) the amendment to Facility Operating License, 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 Assess 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 Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov. The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By September 18, 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

   [[Page 49824]]

   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, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR Reference staff at 1-800-397- 4209, 301-415-4737, or by e-mail to pdr@nrc.gov. 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

   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: Rulemakings 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 the 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 petition for leave to intervene and request for hearing 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 the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

   Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas Nuclear One (ANO), Units 1 and 2, Pope County, Arkansas

   Date of amendment request: February 24, 2003, as supplemented by letters dated March 25, June 30, and July 21, 2003.

   Description of amendment request: The amendments allow the licensee to use the spent fuel crane (L-3 crane) to lift heavy loads in excess of 100 tons. Specifically the licensee received approval to use the upgraded L-3 crane for loads up to 130 tons.

   Date of issuance: July 25, 2003.

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

   Amendment Nos.: 220/248.

   Facility Operating License Nos. (DPR-51 and NPF-6): Amendments allow use of the upgraded L-3 crane to lift loads up to 130 tons.

   Public comments requested as to proposed no significant hazards consideration (NSHC): Yes (68 FR 11157, dated March 7, 2003, and 68 FR 41020, dated July 9, 2003). The notices provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notices also provided an opportunity to request a hearing by April 7, 2003, and July 23, 2003, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

   The July 21, 2003, supplemental letter provided clarifying information that did not change the scope of the Federal Register notice or the NSHC determination published July 9, 2003 (68 FR 41020).

   The Commission's related evaluation of the amendments, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated July 25, 2003.

   [[Page 49825]]

   Attorney for licensee: Winston and Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

   NRC Section Chief: Robert A. Gramm.

   Dated at Rockville, Maryland, this 11th day of August 2003.

   For The Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 03-20839 Filed 8-18-03; 8:45 am]

   BILLING CODE 7590-01-P