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

Table of Contents

I. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, August 23, 2002, through September 5, 2002. The last biweekly notice was published on September 3, 2002 (67 FR 56317).

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

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

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office ofAdministration, 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, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By October 17, 2002, 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, [1] which is available at the Commission's PDR, located at One White Flint North, 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, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, 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, 304-415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: July 31, 2002.

Description of amendment request: The proposed amendment would add a Surveillance Requirement (SR) to Technical Specification 3.2.2, “Minimum Critical Power Ratio (MCPR),” that requires determination of the MCPR limits following completion of control rod scram time testing. The proposed SR would provide for the required evaluation necessary to apply faster scram times to provide for improved MCPR operating limits.

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

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

Response: No.

The proposed change adds a new surveillance requirement (SR) to the Minimum Critical Power Ratio (MCPR) Technical Specification (TS) which requires determination of the MCPR operating limit following the completion of scram time testing of the control rods. Use of the scram speed in determining the MCPR operating limit (i.e., Option B) is an alternative to the current method for determining the operating limit (i.e., Option A). The probability of an accident previously evaluated is unrelated to the MCPR operating limit that is provided to ensure no fuel damage results during anticipated operational occurrences. This is an operational limit to ensure conditions following an assumed accident do not result in fuel failure and therefore do not contribute to the occurrence of an accident. No active or passive failure mechanisms that could lead to an accident are affected by this proposed change.

The consequences of a previously evaluated accident are not significantly increased. The proposed change ensures that the appropriate operating limit is in place. By implementing the correct operating limit the safety limit will continue to be ensured. Ensuring the safety limit is not exceeded will result in prevention of fuel failure. Therefore, since there is no increase in the potential for fuel failure there is no increase in the consequences of any accidents previously evaluated.

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 or a new or different kind of accident from any accident previously evaluated?

Response: No.

The addition of a new SR to the MCPR TS does not involve the use or installation of new equipment. Installed equipment is not operated in a new or different manner. No new or different system interactions are created, and no new processes are introduced. No new failures have been created by the addition of the proposed SR and the use of the alternate method for determining the MCPR operating limit.

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

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

Response: No.

Use of Option B for determining the MCPR operating limit will result in a reduced operating limit in comparison to the use of Option A. However, a reduction in the operating limit margin does not result in a reduction in the safety margin. The MCPR safety limit remains the same regardless of the method used for determining the operating limit. All analyzed transient results remain well within the design values for structure, systems, and components.

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: Mr. Edward J. Cullen, Vice President, General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Anthony J. Mendiola.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of amendment request: August 6, 2002.

Description of amendment request: The proposed amendment would allow the installation of up to four lead fuel assemblies (LFAs) manufactured by Framatome ANP, Inc. (FRA-ANP) into the Unit 2 Cycles 15 and 16 cores. Currently, Technical Specification 4.2.1, Fuel Assemblies, only allows fuel that is clad with either zircaloy or ZIRLO. The FRA-ANP LFA utilizes M5 [TM] alloy for the fuel cladding.

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.

Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies, states that fuel rods are clad with either zircaloy or ZIRLO. This reflects the requirements of 10 CFR 50.44, 10 CFR 50.46, and 10 CFR part 50, Appendix K, which also restricts fuel rod cladding materials to zircaloy or ZIRLO. Calvert Cliffs Nuclear Power Plant, Inc. proposes to insert up to four Framatome ANP, Inc. (FRA-ANP) fuel assemblies into Calvert Cliffs Unit 2 that have fuel rods clad in an alloy that does not meet the definition of zircaloy or ZIRLO. An exemption to the regulations has also been requested to allow these fuel assemblies to be inserted into Unit 2. The proposed change to the Calvert Cliffs Technical Specifications will allow the use of cladding materials that are not zircaloy or ZIRLO for two fuel cycles once the exemption is approved. To obtain approval of new cladding material, 10 CFR 50.12 requires that the applicant show that the proposed exemption is authorized by law, is consistent with common defense and security, will not present an undue risk to the public health and safety, and is accompanied by special circumstances. The proposed change to the Technical Specification is effective only as long as the exemption is effective. The addition of what will be an approved temporary exemption for Unit 2 to Technical Specification 4.2.1 does not change the probability or consequences of an accident previously evaluated.

Supporting analyses indicate that since the lead fuel assemblies (LFAs) will be placed in non-limiting locations, the placement scheme and the similarity of the advanced alloy to zircaloy will assure that the behavior of the fuel rods with this alloy are bounded by the fuel performance and safety analyses performed for the ZIRLO clad fuel rods in the Unit 2 Core. The similarity of ZIRLO to zircaloy was previously approved by theNuclear Regulatory Commission. Therefore, the addition of the advanced cladding M5 [TM] does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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 does not add any new equipment, modify any interfaces with existing equipment, change equipment's function, or change the method of operating the equipment. The proposed change does not affect normal plant operations or configuration. Since the proposed change does not change the design, configuration, or operation, it could not become an accident initiator.

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

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

The margin of safety for the fuel cladding is to prevent the release of fission products. Supporting analyses indicate that since the LFAs will be placed in non-limiting locations, the placement scheme and the similarity of the advanced alloy to zircaloy will assure that the behavior of the fuel rods with this alloy are bounded by the fuel performance and safety analyses performed for the ZIRLO clad fuel rods in the Unit 2 cores. Therefore, the addition of the advanced cladding M5 [TM] does not involve a significant reduction in the margin of safety.

The proposed change will add an approved temporary exemption to the Unit 2 Technical Specifications allowing the installation of up to four FRA-ANP LFAs. The assemblies use the advanced cladding material M5 [TM] that is not specifically permitted by existing regulations or Calvert Cliffs' Technical Specifications. A temporary exemption to allow the installation of these assemblies has been requested. The addition of an approved temporary exemption to Technical Specification 4.2.1 is simply intended to allow the installation of the LFAs under the provisions of the temporary exemption. The license amendment is effective only as long as the exemption is effective. This amendment does not change the margin of safety since it only adds a reference to an approved, temporary exemption to the Technical Specifications.

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

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

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: August 1, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification 3.7.1.1, “Plant Systems: Turbine Cycle Safety Valves,” to reflect results of a reanalysis of overpressurization events to reinstate the capability to operate, at corresponding reduced power levels, with up to four main steam line code safety valves in each main steam line inoperable.

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. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change will revise Specification 3.7.1.1 in accordance with revised overpressurization analyses to reinstate the capability to operate at corresponding reduced power levels with up to four main steam line code safety valves (MSSVs) in each main steam line inoperable. The MSSVs ensure the American Society of Mechanical Engineers (ASME) Code, Section III requirements are maintained to limit secondary system pressure to within 110 percent of design pressure during the most severe anticipated system operational transient. Operation with less than the full number of MSSVs is permitted as long as thermal power is restricted (and the Power Level-High trip setpoint is reset within the specified timeframe). These actions place restrictions on the allowable thermal power so that the energy transfer to the most limiting steam generator (SG) is not greater than the available relief capacity for that generator.

These changes are consistent with the Unit No. 2 Final Safety Analysis Report (FSAR) design description and analysis assumptions where the MSSVs provide the required overpressure protection. The proposed change provides assurance that the secondary side pressure remains within the bounds of the safety analyses; therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change ensures that adequate secondary side overpressure protection is available and properly maintained. This change limits plant power level based on the number of operable MSSVs. The actions require a reduction in power when the number of MSSVs is less than the full complement for each SG and also required a reduction in the Power Level-High trip setpoint.

The proposed change does not involve a physical alteration of the plant or change the plant configuration (no new or different type of equipment will be installed). The proposed change only reinstates a previously authorized mode of operation based upon revised analyses. It does not require any new or unusual operator actions. The change does not alter the way any structure, system, or component functions and does not alter the manner in which the plant is operated. The change does not introduce any new failure modes. Therefore, the proposed change will 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 MSSVs ensure the ASME Code, Section III requirements are maintained to limit the secondary system pressure to within 110 percent of the design pressure. This ensures that the overpressure protection system can cope with all operational and transient events. Plant operation with a reduced number of MSSVs is subject to the same considerations as the condition when all MSSVs are operable, i.e., a transient overpressure event must not exceed the acceptance criteria specified in the Unit No. 2 FSAR. Restricting the thermal power provides this assurance. Reducing the Power Level-High trip setpoint (within the specified timeframe provides additional assurance).

These actions place restrictions on the allowable thermal power so that the energy transfer to the most limiting SG is not greater than the available relief capacity for that generator, consequently these actions ensure the margin of safety is maintained consistent with the analysis bases.

The proposed change does not impact any acceptance criteria for the design basis accidents described in the FSAR and does not impact the consequences of accidents previously evaluated. The proposed change provides assurance that the secondary side pressure remains within the bounds of the safety analyses; therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. Therefore, the proposed change will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: Jacob I. Zimmerman, Acting.

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

Date of amendment request: August 7, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification 6.9.1.8, “Core Operating Limits Report,” to update the list of documents that describe the analytical methods used to determine the core operating limits.

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. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change to delete the document contained in section 6.9.1.8b.4 is required since it has been superceded by the most recent methodology as described in the document contained in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14). Adding the new document associated with the new section 6.9.1.8b.15 to the list of references is required for completeness. This change has no impact on plant equipment operation. Since the changes only affect description of the safety analysis methodology and do not revise any setpoints assumed in the accident analyses, they cannot affect the likelihood or consequences of accidents. Therefore, this change will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. They do not alter the way any structure, system, or component functions and do not alter the manner in which the plant is operated. These changes do not introduce any new failure modes. Therefore, the proposed changes will 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 changes have no impact on plant equipment operation. The proposed changes do not revise any setpoints assumed in the analyses and do not affect the acceptance criteria for the Steam Line Break accident. Therefore, the proposed changes will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: Jacob I. Zimmerman, Acting.

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

Date of amendment request: August 12, 2002.

Description of amendment request: The proposed amendment would change 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,” would be revised to add a new TS to define the program requirements for testing the EDG fuel oil. The TS index and the TS Bases would also be revised to address the proposed changes.

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

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes associated with revising the surveillance requirements for the Millstone Unit No. 2 emergency diesel generators and adding a new specification to define the program requirements for testing of the emergency diesel generator fuel oil will not cause an accident to occur and will not result in any change in the operation of the associated accident mitigation equipment. The ability of the equipment associated with the proposed changes to mitigate the design basis accidents will not be affected. The proposed Technical Specification surveillance requirements are sufficient to ensure the required accident mitigation equipment will be available and function properly for design basis accident mitigation. In addition, the design basis accidents will remain the same postulated events described in the Millstone Unit No. 2 Final Safety Analysis Report, and the consequences of those events will not be affected. Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

The additional proposed changes to the Technical Specifications (e.g., renumbering a requirement, modifying an index page, relocating a footnote requirement, relocating requirements to surveillance notes, relocating part of a surveillance requirement to be a separate surveillance requirement, clarifying the EDGs loads required to be energized for at least 5 minutes, clarifying the EDGs loads that should remain energized by offsite power) will not result in any technical changes to the current requirements. Therefore, these additional changes will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes to the Technical Specifications do not impact any system or component that could cause an accident. The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any unusual operator actions. The proposed changes will not alter the way any structure, system, or component functions, and will not alter the manner in which the plant is operated. There will be no adverse effect on plant operation or accident mitigation equipment. The response of the plant and the operators following an accident will not be different. In addition, the proposed changes do not introduce any new failure modes. Therefore, the proposed changes will 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 Technical Specification changes associated with revising the surveillance requirements for the Millstone Unit No. 2 emergency diesel generators and adding a new specification to define the program requirements for testing of the emergency diesel generator fuel oil will not cause an accident to occur and will not result in any change in the operation of the associated accident mitigation equipment. The equipment associatedwith the proposed Technical Specification changes will continue to be able to mitigate the design basis accidents as assumed in the safety analysis. The proposed surveillance requirements are adequate to ensure proper operation of the affected accident mitigation equipment. In addition, the proposed changes will not affect equipment design or operation, and there are no changes being made to the Technical Specification required safety limits or safety system settings. The proposed Technical Specification changes will provide adequate control measures to ensure the accident mitigation functions are maintained. Therefore, the proposed changes will not result in a reduction in a margin of safety.

The additional proposed changes to the Technical Specifications (e.g., renumbering a requirement, modifying an index page, relocating a footnote requirement, relocating requirements to surveillance notes, relocating part of a surveillance requirement to be a separate surveillance requirement, clarifying the EDGs loads required to be energized for at least 5 minutes, clarifying the EDGs loads that should remain energized by offsite power) will not result in any technical changes to the current requirements. Therefore, these additional changes will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Andersen, Acting.

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

Date of amendment request: August 14, 2002.

Description of amendment request: The proposed amendment would revise the Technical Specifications related to reactivity control systems, power distribution limits, and special test exceptions. The purpose of the proposed changes are to remove ambiguity and improve usability of the current 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. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes associated with the deletion of special test exceptions in Specifications 3/4.10.3, 3/4.10.4 and 3/4.10.5, changes to reflect the current Millstone Unit No. 2 design (i.e. full length CEAs [control element assemblies]), changes that limit the Mode applicability requirement for Shutdown Margin requirements (Specifications 3/4.1.1.1 and 3/4.1.1.2), and changes to action requirements and surveillance requirements will not cause an accident to occur and will not result in any change in operation of the mitigation equipment. The proposed changes in Specification 3/4.1.3.1 have no effect on the operability and alignment ofCEAs. The proposed allowed outage times and shutdown times are reasonable and consistent with the industry guidelines to ensure the accident mitigation equipment will be restored in a timely manner. In addition the design basis accident will remain the same postulated events described in the Millstone Unit No. 2 Final Safety Analysis Report. Since the initial conditions and assumptions included in the safety analyses are unchanged, the consequences of the postulated events remain unchanged. Therefore the proposed changes will not increase the probability or consequences of an accident previously evaluated.

The additional proposed changes to the Technical Specifications (e.g.[,] combining requirements, re-ordering requirements, relocating information to the Bases, modifying index pages, deletion or addition of footnotes) will not result in any technical changes to the current requirements. Therefore, these additional changes will not increase the probability or consequences of an accident previously evaluate[d].

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. Since the requirements remain the same, the proposed changes do not alter the way any system, structure, or component functions and do not alter the manner in which the plant is operated. The proposed changes do not introduce any new failure modes. Therefore, the proposed changes will 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 changes will not reduce the margin of safety since they have no impact on any accident analysis assumptions. The proposed changes do not decrease the scope of equipment currently required to operate or subject to surveillance testing, nor do the proposed changes affect any instrument setpoints or equipment safety functions. The effectiveness of Technical Specifications will be maintained since the changes will not alter the operation of any component or system, nor will the proposed changes affect any safety limits or safety system settings which are credited in a facility accident analysis. Therefore, there is no 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: Victor Nerses, Acting.

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

Date of amendment request: August 14, 2002.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) related to Containment Systems. Specifically, the proposed changes would: (1) Add a new requirement for a Containment Tendon Surveillance Program to TS Section 6.0, “Administrative Controls;” (2) delete TS 3/4.6.1.6, “Containment Structural Integrity;” (3) revise TS 3/4.6.1.1, “Containment Integrity,” to add a new surveillance requirement that would require that containment structural integrity be verified in accordance with the Containment Tendon Surveillance Program; (4) revise TS 3/4.6.3.1, “Containment Isolation Valves” to add a new action statement that would increase the allowed outage time (AOT) from 4 hours to 72 hours for Containment Isolation Valves (CIVs) in closed systems; (5) make other changes to the TSs for Containment Integrity and CIVs to provide clarity to the TSs; and (6) make other administrative type changes. In addition, the TS Bases would be revised to address the proposed changes.

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

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes associated with both containment integrity and CIVs that will remove ambiguity, improve usability, and increase AOT for CIVs in closed systems, will not cause an accident to occur. Operablity requirements for containment integrity and CIVs will remain the same. The ability of the equipment associated with the proposed changes to mitigate the design basis accidents will not be affected. The proposed Technical Specification requirements are sufficient to ensure the required accident mitigation equipment will be available and function properly for design basis accident mitigation. The proposed allowed outage time is reasonable and consistent with standard industry guidelines to ensure the accident mitigation equipment will be restored in a timely manner. In addition, the design basis accidents will remain the same postulated events described in the Millstone Unit No. 2 Final Safety Analysis Report, and the consequences of those events will not be affected. Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

The additional proposed changes to the Technical Specifications (e.g., changes to index, renumbering a requirement) will not result in any technical changes to the current requirements. Therefore, these additional changes will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes to the Technical Specifications do not impact any system or component that could cause an accident. The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any unusual operator actions. The proposed changes will not alter the way any structure, system, or component functions, and will not alter the manner in which the plant is operated. The response of the plant and the operators following an accident will not be different. In addition, the proposed changes do not introduce any new failure modes. Therefore, the proposed changes will 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 Technical Specification changes associated with both containment integrity and CIVs that will remove ambiguity, improve usability, and increase AOT for CIVs in closed systems, will not cause an accident to occur. Operablity requirements for containment integrity and CIVs will remain the same. Although, Containment Structural Integrity and Containment Integrity Specifications are combined, operability of the containment structure will continue to be maintained as part of a surveillance program. The equipment associated with the proposed Technical Specification changes will continue to be able to mitigate the design basis accidents as assumed in the safety analysis. The proposed allowed outage time is reasonable and consistent with standard industry guidelines to ensure the accident mitigation equipment will be restored in a timely manner. In addition, the proposed changes will not affect equipment design or operation, and there are no changes being made to the Technical Specification required safety limits or safety system settings. The proposed Technical Specification changes will provide adequate control measures to ensure the accident mitigation functions are maintained. Therefore, the proposed changes will not result in a reduction in a margin of safety.

The additional proposed changes to the Technical Specifications (e.g., changes to index, renumbering a requirement) will not result in any technical changes to the current requirements. Therefore, these additional changes will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Andersen, Acting.

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: August 22, 2002.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.8.1 to allow a one-time extension of the completion times for each Keowee Hydro Unit (KHU). This would accommodate a complete inspection and overhaul of each KHU that is expected to take more time than the current TS 3.8.1 completion time would allow.

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. Involve a significant increase in the probability or consequences of an accident previously evaluated[.]

No. The change involves an extension of the Completion Times for TS 3.8.1 Required Action C.2.2.5 and Required Action H.2. During the time that one KHU is inoperable for 72 hours or both KHUs are inoperable, a LCT [Lee combustion turbine] will be energizing both standby buses, two available offsite power sources will be maintained available, and maintenance on electrical distribution systems will not be performed unless necessary. Extending the Completion Times will decrease the likelihood of an unplanned forced shutdown of all three Oconee Units and the potential safety consequences and operational risks associated with that action. Avoiding this risk offsets the risks associated with having a design basis event during the additional completion time for having one or both KHUs inoperable.

Extending the Completion Time does not involve: (1) A physical alteration to the Oconee Units; (2) the installation of new or different equipment; (3) operating any installed equipment in a new or different manner; or (4) a change to any set points for parameters which initiative protective or mitigation action.

There is no adverse impact on containment integrity, radiological release pathways, fuel design, filtration systems, main steam relief valve set points, or radwaste systems. No new radiological release pathways are created.

The consequences of an event occurring during the extended Completion Time are the same as those that would occur during the existing Completion Time. A risk assessment shows that the additional time coupled with compensatory measures results in an acceptable level of risk.

Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.

2. Create the possibility of a new or different kind of accident from any kind of accident previously evaluated[.]

No. This change involves an extension of the Completion Times for TS 3.8.1 Required Actions C.2.2.5 and H.2 associated with restoring compliance with TS LCO 3.8.1.C. During the time period that both KHUs are inoperable, the safety function for the emergency power source will be fulfilled by the LCTs. Compensatory measures previously specified will be in place.

Extending the Completion Times does not involve a physical effect on the unit, nor is there any increased risk of a unit trip or reactivity excursion. No new failure modes or credible accident scenarios are postulated from this activity.

Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created.

3. Involve a significant reduction in a margin of safety.

No. This change involves an extension of the Completion Times for TS 3.8.1 Required Actions C.2.2.5 and H.2 associated with restoring compliance with TS LCO 3.8.1.C. During the time period that both KHUs are inoperable, the safety function for the emergency power source will be fulfilled by the LCTs. Compensatory measures previously specified will be in place to minimize electrical power system vulnerabilities.

Extending the Completion Time does not involve: (1) A physical alteration of the Oconee Units; (2) the installation of new or different equipment; (3) operating any installed equipment in a new or different manner; (4) a change to any set points for parameters which initiate protective or mitigation action; or (5) any impact on the fission product barriers or safety limits.

Therefore, this request 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: Anne W. Cottington, Winston and Strawn, 1200 17th Street, NW., Washington, DC 20005.

NRC Section Chief: John A. Nakoski.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: July 10, 2002.

Description of amendment request: The proposed change would increase the control rod scram time testing interval from 120 days to 200 days of full power operation.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of theissue 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 will not adversely impact plant operation. There will be no change in the method of performing the tests. The extended test frequency will provide some positive safety benefits by reducing the complexity of half of the control rod sequence exchange maneuvers, reducing the likelihood of a reactivity or fuel related event.

The actual rod insertion times and control rod reliability are not impacted by this proposed change; only the probability of detecting slow rods is impacted. The potential consequence of the proposed change is that one or more slow rods that would have been detected under the current 120-day frequency, may not be detected due to a reduced number of tests under the 200-day frequency.

Historical data shows that the River Bend Station control rod insertion frequency is highly reliable and rod insertion tests meet the scram time limits 99.949% of the time. Statistical analysis also demonstrates that the extended frequency would have little impact on the ability to detect slow rods in the sampling tests.

There is no safety consequence resulting from “slow” rods so long as the plant does not exceed the Technical Specification 3.1.4 Limiting Condition for Operation requirement of no more than 10 slow rods in the entire core or no two OPERABLE “slow” rods occupying adjacent positions. It is highly unlikely that a combination of missed detections and known “slow” rods would lead to the requirement to take action in accordance with Technical Specification 3.1.4. as discussed in the supporting analysis. Therefore, it is highly unlikely that the reduction in test frequency would have any impact on plant operation or safety.

The plant safety analysis assumes that all 10 slow rods take 7 seconds to reach notch position 13 which is very conservative based on actual rod performance. Control rod data shows that rods that have failed the time requirements are usually only a fraction of a second slower. The low probability of MODE 1 operation with excess slow rods combined with the historically low incidence of failure, leads to the conclusion that the probability or consequences of accidents previously evaluated are not significantly increased.

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 previously evaluated?

Response: No.

The proposed change will make no change to plant configuration or test procedures. The proposed change does not impact the operation of the plant except to reduce the number of required tests and slightly increase the probability of failing to detect a slow control rod. Operating with possibly one or two undetected slow rods does not create the possibility of an accident, since sudden control rod insertion by scram is an accident mitigation action.

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

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

Response: No.

The River Bend Station accident analyses assume a certain negative reactivity time function associated with scrams. So long as the Limiting Condition for Operation of Technical Specification 3.1.4 is met, that is, there are no more than 10 slow control rods in the entire core or two operable “slow” rods occupying adjacent locations, all accident analysis assumptions are met and there is no reduction in any margin of safety. The proposed change does not impact the Technical Specification Limiting Condition for Operation or any other allowable operating condition. The potential for an increase in the probability of being outside acceptable operating conditions due to this proposed change is insignificant. Calculations have demonstrated that the likelihood of detecting four slow rods with proposed testing frequency over a fuel cycle is lower than that with the current testing frequency by a negligible amount. The difference is even smaller for detecting a greater number of slow rods over a cycle. Therefore, since there is no impact on allowable operating parameters and the likelihood of detecting significant numbers of slow rods is only negligibly affected, there is no significant reduction in a margin of safety.

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: Mark Wetterhahn, Esq., Winston Strawn, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket No. 50-10, Dresden Nuclear Power Station (DNPS), Unit 1, Grundy County, Illinois

Date of amendment request: August 1, 2002.

Description of amendment request: The proposed changes revise the Operating License to update references to plant documents and delete Technical Specification limiting conditions for required equipment and surveillance requirements that no longer apply or are being relocated to the DNPS Technical Requirements Manual. In addition, the proposed changes delete or revise administrative control and staffing requirements that either no longer apply or have changed due to the Unit 1 Fuel Storage Pool no longer containing irradiated fuel assemblies.

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 accidents previously evaluated in the Defueled Safety Analysis Report (DSAR) affecting nuclear safety only involve the storage and handling of irradiated fuel. In each analyzed accident, irradiated fuel is assumed to be stored in the Dresden Nuclear Power Station (DNPS), Unit 1 Fuel Storage Pool. Since irradiated fuel has been permanently removed from the Unit 1 Fuel Storage Pool, the previously analyzed accidents are no longer credible, and therefore can not possibly occur. The proposed Technical Specifications (TS) changes delete requirements involving storage and handling of irradiated fuel, sealed source contamination, liquid radwaste storage radioactivity, written procedures, the Process Control Program and the unit staff, and reassign plant management responsibilities. The proposed Amended Facility Operating License (OL) changes are administrative in nature in that they only correct references to superseded plant documents. Based on the above, the proposed OL and TS 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 delete requirements involving storage and handling of irradiated fuel, sealed source contamination, liquid radwaste storage radioactivity, written procedures, the unit staff and the Process Control Program; and reassign plant management responsibilities. Deletion of requirements involving storage and handling of irradiated fuel is consistent with the current plant configuration with irradiated fuel permanently removed from the Unit 1 Fuel Storage Pool and stored in either the ISFSI [independent spent fuel storage installation] or the Unit 3 Spent Fuel Pool. Irradiated fuel in the ISFSI is controlled in accordance with 10 CFR [part] 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.” Irradiated fuel in the Unit 3 Spent Fuel Pool is controlled by the DNPS Units 2 and 3 TS in accordance with 10 CFR [part] 50, “Domestic Licensing of Production and Utilization Facilities.” Since accidentanalysis for Unit 1 irradiated fuel is now controlled by either 10 CFR [part 72 or the DNPS Units 2 and 3 TS, the deletion of DNPS Unit 1 TS requirements involving storage and handling of irradiated fuel will not create new or different kinds of accidents. Relocation of requirements for liquid radwaste storage radioactivity and sealed source contamination will not create new or different kinds of accident[s] since the requirements will still be applicable, but specified in the DNPS Technical Requirements Manual (TRM) not the DNPS Unit 1 TS. Similarly, Process Control Program requirements are redundantly contained in the DNPS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR). Therefore, deletion of requirements for the Process Control Program will not contribute to the creation of a new or different kind of accident from any accident previously evaluated. Deletion of requirements for written procedures and the unit staff, and reassignment of plant management responsibilities are administrative changes only and will not contribute to the creation of a new or different kind of accident from any accident previously evaluated. In addition, the proposed OL changes are also administrative in nature and will not contribute to the creation of a new or different kind of accident from any accident previously evaluated. The proposed changes do not physically alter the plant and will not alter the operation of the structures, systems, and components as described in the DSAR. Therefore, a new or different kind of accident from any accident previously evaluated will not be created.

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

The removal of TS requirements involving storage and handling of irradiated fuel only corrects the TS to conform to the current plant conditions (i.e., irradiated fuel permanently removed from the Fuel Storage Pool). Unit 1 irradiated fuel storage and handling is now controlled in accordance with either 10 CFR [part] 72 or the DNPS Units 2 and 3 TS (required by 10 CFR [part] 50), not the current DNPS Unit 1 TS. Thus, any changes to the DNPS Unit 1 TS involving storage and handling of irradiated fuel do not reduce any margin of safety. The relocation of the sealed source contamination and liquid storage radioactivity requirements from the DNPS Unit 1 TS to the DNPS TRM does not reduce any safety margin since the requirements still pertain. Process Control Program requirements are redundantly contained in the DNPS Units 2 and 3 UFSAR. Therefore, deletion of Process Control Program requirements from the DNPS Unit 1 TS does not reduce any safety margin since UFSAR changes are controlled under the provisions of 10 CFR 50.59, “Changes, tests, and experiments.” The deletion of written procedure requirements and unit staff requirements, and reassignment of plant management responsibilities are administrative changes only. In addition, the proposed OL changes are also administrative in nature in that they only correct references to obsolete plant documents. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC—Legal, 2301 Market Street, Philadelphia, PA 19101.

NRC Section Chief: Stephen Dembek.

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: May 31, 2002.

Description of amendment request: The proposed amendment would provide specific actions and increase restoration time for an inoperable battery charger; relocate preventative maintenance surveillance requirements for the battery charger from the Technical Specifications (TSs) to the Technical Requirements Manual (TRM); replace battery specific gravity monitoring with battery float monitoring; relocate battery float voltage and battery cell voltage, level, and temperature from the TSs to the TRM, and revise the associated surveillance requirements; create a new battery monitoring and maintenance program; provide specific actions with increased restoration time for certain battery and battery cell parameter out-of-limits conditions; eliminate the once per 60-month restriction on crediting performance discharge test for service test and restrict its use to the modified performance discharge test; revise the duration of the battery charger service test from 8 hours to 4 hours; revise the frequency of the battery performance discharge test; and delete surveillance requirements that provide excessive detail.

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

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

The class 1E direct current (DC) electrical power system including associated battery chargers are not initiators to any accident sequence analyzed in the Updated Final Safety Analysis Report (UFSAR). Operation in accordance with the proposed Technical Specification (TS) ensures that the DC system is capable of performing its function as described in the UFSAR, therefore the mitigative functions supported by the DC system will continue to provide the protection assumed by the analysis. The relocation of preventive maintenance surveillances, certain operating limits and actions to either the Technical Requirements Manual (TRM), TS Bases, or newly-created TS 6.8.4.h, “Battery Monitoring and Maintenance Program,” will not challenge the ability of the DC system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC system.

These changes do not involve any physical change to structures, systems, or components (SSCs) and do not alter the method of operation or control of SSCs. The current assumptions in the safety analysis regarding accident initiators and mitigation of accidents are unaffected by these changes. No additional failure modes or mechanisms are being introduced and the likelihood of previously analyzed failures remains unchanged.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the UFSAR will not be affected by these changes. Therefore, the consequences of previously analyzed accidents will not increase because of these changes.

Based on the above discussion, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed changes do not involve a physical alteration of the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. These changes will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. Any alteration in procedures will continue to ensure that the plant remains within analyzed limits, and no change is being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR. As such, no new failure modes are being introduced. The changes do not alter assumptions made in the safety analysis and licensing basis.

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

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

The margin of safety is established through equipment design, operating parameters, andthe setpoints at which automatic actions are initiated. The proposed changes are acceptable because the operability of the DC system is unaffected, there is no detrimental impact on any equipment design parameter, and the plant will still be required to operate within assumed conditions. Operation in accordance with the proposed TS ensures that the DC system is capable of performing its function as described in the UFSAR; therefore, the support of the DC system to the plant response to analyzed events will continue to provide the margins of safety assumed by the analysis. The relocation of preventive maintenance surveillances, certain operating limits and actions to either the TRM, TS Bases, or newly-created TS 6.8.4.h, “Battery Monitoring and Maintenance Program,” will not challenge the ability of the DC system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC system. This provides sufficient management control of the requirements that assure the batteries are maintained in a highly reliable condition.

The increased restoration times and revised criteria for monitoring the capacity of the battery and battery chargers to perform their intended function, are reasonable and generally consistent with approved standards, guidance and regulations.

Based on the above discussion, the proposed TS changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. Edward Cullen, Vice President General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Jacob I. Zimmerman, Acting.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

Date of amendment request: May 31, 2002.

Description of amendment request: The proposed license amendment would revise the Technical Specifications (TSs) to allow the Unit No.1 core to be operated with a positive moderator temperature coefficient (PMTC). TS 3/4.1.1.4, “Reactivity Control System—Moderator Temperature Coefficient (MTC),” would be changed from the current MTC limit of 0×10 [4] Δk/k/°F to +0.2×10 [4] Δk/k/°F for power levels up to 70 percent of Rated Thermal Power (RTP) and then ramping lineally from +0.2×10 [4] Δk/k/°F at 70 percent RTP to 0×10 [4] Δk/k/°F at 100% RTP. This change is being requested to address future core design requirements associated with plant operations at higher capacity factors. The amendment would include editorial and format changes as well as repagination in order to incorporate the revision into the TSs.

Basis or 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?

No. The proposed change from a moderator temperature coefficient (MTC) limit of 0 × 10 [4] Δk/k/°F to a positive moderator temperature coefficient (PMTC) of +0.2 × 10 [4] Δk/k/°F does not introduce an initiator of any design basis accident or event. The proposed change does not adversely affect accident initiators or precursors nor alter the configuration of the facility or the manner in which the plant is maintained. Thus, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed change to a PMTC does not alter or prevent the ability of structures, systems and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Accident analyses affected by the proposed change have been reanalyzed and all applicable acceptance criteria have been met. Thus, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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?

No. The change to a PMTC does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed), subsequently no new or different failure modes or limiting single failures are created. The plant will not be operated in a different manner due to the proposed change. All SSCs will continue to function as currently designed. Thus, the proposed change does not create any new or different accident scenarios.

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

No. The proposed change to a PMTC does not involve revisions to any safety limits or safety system settings that would adversely impact plant safety. The proposed amendment does not alter the functional capabilities assumed in a safety analysis for any SSCs important to the mitigation and control of design bases accident conditions within the facility.

All of the applicable acceptance criteria (i.e., preventing reactor coolant system [RCS] or main steam system overpressurization, maintaining the minimum departure from nucleate boiling ratio [DNBR], preventing core uncovery, preventing fuel temperatures from exceeding their limit, preventing clad damage, and limiting the number of fuel rods that enter a departure from nucleate boiling [DNB] condition) for each of the analyses affected by the proposed change continue to be met. The conclusions of the Updated Final Safety Analysis Report (UFSAR) remain valid. Thus, since the operating parameters and system performance will remain within design requirements and safety analysis assumptions, safety margin is maintained.

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: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Sation, Unit 2, Beaver County, Pennsylvania

Date of amendment request: July 24, 2002.

Description of amendment request: The proposed license amendment would revise the Beaver Valley Power Station, Unit No. 2, (BVPS 2) Technical Specifications (TS) Surveillance Requirement (SR) 4.7.1.5 to change the valve stroke time limit for full closure of each Main Steam Isolation Valve (MSIV) to within 6 seconds from its current 5-second limit. The amendment would also replace the quarterly partial stroke exercise requirement with criteria to test each MSIV pursuant to Specification 4.0.5. TS 4.0.5 requires testing in accordance with Section 11 of the American Society of Mechanical Engineering (ASME) Code.

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?

No. The proposed changes to the surveillance criteria for the Main Steam Isolation Valves (MSIVs) do not introduce any new initiator of a design basis accident. These proposed changes do not involve any physical modifications to the MSIVs. The proposed changes do not adversely affect accident initiators or precursors nor alter the configuration of the facility or the manner in which the plant is maintained. The proposed frequency change would reduce the potential for an (inadvertent) event initiator of full MSIV closure and resulting plant transient while retaining a sufficient test frequency to identify potential MSIV malfunctions, based on industry operating experience. Thus, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes are consistent with the safety analyses assumptions and resultant consequences. Accident analyses potentially affected by the proposed change have been reviewed and all applicable acceptance criteria continue to be met. Thus, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, the proposed changes do 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?

No. The proposed change to the surveillance criteria for MSIVs do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). Subsequently, no new or different failure modes or limiting single failures are created. The plant will not be operated in a different manner due to the proposed change. All SSCs will continue to function as currently designed. Thus, the proposed changes do not create any new or different accident scenarios.

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

No. The proposed change to the surveillance criteria for MSIVs do not involve revisions to any safety limit or safety system settings that would adversely impact plant safety. The proposed amendment does not alter the functional capabilities assumed in a safety analysis for any SSCs important to the mitigation and control of design basis accident conditions within the facility. The proposed frequency change would reduce the potential for an (inadvertent) event initiator of full MSIV closure and resulting plant transient while retaining a sufficient test frequency to identify potential MSIV malfunctions, based on industry operating experience.

All of the applicable acceptance criteria for each of the analyses affected by the proposed changes continue to be met. The conclusions of the Updated Final Safety Analysis Report (UFSAR) remain valid. Thus, since the operating parameters and system performance will remain within designed requirements and safety analysis assumptions, safety margin is maintained.

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: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Sation, Unit 2, Beaver County, Pennsylvania

Date of amendment request: May 31, 2002.

Description of amendment request: The proposed license amendment would revise the Beaver Valley Power Station (BVPS) Unit No. 2 Technical Specification (TS) Design Feature 5.3.1, Criticality, where the new fuel (fresh fuel) racks enrichment limit specified in Section 5.3.1.2.a would be increased to 5.00 weight percent (w/o) from its current 4.85 w/o limit.

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?

No. The proposed change to the new fuel storage racks enrichment limit does not introduce an initiator of any design basis accident. The text change on [the] tolerance is added for clarification of the criteria associated with [the] new fuel enrichment limit. The proposed changes do not adversely affect accident initiators or precursors nor alter the configuration of the facility or the manner in which the plant is maintained. Thus, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

The proposed changes do not alter or prevent the ability of structures, systems and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptable limits. The proposed changes are consistent with the safety analyses assumptions and resultant consequences. Accident analyses potentially affected by the proposed change have been reviewed and all applicable acceptance criteria continue to be met. Thus, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, the proposed changes do 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?

No. The proposed change to the new fuel storage racks enrichment limit and its associated text clarifications do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). Subsequently, no new or different failure modes or limiting single failures are created. The plant will not be operated in a different manner due to the proposed change. All SSCs will continue to function as currently designed. Thus, the proposed changes do not create any new or different accident scenarios.

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

No. The proposed change to the new fuel storage racks enrichment limit and its associated text clarifications do not involve revisions to any safety limit or safety system settings that would adversely impact plant safety. The proposed amendment does not alter the functional capabilities assumed in a safety analysis for any SSCs important to the mitigation and control of design basis accident conditions within the facility.

All of the applicable acceptance criteria for each of the analyses affected by the proposed changes continue to be met. The conclusions of the Updated Final Safety Analysis Report (UFSAR) remain valid. Thus, since the operating parameters and system performance will remain within design requirements and safety analysis assumptions, safety margin is maintained.

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: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: November 21, 2001, as supplemented January 25, 2002, and August 15, 2002.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS)Surveillance Requirement (SR) 4.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period would be extended from “* * * up to 24 hours to permit completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours” to “* * * up to 24 hours or up to the limit of the specified frequency, whichever is greater.” In addition, the following requirement would be added to SR 4.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.” The proposed amendments are consistent with TS Task Force traveler TSTF-358, which has been approved by the NRC. The TS Bases will be revised under the licensee's existing TS Bases control program to be consistent with TSTF-358. Lastly, a proposed administrative change moves two sentences dealing with operability requirements from SR 4.0.3 to SR 4.0.1 to make the revised TS consistent with the Standard TS for Combustion Engineering plants.

With regard to the first two changes, the NRC staff issued a notice of opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400), on possible amendments concerning missed surveillances, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 28, 2001 (66 FR 49714). The licensee affirmed the applicability of the model NSHC determination in its application dated November 21, 2001, as supplemented January 25, 2002, and August 15, 2002.

With respect to the administrative changes, the licensee provided an additional NSHC determination in its August 15, 2002, supplement.

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 for the changes associated with extending the delay period for a missed surveillance is presented below:

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

The proposed change relaxes the time allowed to perform a missed surveillance. The time between surveillances is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment being tested is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected. Any reduction in confidence that a standby system might fail to perform its safety function due to a missed surveillance is small and would not, in the absence of other unrelated failures, lead to an increase in consequences beyond those estimated by existing analyses. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. A missed surveillance will not, in and of itself, introduce new failure modes or effects and any increased chance that a standby system might fail to perform its safety function due to a missed surveillance would not, in the absence of other unrelated failures, lead to an accident beyond those previously evaluated. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The extended time allowed to perform a missed surveillance does not result in a significant reduction in the margin of safety. As supported by the historical data, the likely outcome of any surveillance is verification that the LCO [Limiting Condition for Operation] is met. Failure to perform a surveillance within the prescribed frequency does not cause equipment to become inoperable. The only effect of the additional time allowed to perform a missed surveillance on the margin of safety is the extension of the time until inoperable equipment is discovered to be inoperable by the missed surveillance. However, given the rare occurrence of inoperable equipment, and the rare occurrence of a missed surveillance, a missed surveillance on inoperable equipment would be very unlikely. This must be balanced against the real risk of manipulating the plant equipment or condition to perform the missed surveillance. In addition, parallel trains and alternate equipment are typically available to perform the safety function of the equipment not tested. Thus, there is confidence that the equipment can perform its assumed safety function. Therefore, this change does not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

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

(1) Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed amendments are administrative in nature and they do not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor do they affect Technical Specifications that preserve safety analysis assumptions. These proposed changes do not change the existing administrative controls on performance of Surveillance Requirements. The changes only relocate the existing requirements to SR 4.0.1 to closely conform to the Standard Technical Specifications. Further, the proposed changes do not alter the design, function, or operation of any plant component. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The changes being proposed are administrative in nature and do not introduce a new mode of plant operation or surveillance requirement, nor involve a physical modification to the plant. Therefore, the design, function, or operation of any plant component is not altered. The changes propose to relocate specific controls from SR 4.0.3 to SR 4.0.1 to closely conform to the Standard Technical Specifications. Therefore, operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

The proposed changes conform closely to the industry and NRC approved TSTF-358 and relates to the relocation of TS specific controls for Surveillance Requirements from SR 4.0.3 to SR 4.0.1. The specific controls are not changed only relocated to closely conform to the Standard Technical Specifications. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Section Chief: Kahtan N. Jabbour, Acting.

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

Date of amendment request: August 15, 2002.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) Section 6.8.4.h, Containment Leakage Rate Testing Program, to allow a one-time 5-year extension to the current 10-year test interval for the containment integrated leak rate test (ILRT). The proposed changes are submitted on a risk-informed basis as described in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. The risk-informed analysis supporting the proposed changes indicates that the increase in risk from extending the ILRT test interval from 10 to 15 years is insignificant.

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) Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments of the Technical Specifications add a one time extension to the current surveillance interval for Type A testing (ILRT). The current test interval of 10 years, based on performance history, would be extended on a one time basis to 15 years from the last Type A test. The proposed extension to Type A testing cannot increase the probability of an accident previously evaluated since the containment Type A test is not a modification, nor a change in the way that plant systems, structures, or components (SSC) are operated, and is not an activity that could lead to equipment failure or accident initiation. The proposed extension of the test interval does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493, Performance Based Containment Leak-Test Program, has found that generically, very few potential leak paths are not identified with Type B and C tests. NUREG-1493 concluded that an increase in the test interval to 20 years resulted in an imperceptible increase in risk. St. Lucie Units 1 and 2 provide a high degree of assurance through testing and inspection that the containment will not degrade in a manner only detectable by Type A testing. Inspections required by the ASME code and the Maintenance Rule are performed in order to identify indications of containment degradation that could affect leak-tightness. Type B and C testing required by 10 CFR part 50 part Appendix J are not affected by this proposed extension to the Type A test interval and will continue to identify containment penetrations leakage paths that would otherwise require a Type A test.

(2) Operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes do not result in operation of the facility that would create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed extension to Type A testing does not create a new or different type of accident for St. Lucie because no physical plant changes are made and no compensatory measures are being imposed that could potentially lead to a failure. There are no operational changes that could introduce a new failure mode or create a new or different kind of accident. The proposed changes only add a one time extension to the current interval for Type A testing and do not change implementation aspects of the test.

(3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

The proposed changes would not result in operation of the facility involving a significant reduction in a margin of safety. The proposed license amendments add a one time extension to the current interval for Type A testing. The current test interval of 10 years, based on historical performance, would be extended on a one time basis to 15 years from the last Type A test. The NUREG-1493 generic study of the effects of extending the Type A test interval out to 20 years concluded that there is an imperceptible increase in plant risk. Further, the extended test interval would have a minimal affect on such risk since Type B and C testing detect over 95 percent of potential leakage paths. A plant specific risk calculation, as part of the CEOG [Combustion Engineering Owners Group] joint application report, on this topic obtained results consistent with the generic conclusions of NUREG-1493. The overall increase in risk contribution was determined as 0.49 percent for Unit 1 and 0.30 percent for Unit 2.

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Section Chief: Kahtan N. Jabbour, Acting.

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

Date of amendment requests: July 23, 2002.

Description of amendment requests: The proposed amendments would revise certain 18 month surveillance requirements by eliminating the condition that testing be conducted “during shutdown,” or “during the cold shutdown or refueling mode”(i.e., shutdown conditions).

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 of occurrence or consequences of an accident previously evaluated?

Response: No.

Probability of Occurrence of an Accident Previously Evaluated

The proposed change would eliminate the requirement to perform certain 18-month surveillance tests during a shutdown condition. These surveillance tests verify that equipment will perform its intended safety function of mitigating an accident. Performing the surveillance tests during power operation does not affect any existing accident initiators or precursors. The proposed change will not create any adverse interactions with other systems that could result in initiation of a design basis accident. The format and capitalization changes are proposed to improve readability and appearance, and do not alter any requirements. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

The proposed change does not reduce the ability of the mitigating equipment to perform its safety function. The [technical specification] TS will continue to require the surveillance tests be performed on an 18 month periodicity to verify operability. One train will be verified as operable prior to testing equipment in the other train, thereby making it available to mitigate an accident. The accident analyses assume only one train is operable in the event of an accident. As aresult, the ability of the mitigating equipment to perform its safety function is unaffected by the proposed change. The format and capitalization changes are proposed to improve readability and appearance, and do not alter any requirements. Therefore, the safety related systems and components that are supported by the equipment to mitigate the consequences of an accident are not affected by the proposed change.

In summary, the probability of occurrence and the consequences of an accident previously evaluated are not significantly increased.

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

Response: No.

The proposed change does not create any new or different accident initiators or precursors. The mitigating equipment will continue to function as before the change, and will continue to be tested at the same surveillance test interval for operability. The proposed change does not create any new failure modes for the mitigating equipment and does not affect the interaction between the equipment and any other system. The format and capitalization changes are proposed to improve readability and appearance, and do not alter any requirements. Thus, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The margins of safety applicable to the proposed change are those associated with the capability of the mitigating equipment to perform its safety function. The proposed change allows the surveillance test to be performed during power operation without significantly reducing the capability of the mitigating equipment to perform in accordance with its safety margin. The format and capitalization changes are proposed to improve readability and appearance, and do not alter any requirements. Therefore, the proposed change does not involve a significant reduction in 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: David W. Jenkins, Esq.,500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: L. Raghavan.

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

Date of amendment requests: July 31, 2002.

Description of amendment requests: The proposed change will revise Diablo Canyon Power Plant (DCPP) Technical Specification (TS) 5.6.6, “Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),” to reference WCAP-14040-NP-A, “Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,” as the approved methodology for the PTLR.

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 Technical Specification (TS) changes provide the reference for the NRC approved methodology for the Diablo Canyon Power Plant (DCPP) Pressure And Temperature Limits Report (PTLR). The TS and PTLR were developed using the guidance of NRC Generic Letter (GL) 96-03, “Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,” dated January 31, 1996, which provides guidance on relocating reactor coolant system (RCS) pressure/temperature (P/T) limit curves and low-temperature overpressure (LTOP) system limits from TS to a PTLR. NRC approval of the DCPP specific application of the PTLR methodology will allow PGE [Pacific Gas and Electric Company] to use the approved PTLR methodology in the future to calculate new P/T and LTOP limits without prior NRC staff approval.

The proposed PTLR was developed using methodology previously approved by the NRC, primarily WCAP-14040-NP-A, Revision 2, “Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,” dated January 1996. PGE has evaluated this methodology and concludes it is applicable for use at DCPP. As a result, use of this methodology does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change completes relocation of the RCS P/T and LTOP limits from the TS to the PTLR. The DCPP PTLR submitted with this amendment has been developed primarily using the NRC-approved methodology of WCAP-14040-NP-A, Revision 2.

The proposed change makes no changes to plant equipment, and does not physically alter or change the function of any structures, systems or components that could initiate an accident. Through the PTLR, it provides operational controls to assure that current RCS P/T and LTOP limits are not violated. It provides for use of NRC-approved methodology for changing the RCS P/T and LTOP limits in the future without requiring prior NRC approval. As a result, 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 completes relocation of the RCS P/T and LTOP limits from the TS to the PTLR, and submits the DCPP PTLR methodology for NRC approval. The DCPP PTLR submitted with this amendment has been developed using the methodology of WCAP-14040-NP-A, Revision 2, which has previously been approved by the NRC.

The proposed change makes no changes to plant equipment, and does not physically alter or change the function of any structures, systems or components that could affect any margin of safety. Through the PTLR, it provides operational controls to assure that current RCS P/T and LTOP limits are not violated. It provides for use of NRC approved methodology for changing the RCS P/T and LTOP limits in the future without requiring prior NRC approval. As a result, the proposed change has no affect on any margin of safety.

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

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

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

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

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

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

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

Date of amendment request: July 19, 2002.

Brief description of amendment request: The proposed amendment would revise Technical Specification Surveillance Requirement (SR) 4.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period would be extended from the current limit of “* * * up to 24 hours” to “* * * up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater.” In addition, the following requirement would be added to SR 4.0.3: “A risk evaluation shall be performed for any surveillance delayed greater than 24 hours and the risk impact shall be managed.” The proposed amendment would also make administrative changes to SRs 4.0.1 and 4.0.3 to be consistent with NUREG-1432, Revision 2.

Date of publication of individual notice in the Federal Register: August 22, 2002 (67 FR 54497).

Expiration date of individual notice: September 23, 2002.

Duke Energy Corporation, et al., (the Licensee) for Operation of the Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 50-414 , and McGuire Nuclear Station, Units 1 and 2, Docket Nos 50-369 and 50-370, located in York County, South Carolina and Mecklenburg County, North Carolina

Date of amendment request: October 7, 2001, as supplemented by letter dated August 7, 2002.

Brief description of amendment request: The proposed amendments would revise Technical Specification (TS) 5.6.5 regarding the Core Operating Limits Report (COLR). TS 5.6.5.a lists the parameters for which the limiting values have been relocated by previous TS amendments from the TS to the COLR. Specifically, for both Catawba and McGuire Nuclear Stations, the amendments would revise the TS 5.6.5.a by (1) adding “60 ppm” to Item 5.6.5.a.1 regarding the moderator temperature coefficient (MTC) surveillance limit for Specification 3.1.3, and (2) by adding Item 5.6.5.a.12, “31 EFPD [effective full-power day] surveillance penalty factors for Specifications 3.2.1 and 3.2.2.” In addition, for Catawba Nuclear Station, the amendments would add Item 5.6.5.a.13, “Reactor makeup water pumps combined flow rates limit for Specifications 3.3.9 and 3.9.2.”

The limiting values for these parameters were previously relocated from the TS to the COLR without the parameter identifier being retained in the TS. Inclusion of the parameter identifier in the TS will improve consistency between the TS and the COLR. The amendments would also change Bases 3.2.1 and 3.2.3 to remove the specific date of the referenced topical report.

Date of publication of individual notice in the Federal Register: August 23, 2002 (67 FR 54680).

Expiration date of individual notice: September 23, 2002.

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: May 14, 2002.

Brief description of amendment request: The proposed changes would modify technical specification (TS) requirements for a missed surveillance through revision of Specifications 4.0.1 and 4.0.3. The delay period would be extended from the current limit of “* * * up to 24 hours to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours” to “* * * up to 24 hours or up to the limit of the specified Surveillance time interval, whichever is greater.” In addition, the following requirement would be added to Surveillance Requirement 4.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.” The proposed revision would also add a TS Bases Control Program to the Limerick Generating Station (LGS) TS.

The proposed amendment would make administrative changes to TS 6.2.2.g to revise the designation of which manager in the operations department shall hold a senior reactor operator license and to TS 6.5.1.2 to revise the LGS Plant Operations Review Committee (PORC) member composition.

Date of publication of individual notice in the Federal Register: August 27, 2002 (67 FR 55041).

Expiration date of individual notice: September 26, 2002.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: September 10, 2001.

Brief description of amendment: The amendment revised the requirements in Technical Specification Section 3.9, changing the number of operable source range monitors (SRMs) from one SRM nearest the core alteration to two SRM channels, one with its detector located in the core quadrant where core alterations are being performed, andanother with its detector located in an adjacent quadrant.

Date of Issuance: September 5, 2002.

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

Amendment No.:229.

Facility Operating License No. DPR-16: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: November 28, 2001 (66 FR 59501). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: No.

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 application for amendments: March 25, 2002.

Brief Description of amendments: The proposed amendment would revise Surveillance Requirement 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance at the Brunswick Steam Electric Plant, Units 1 and 2.

Date of issuance: August 26, 2002.

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

Amendment Nos.:224 and 249.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in the Federal Register: May 28, 2002 (67 FR 36927). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 26, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 1, 2001 as supplemented on May 13 and July 1, 2002.

Brief description of amendment: The amendment modifies the Millstone Power Station, Unit No. 3 Technical Specifications to increase the emergency diesel generator (EDG) allowed outage time, to perform a verification of the offsite circuits within 1 hour prior to, or after entering, the condition of either an inoperable offsite source or inoperable EDG, to revise the requirements for the pressurizer heaters and the pressurizer power operated relief and block valves, and to improve the format of the electrical power sources action requirements.

Date of issuance: August 26, 2002.

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

Amendment No.:210.

Facility Operating License No. NPF-49: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: January 22, 2002 (67 FR 2920). The May 13 and July 12, 2002, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination and was within the scope of the original application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 26, 2002.

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: March 26, 2002, as supplemented by letter dated June 3, 2002.

Brief description of amendments: The amendments revise the Technical Specification section 1.1, “Definitions,” to eliminate response time testing requirements for selected sensors and specified instrumentation loops for the engineered safety features system and the reactor trip system.

Date of issuance: August 23, 2002.

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

Amendment Nos.:206/187.

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

Date of initial notice in the Federal Register: April 30, 2002 (67 FR 21286). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 23, 2002.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: February 20, 2001, as supplemented by letters dated July 5, 2001, March 28, 2002, and June 14, 2002.

Brief description of amendment: The amendment consists of changes to Columbia Generating Station Physical Security Plan pertaining to the independent spent fuel storage facility installation (ISFSI).

Date of issuance: August 27, 2002.

Effective date: August 27, 2002, and shall be implemented within 30 days from the date of issuance.

Amendment No.:178.

Facility Operating License No. NPF-21: The amendment revised the operating license.

Date of initial notice in the Federal Register: April 4, 2002 (66 FR 17966). The July 5, 2001, September 13, 2001, March 28, 2002, and June 14, 2002, supplemental letters provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 27, 2002.

The Safety Evaluation contains Safeguards information and is not publicly available.

No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: June 21, 2001, as supplemented on February 8, 2002.

Brief description of amendment: The amendment revises the control rod block instrumentation requirements contained in Technical Specifications (TS) 2.1.B, Figure 2.1.1, and Tables 3.2.5 and 4.2.5. Some of the control rod block trip functions are being relocated to the Vermont Yankee Technical Requirements Manual and some of the requirements for the retained trip functions are clarified. Two trip functions are being added to the TSs.

Date of issuance: August 27, 2002.

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

Amendment No.:211.

Facility Operating License No. DPR-28: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: March 19, 2002 (67 FR 12608). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated August 27, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 12, 2002.

Brief description of amendment: The amendment revises Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any surveillance delayed greater than 24 hours and the risk impact shall be managed.”

Date of issuance: August 26, 2002.

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

Amendment No.:152.

Facility Operating License No. NPF-29: The amendment revises the Technical Specifications.

Date of initial notice in the Federal Register: July 23, 2002 (67 FR 48217). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 26, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 15, 2002, as supplemented by letter dated June 25, 2002.

Brief description of amendments: These amendments revised the reactor water cleanup system (RWCS) steam leak detection temperature isolation actuation instrumentation setpoints contained in Table 3.3.2-2 concerning items 3.b and 3.c for RWCS area temperature—high and RWCS area ventilation differential temperature—high.

Date of issuance: August 30, 2002.

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

Amendment Nos.:161/123.

Facility Operating License Nos. NPF-39 and NPF-85: The amendments revised the Technical Specifications.

Date of initial notice in the Federal Register: June 11, 2002 (67 FR 40024). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 30, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 25, 2002.

Brief description of amendments: The amendments revise the Technical Specifications, Appendix B, “Environmental Protection Plan (Non-Radiological)” to incorporate by reference the revised terms and conditions of the Incidental Take Statement included in the Biological Opinion issued by the National Marine Fisheries Service (NMFS) on May 4, 2001, as modified by NMFS letter dated October 8, 2001. They also incorporate administrative revisions necessary to change references to the National Pollutant Discharge Elimination System Permit to the Wastewater Permit, based on a change in administrative authority over these permits.

Date of issuance: August 28, 2002.

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

Amendment Nos.:183 and 126.

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

Date of initial notice in the Federal Register: February 19, 2002 (67 FR 7419). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 28, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 23, 2002, as supplemented July 15, 2002.

Brief description of amendments: Revised Technical Specifications to remove the requirement for operability of certain systems when handling fuel assemblies that have decayed a sufficient period of time such that dose consequences of the postulated fuel handling accident remain below the limits of 10 CFR part 100 and the NRC Standard Review Plan with these systems unavailable.

Date of issuance: August 30, 2002.

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

Amendment Nos.:184 and 127.

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

Date of initial notice in the Federal Register: June 25, 2002 (67 FR 42827). The July 15, 2002, supplement did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 30, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 28, 2002.

Brief description of amendment: The amendment revises Section 6.14, “Systems Integrity,” of the Technical Specifications to eliminate the Post Accident Sampling System (PASS) as a potential leakage path outside the primary containment. In addition, the amendment supersedes the previous requirements for installing and maintaining the PASS, which were imposed by NRC confirmatory orders dated March 14, 1983, and June 12, 1984.

Date of issuance: August 26, 2002.

Effective date: August 26, 2002.

Amendment No.:174.

Facility Operating License No. DPR-63: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: July 23, 2002 (67 FR 48219). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 26, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 15, 2002.

Brief description of amendment: The amendment revised Section 4.6.4, “Shock Suppressors (Snubbers),” following the guidance of Generic Letter 90-09, “Alternative Requirements for Snubber Visual Inspection and Corrective Actions,” dated December 11, 1990.

Date of issuance: August 28, 2002.

Effective date: As of the date of its issuance and shall be implemented prior to the spring 2003 refueling outage.

Amendment No.:175.

Facility Operating License No. DPR-63: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: April 16, 2002 (67 FR 18645). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 28, 2002.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: September 27, 2001, as supplemented by letter dated May 14, 2002.

Brief description of amendment: The amendment (1) revises the diesel fuel supply volume required for diesel generator (DG) operability, (2) clarifies existing wording in the Technical Specifications (TS), (3) adds a TS limiting condition for operation (LCO), and a TS Surveillance Requirement (SR) regarding the DG air receivers, (4) deletes a current TS SR concerning DG starting air compressors, and (5) restructures and renumbers the TS LCOs and SR for applicability and administrative purposes.

Date of issuance: August 27, 2002.

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

Amendment No.:129.

Facility Operating License No. DPR-22: Amendment revised the Technical Specifications.

Date of initial notice in the Federal Register: October 17, 2001 (66 FR 52801). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 27, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 29, 2002, as supplemented July 12, 2002.

Brief description of amendments: These amendments revise Technical Specification 3.8.1, “AC Sources—Operating,” to allow portions of Surveillance Requirement 3.8.1 to be performed with the units in Mode 1, 2, 3, or 4. The proposed amendments are consistent with changes made to NUREG-1431, Standard Technical Specifications, Westinghouse Plants, by Technical Specification Task Force (TSTF) Traveler, TSTF-283, Revision 3.

Date of issuance: August 29, 2002.

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

Amendment Nos.:204 and 209.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications.

Date of initial notice in the Federal Register: July 9, 2002 (67 FR 45571). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 29, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 20, 2002.

Brief description of amendments: The amendments revise Technical Specification 3.7.8 to allow the service water (SW) system to be operable with five operable SW pumps, provided one unit is in Mode 5 or 6, or defueled, and the SW system is capable of providing the required cooling water flow to required equipment.

Date of issuance: August 29, 2002.

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

Amendment Nos.:205 and 210.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications.

Date of initial notice in the Federal Register: May 14, 2002 (67 FR 34490). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 29, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 24, 2001, as supplemented April 4, 2002, May 7, 2002, June 17, 2002, July 2, 2002, July 15, 2002, and July 25, 2002.

Brief description of amendment: This amendment increases the spent fuel pool storage capacity by replacing all 11 existing rack modules with 12 new storage racks. The rerack increases the storage capacity from 1,276 storage cells to 1,712 storage cells. The degrading Boraflex neutron-absorbing material in the existing racks will be replaced by Boral material that will be used in the new racks.

Date of issuance: August 30, 2002.

Effective date: August 30, 2002.

Amendment No.:160.

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

Date of initial notice in the Federal Register: June 25, 2002 (67 FR 42810), and repeated on August 20, 2002 (67 FR 53993). The supplements listed above contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 30, 2002.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 10, 2001, as supplemented by letter dated May 23, 2002.

Brief description of amendments: The amendments revise Technical Specifications (TSs) Surveillance Requirements (SRs) 4.0.1 and 4.0.3, and incorporate a Bases Control Program in new TS 6.8.3m, in accordance with the U. S. Nuclear Regulatory Commission (NRC) staff's position on missed surveillances as described in TS Task Force—358, Revision 6. The change to SR 4.0.3 extends the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours” to “* * * up to 24 hours or up to the limit of the specified surveillance interval, whichever is greater.” In addition, the following requirement is added to SR 4.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.” In addition to revising SR 4.0.3, part of SR 4.0.3 is relocated to SR 4.0.1 and SR 4.0.1 is revised to conform to wording contained in the improved Standard TSs.

Date of issuance: August 27, 2002.

Effective date: August 27, 2002.

Amendment Nos.: Unit 1-141; Unit 2-129.

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

Date of initial notice in the Federal Register: February 5, 2002 (67 FR 5337). The May 23, 2002, supplemental letter provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 27, 2002.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Date of application for amendment: November 29, 2001, as supplemented June 18, 2002.

Brief description of amendment: These amendments establish a new operating domain for the containment partial pressure.

Date of issuance: September 5, 2002.

Effective date: As of the date of issuance and shall be implemented at the end of the Cycle 16/17 refueling outage for Unit 1, and at the end of the Cycle 15/16 refueling outage for Unit 2.

Amendment Nos.:232/214.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the Technical Specifications.

Date of initial notice in the Federal Register: April 30, 2002 (67 FR 21295). The June 18, 2002, supplement contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-339, North Anna Power Station, Unit 2, Louisa County, Virginia

Date of application for amendment: February 11, 2002, as supplemented May 16, 2002.

Brief description of amendment: This amendment revises the Facility Operating License (FOL) to allow the operation of one lead test assembly containing zirconium-based alloy for one cycle, with a lead rod burnup not to exceed 75,000 MWD/MTU.

Date of issuance: September 4, 2002.

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

Amendment No.:213.

Facility Operating License No. NPF-7: Amendment changes the FOL.

Date of initial notice in the Federal Register: April 30, 2002 (67 FR 21296). The May 16, 2002, supplement contained clarifying information only, and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No

Dated at Rockville, Maryland, this 6th day of September, 2002.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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

Footnotes

1. The most recent version of Title 10 of the Code of Federal Regulations, published January 1, 2002, inadvertently omitted the last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), regarding petitions to intervene and contentions. Those provisions are extant and still applicable to petitions to intervene. Those provisions are as follows: “In all other circumstances, such ruling body or officer shall, in ruling on—

(1) A petition for leave to intervene or a request for hearing, consider the following factors, among other things:

(i) The nature of the petitioner's right under the Act to be made a party to the proceeding.

(ii) The nature and extent of the petitioner's property, financial, or other interest in the proceeding.

(iii) The possible effect of any order that may be entered in the proceeding on the petitioner's interest .

(2) The admissibility of a contention, refuse to admit a contention if:

(i) The contention and supporting material fail to satisfy the requirements of paragraph (b)(2) of this section; or

(ii) The contention, if proven, would be of no consequence in the proceeding because it would not entitle petitioner to relief.”

References

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