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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, September 20, 2002, through October 3, 2002. The last biweekly notice was published on October 1, 2002 (67 FR 61674).

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

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

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

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

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

By November 14, 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 Start Printed Page 63688petition 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: August 28, 2002.

Description of amendments request: The proposed change will revise the expiration date of the facility operating licenses for Palo Verde Nuclear Generating Station Units 1, 2, and 3, to recapture low-power testing time.

Basis for proposed no significant hazards consideration determination: Start Printed Page 63689As 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.

Response: No.

The proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated because they do not involve a change to design configuration or operation of the facilities. In addition, each PVNGS [Palo Verde Nuclear Generating Station] unit was designed and constructed to ensure a 40-year service life. Design features were incorporated that provide for inspectability of structures, systems and components during the 40-year service life. Surveillance, inspectability and maintenance practices which have been implemented in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and the unit Technical Specifications provide assurance that any degradation in plant safety-related equipment will be identified and corrected to provide continued safe operation of each unit throughout the duration of the applicable facility operating license.

The largest recapture period requested by the proposed amendment requests is 8 months (Unit 3). This recapture period represents less than 1.7% of the 40-year service life of the respective unit, and is insignificant from an aging effects perspective. Therefore, the proposed amendments 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.

Response: No.

The proposed amendments would revise the expiration of each facility operating license such that the expiration of each facility operating license is based upon issuance of the respective FPOL [full power operating license] and not upon issuance of the respective LPOL [low power operating license]. No physical changes are being made to the design features or operation of the facilities. Therefore, the proposed amendments do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed amendments would revise the expiration of each facility operating license such that the expiration of each facility operating license is based upon issuance of the respective FPOL and not upon issuance of the respective LPOL. No physical changes are being made to the design features or operation of the facilities.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary and the containment structure) to limit the radiological dose to the public and control room operators in the event of an accident. The proposed amendments to the facility operating licenses are administrative in nature and have no impact on the margin of safety and robustness provided in the design and construction of the facilities. In addition, the proposed amendments will not relax any of the criteria used to establish safety limits, nor will the proposed amendments relax safety system settings or limiting conditions of operation as defined in the Technical Specifications. Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.

Based on the above information, APS [Arizona Public Service Company] concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999.

NRC Section Chief: Stephen Dembek.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: June 11, 2002.

Description of amendments request: The proposed amendment revises Technical Specification (TS) 3.7.11, Spent Fuel Pool Exhaust Ventilation System, for Units 1 and 2 to redefine the applicability of the TS to limit the types of fuel assemblies to which it applies. This proposed amendment revises TS 3.7.11 to not require the ventilation be operable or in operation for the movement of fuel assemblies with an appropriate amount of decay time. An evaluation has determined that 32 days is adequate time to allow for sufficient radioactive decay of short lived isotopes resulting in no increase in offsite dose if the ventilation system were not operable. This change is consistent with changes previously approved for the Improved Standard Technical Specifications as described in Technical Specification Task Force—51.

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

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

The system affected by this proposed amendment is the spent fuel pool exhaust ventilation system (SFPEVS). This system mitigates the consequences of a Fuel Handling Incident (FHI) by filtering radioactive iodine from the air above the spent fuel pool prior to that air being exhausted to the environment. This limits the offsite dose possible from a[n] FHI. This proposed amendment revises the Technical Specification applicability for the SFPEVS by defining when the ventilation system is required to limit offsite dose due to a[n] FHI. Because this system is used for the mitigation of an accident, it is not an accident initiator. Therefore, the probability of an accident previously evaluated is not increased.

The only design basis accident originating in the spent fuel pool is the FHI. This accident is evaluated in the Updated Final Safety Analysis Report. The analysis assumed credit for the filtration system. However, a more recent evaluation shows that 32 days after a fuel assembly has been removed from the critical reactor core, adequate radioactive decay has occurred which compensates for the filtration of the ventilation system. Thus, no increase in offsite dose occurs under these conditions. Therefore, the consequences of an accident previously evaluated have not increased.

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

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

The SFPEVS is not being altered by this amendment request. No changes are made in the way in which the SFPEVS is operated or in the way fuel is moved in the spent fuel pool. The only change made would allow some irradiated fuel assemblies to be moved in the spent fuel pool without requiring the operation of the ventilation system. Since no changes are being made to the operation of the SFPEVS when it is needed for offsite dose control and the SFPEVS is a[n] accident mitigating system only, changes in when this system is needed to operate cannot create a new type of accident.

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

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

The margin of safety provided by the SFPEVS is to limit offsite dose due to a[n] FHI to the limits described in the Updated Final Safety Analysis Report. The evaluation performed indicates that radioactive decay can compensate for the filtration system. Thirty-two days after fuel occupied a critical reactor core, enough radioactive decay has occurred that the offsite dose from a[n] FHI assuming no filtration is the same as the dose determined in the Updated Final Safety Analysis Report. Therefore, no reduction in the margin of safety has occurred because the Start Printed Page 63690offsite dose is the same as the previously approved dose limits.

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 amendments 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.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: July 17, 2002.

Description of amendment request: The proposed amendment would allow the installation of up to four lead fuel assemblies (LFAs) Manufactured by Westinghouse Electric Company (Westinghouse) into the Unit 2 Cycles 15 and 16 cores. Currently, Technical Specification (TS) 4.2.1, Fuel Assemblies, only allows fuel that is clad with either zircaloy or ZIRLO. The Westinghouse LFAs utilizes advance zirconium-based material for cladding. In addition, the statements currently in TS 4.2.1 concerning the lead test assemblies that were allowed to be inserted for Unit 1 Cycles 13, 14, and 15 will be deleted.

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, 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 Westinghouse fuel assemblies into Calvert Cliffs Unit 2 that have some fuel rods clad in zirconium alloys that do 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 materials, 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. In addition, the statements concerning the exemption for Unit 1 Cycles 13, 14, and 15 have been deleted, since Unit 1 Cycle 15 is completed, and therefore the exemption has expired. The addition of what will be an approved temporary exemption for Unit 2 and the deletion of an expired exception 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 ZIRLO 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. Therefore, the addition of these advanced claddings 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 ZIRLO will assure that the behavior of the fuel rods with these alloys 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 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 Westinghouse LFAs. The assemblies use the advanced cladding materials that are 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.

In addition, the words concerning the exemption for Unit 1 Cycles 13, 14, and 15 will be deleted since Unit 1 Cycle 15 is completed, and therefore, the exemption has expired. This change does not change the margin of safety since it only deletes a reference to an expired 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.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: August 6, 2002.

Description of amendments request: The proposed amendment would revise Technical Specification (TS) 3.9, Refueling Operations, to incorporate two changes previously approved in NUREG-1432, Revision 2, “Combustion Engineering Improved Standard Technical Specifications” dated April 2001. One change would add a note to Limiting Condition for Operation 3.9.3 allowing penetration flow path(s) that have direct access from containment atmosphere to the outside atmosphere to be unisolated under administrative control. The other change would replace the requirement in TSs 3.9.4 and 3.9.5 to “[c]lose all containment penetrations providing direct access from the containment atmosphere to outside atmosphere” with a set of more detailed and less restrictive requirements.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR part 50.91(a), the licensee has provided its analysis of the issue of no significant hazards Start Printed Page 63691consideration, which is presented below:

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

Closing the containment penetrations is considered to be a mitigator of the radiological consequences of a fuel handling incident and a loss of SDC [Shutdown Cooling], not an initiator. Therefore, allowing containment penetration flow paths to be unisolated and the containment purge valves to be opened during these outage activities does not involve a significant increase in the probability of an accident previously evaluated.

The consequence of a fuel handling incident is the release of radioactivity from Containment. The impact of the proposed change to the calculated offsite dose resulting from a fuel handling incident has been evaluated and determined to be acceptable. The fuel handling incident analysis assumes no containment closure. The amount of radioactivity that could be released as a result of the proposed change is bounded by the current analysis of record. Therefore, having containment penetration flow paths unisolated during core alterations and fuel handling does not involve an increase in the consequences of an accident previously evaluated.

The consequences of a loss of SDC is the potential for release of radioactivity to the atmosphere outside Containment. Closing containment penetrations is a mitigator of that consequence. Administrative controls will be put in place to ensure that in an emergency containment closure can be quickly achieved. The containment purge system isolation valves are closed automatically on a containment high radiation signal and can be shut by remote manual operation. Therefore, the proposed changes do not involve a significant increase in the consequences of a loss of SDC.

Therefore, the proposed Technical Specification changes do 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.

This requested change does not involve a significant change in the operation of the plant and no new accident initiation mechanism is created by the proposed changes. Closing containment penetrations is considered to be a mitigator of the radiological consequences of any accident in the Containment, not an initiator. The containment penetration flow paths are currently opened and closed during the course of an outage. The proposed changes allow them to remain open during a period when they are currently required to be closed.

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

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

The margin of safety for containment closure during core alteration/fuel handling is based on the amount of offsite dose resulting from a fuel handling incident. An offsite dose calculation previously approved by the Nuclear Regulatory Commission for a fuel handling incident assumes no containment closure, and any activity released from the Containment is unfiltered. The analysis will apply to the containment penetration flow paths that could be opened under administrative controls and therefore, does not involve a significant reduction in the margin of safety.

The margin of safety for containment closure in the case of a loss of SDC is twofold: (1) The time required to close the Containment to prevent a radioactive release to the atmosphere outside Containment if SDC is lost; and (2) the ability to retain the pressure generated by boiling of reactor coolant as a result of a loss of SDC.

Currently the Technical Specifications are vague and overly restrictive concerning the requirement for containment closure when SDC is lost. The proposed change eliminates unclear requirements and provides a clear way to establish containment closure that meets the Bases description for the Action, which is to prevent fission products from being released from the Containment during a loss of SDC incident. The containment purge isolation valves close rapidly on a high radiation signal or are closed by remote manual operation. The proposed changes do not increase the possibility of a release of radiation following a loss of SDC incident.

Therefore, the ability to provide containment closure is maintained and the margin of safety is not significantly reduce[d] by this proposed activity.

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

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

NRC Section Chief: Richard J. Laufer.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of amendment request: August 28, 2002.

Description of amendment request: The amendment, proposed by Carolina Power & Light Company to the Harris Nuclear Plant (HNP) Technical Specifications (TS), revises TS 6.9.1.6.2 to add analytical methodology references, which are used to determine 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. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved by the NRC for use in licensing applications. Plant structures, systems, and components will not be operated in a different manner as a result of these proposed changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not increase 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 incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved by the NRC for use in licensing applications. Plant structures, systems, and components will not be operated in a different manner as a result of these changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not create the possibility of a new or different type of accident from any previously evaluated.

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

The proposed changes incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved by the NRC for use in licensing applications. Plant structures, systems, and components will not be operated in a different manner as a result of these changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: William D. Johnson, Vice President and Corporate Secretary, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe Start Printed Page 63692

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

Date of amendment request: August 7, 2002.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) related to safety system settings. Specifically, the proposed changes would revise: (1) TS 1.0 “Definitions;” (2) TS 2.2.1 “Limiting Safety System Settings—Reactor Trip System Instrumentation Setpoints;” (3) TS 3.3.1 “Reactor Trip System Instrumentation;” (4) TS 3.3.2 “Engineered Safety Features Actuation System Instrumentation;” (5) TS 3.7.7 “Control Room Emergency Ventilation System;” (6) TS 3.8.3.1 “Onsite Power Distribution—Operating.” In addition, the appropriate TS Bases would be revised to conform with 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 changes associated with the operability requirements, surveillance requirements and allowed outage times will improve usability of the facility Technical Specifications. The proposed changes will clearly reflect the existing plant design for the Reactor Trip System (RTS), Engineered Safety Features Actuation System (ESFAS), Control Room, Emergency Ventilation System, and Electrical Power Systems Instrumentation. The proposed changes will also provide consistency within the individual technical specifications tables (e.g. Table 2.2-1, Table 3.3-1, and Table 4.3-1). In addition, there are no hardware changes associated with the proposed changes. Therefore, these systems will continue to perform within the bounds of the previously performed accident analyses.

The proposed changes to the operability requirements will not affect the instrumentation's ability to mitigate the design basis accidents. The proposed allowed outage times (i.e. the required action times) are reasonable and consistent with industry guidelines to ensure the affected instrumentation will be restored in a timely manner and provide consistency with the existing plant design. The design basis accidents will remain the same postulated events described in the Millstone Unit No. 3 Final Safety Analysis Report (FSAR), and the consequences of these events will not be affected. Therefore, the proposed 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 do not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. The proposed changes do not alter the way any structure, system, 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 do 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 assumption. The proposed changes do not decrease the scope of equipment currently required to be operable 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: James W. Andersen, Acting.

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

Date of amendment request: September 12, 2002.

Description of amendment request: The proposed amendments would temporarily revise Technical Specification (TS) 3.5.2, “Emergency Core Cooling System (ECCS);” TS 3.6.6, “Containment Spray System;” TS 3.7.5, “Auxiliary Feedwater (AFW) System;” TS 3.7.7, “Component Cooling Water (CCW) System;” TS 3.7.8, “Nuclear Service Water System (NSWS);” and TS 3.8.1, “AC Sources—Operating” for Catawba Nuclear Station, Units 1 and 2. The proposed TS changes will allow the “A” NSWS header for each unit to be taken out of service for 7 days for pipe replacement. This pipe replacement is scheduled to occur when Units 1 and 2 are at 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 the issue of no significant hazards consideration, which is presented below:

Catawba is currently pursuing a project to replace a portion of the ‘A’ train of the nuclear service water system (NSWS) piping for both units. This is necessary to maintain the long-term reliability of the NSWS. This project represents a challenge in that it is not possible to isolate, drain, replace, restore and test the NSWS during the current TS action time frame. The purpose of this submittal is to request a temporary change to the existing TS for the systems affected during the project. This will permit an orderly and efficient project implementation during power operation on both units. The specific change is to extend the TS required action time from 72 hours to 168 hours.

The following discussion is a summary of the evaluation of the changes contained in this proposed amendment against the 10 CFR 50.92(c) requirements to demonstrate that all three standards are satisfied. A no significant hazards consideration is indicated if operation of the facility in accordance with the proposed amendment would not:

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

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

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

First Standard

The pipe replacement project for the NSWS and proposed TS changes have been evaluated to assess their impact on normal operation of the systems affected and to ensure that the design basis safety functions are preserved. During the pipe replacement the other NSWS train will be operable and no major maintenance or testing will be done on the operable train. The operable train will be protected to help ensure it would be available if called upon.

This pipe replacement project will enhance the long term structural integrity in the NSWS system. This will ensure that the ‘A’ NSWS header maintains its flow margin to ensure its ability to comply with design basis requirements and increase the overall reliability for many years.

The increased NSWS train unavailability as a result of the implementation of this amendment does involve a one time increase in the probability or consequences of an accident previously evaluated during the time frame the NSWS header is out of service for pipe replacement. Considering this small time frame for the ‘A’ NSWS train outage with the increased reliability and the decrease in unavailability of the NSWS system in the future because of this project, the overall probability or consequences of an accident previously evaluated will decrease.

An evaluation was performed utilizing PRA [probabilistic risk analysis] for Start Printed Page 63693extending the NSWS TS time limit from 72 hours to 168 hours. The [CDF] core damage frequency contribution from the proposed outage extension is judged to be acceptable for a one-time, or rare, evolution. Considering the change in CDF associated with the outage extension in the framework of an average over a five-year period, the average annual contribution is considered a low-to-moderate increase in the CDF for consideration of permanent changes to the licensing basis.

Therefore, because this is a temporary and not a permanent change, the time averaged risk increase is acceptable. The increase in the overall reliability of the NSWS along with the decreased unavailability in the future because of the pipe replacement project will result in an overall increase in the safety of both Catawba units. Therefore, the consequences of an accident previously evaluated remains unaffected and there will be minimal impact on any accident consequences.

Second Standard

Implementation of this amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed temporary TS changes do not affect the basic operation of the ECCS [emergency core cooling system], containment spray system, NSWS, AFW [auxiliary feedwater], CCW [component cooling water], or EDG [emergency diesel generator] systems. The only change is increasing the required action time frame from 72 hours to 168 hours (ECCS, containment spray system, NSWS, AFW, CCW, and EDG). During the project, contingency measures will be in place to provide additional assurance that the affected systems will be able to complete their design functions. No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant, which will introduce any new accident causal mechanisms.

Third Standard

Implementation of this amendment would not involve a significant reduction in a margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed temporary TS amendment. During the ‘A’ NSWS train outage, the affected systems will still be capable of performing their required functions and contingency measures will be in place to provide additional assurance that the affected systems will be maintained in a condition to be able to complete their design functions. No safety margins will be impacted.

The probabilistic risk analysis conducted for this proposed amendment demonstrated that the CD[F] associated with the outage extension is judged to be acceptable for a one-time or rare evolution. Therefore, there is not a significant reduction in the margin of safety.

Based upon the preceding discussion, Duke Energy has concluded that the proposed amendment for a temporary one time TS change does not involve a significant hazards consideration.

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

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

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: September 3, 2002.

Description of amendment request: The proposed changes to the Columbia Generating Station Technical Specifications (TSs) are to: (1) Add depleted uranium to the fuel assembly composition description in TS 4.2.1, (2) revise TS 5.6.5.b to incorporate references to the analytical methods used to determine core operating limits and remove those that are no longer used, and (3) format the revised references as described in Industry/Technical Specification Task Force (TSTF) Traveler, TSTF-363, “Revised Topical Report References in ITS [Improved Technical Specifications] 5.6.5, COLR [Core Operating Limits Report].”

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.

Assembly and core designs employing depleted uranium are employed in other reactors and are within the FRA-ANP fuel design methods and experience base. There will be no change to the composition of the fuel pellets (i.e., UO2) containing the depleted uranium except for a slight decrease in the amount of U235. Therefore the use of depleted uranium in the fuel rods does not affect the mechanical performance of the rods. Flux profile measurements performed on these core designs correlate with calculated values in a manner consistent with fuel assembly designs that do not include depleted uranium.

Core operating limits are established to support Technical Specification 3.2, Power Distribution, requirements which ensure that fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). The methods used to determine the core operating limits for each operating cycle are based on methods previously found acceptable by the NRC and listed in TS section 5.6.5.b. A change to TS section 5.6.5.b is requested to include the FRA-ANP methods in the list of approved methods applicable to Columbia Generating Station. Application of these approved methods will continue to ensure that acceptable operating limits are established to protect the fuel cladding integrity during normal operation and AOOs.

The requested Technical Specification changes do not involve any plant modifications or operational changes that could affect system reliability, performance, or possibility of operator error. The requested changes do not affect any postulated accident precursors, do not affect any accident mitigation systems, and do not introduce any new accident initiation mechanisms.

Therefore, these changes do not increase the probability or consequences of any 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.

Assembly and core designs employing depleted uranium are within the capability of the NRC-approved FRA-ANP fuel design methods. There will be no change to the composition of the fuel pellets (i.e., UO2) containing the depleted uranium except for a slight decrease in the amount of U235. Therefore the use of depleted uranium in the fuel rods does not affect the mechanical performance of the rods.

Changes to the methodologies listed in the TS are administrative. The proposed changes do not involve any new modes of operation, any changes to setpoints, or any plant modifications. The core operating limits will continue to be developed using NRC-approved methods that account for the mixed fuel core design. The proposed methods do not result in any new precursors to an accident.

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

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

Assembly and core designs employing depleted uranium are within the capability of the NRC-approved FRA-ANP fuel design methods. There will be no change to the composition of the fuel pellets (i.e., UO2) containing the depleted uranium except for a slight decrease in the amount of U235. Therefore the use of depleted uranium in the fuel rods does not affect the mechanical performance of the rods.

The core operating limits will continue to be determined using methodologies that have been approved by the NRC.

On this basis, the implementation of the changes does not involve a significant reduction in margin of safety.

Start Printed Page 63694

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

Attorney for licensee: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.

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

Date of amendment request: September 19, 2002.

Description of amendment request: The proposed amendments would add a new analytical method to Technical Specifications (TS) Section 5.6.5, “Core Operating Limits Report (COLR).” The proposed change supports the core design efforts currently in process for the upcoming Unit 2 refueling outage scheduled to begin in January 2003.

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 to LaSalle County Station, Unit 1 and Unit 2 Technical Specifications (TS), involves reference to a new fuel analytical method in TS Section 5.6.5, “Core Operating Limits Report (COLR).” This code package supports the methodology currently being used by Framatome-ANP in the reload design and analysis process.

The proposed change to TS Section 5.6.5 will add to the list of methods used to determine the core operating limits, the fuel analytical method that supports design of the LaSalle County Station Unit 2 Cycle 10 reload that is currently scheduled to startup on February 5, 2003. The addition of the approved method to TS Section 5.6.5 has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated. The NRC approved method has been reviewed to ensure that the output accurately models predicted core behavior, has no affect on the type or amount of radiation released, and has no affect on predicted offsite doses in the event of an accident.

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

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

Response: No.

The proposed change to TS Section 5.6.5 does not affect the performance of any LaSalle County Station structure, system, or component credited with mitigating any accident previously evaluated. The use of a new analytical method, which has been reviewed and approved by the NRC for the design of a core reload, will not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. The proposed change does not introduce any new modes of system operation or failure mechanisms.

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

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

Response: No.

The proposed change to TS Section 5.6.5 adds the current analytical method for design and analysis of core reloads to the list of methods used to determine the core operating limits. The NRC has approved for use by licensees the analytical method being added. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.

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

Based on the above, Exelon Generation Company concludes that the proposed amendment presents a no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of “no significant hazards consideration” is justified.

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

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

NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, Pennsylvania

Date of amendment request: August 7, 2002.

Description of amendment request: The proposed amendments would: (1) Revise the surveillance frequency for air or smoke flow testing of containment spray nozzles, as specified in surveillance requirements (SRs) 4.6.2.1.d and 4.6.2.2.f, from once per 10 years to following maintenance which results in the potential for nozzle blockage and allows the use of a visual examination in lieu of an air or smoke flow test; (2) eliminate the SR 4.6.2.2.e.3 criteria for the river water flow rate through the Recirculation Spray System heat exchangers; and (3) make minor clarifying changes to the text in Technical Specification 3.3.1.1.

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 containment spray system nozzle surveillance frequency, the manner in which the nozzles are verified to be unobstructed, and the elimination of the associated Recirculation Spray System (RSS) flow rate requirement does not introduce an initiator of any design basis accident or event. 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 river/service water system monitoring program ensures that the river/service water flow through the RSS heat exchangers will be maintained. The proposed changes to provide alternate wording for the P-13 function in the Reactor Protection System solely for clarification of the current criteria does not adversely affect accident initiators or precursors. 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. Introduction of foreign materials into the containment spray system from the exterior is unlikely due to the location of the spray headers, the passive nature of the nozzles, station foreign material controls, and the fact that the containment spray headers are maintained dry above the water level maintained in the Recirculation Water Storage Tank which inhibits active degradation mechanisms such as corrosion. The proposed amendment to eliminate the associated RSS flow rate requirements and the text clarification for the P-13 function do not introduce an initiator of any design basis accident or event. The proposed changes are consistent with the safety analysis assumptions and resultant consequences. Accident analyses potentially affected by the proposed change have been reviewed and none are adversely affected. Thus, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Start Printed Page 63695

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

No. The proposed changes to the containment spray nozzle surveillance frequency, the manner in which the nozzles are verified to be unobstructed, the elimination of the associated RSS flow rate requirement, and the text clarifications for the P-13 function do not involve any 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 changes to the containment spray system nozzle surveillance frequency, the manner in which the nozzles are verified to be unobstructed, the elimination of the associated RSS flow rate requirement, and the text clarifications for the P-13 function do not involve revisions to any safety limits or safety system settings that would adversely impact plant safety. No current setpoints are altered by this change. 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. The river/service water system monitoring program ensures that the river/service water flow through the RSS heat exchangers will be maintained.

All of the applicable acceptance criteria for each of the analyses affected by the proposed change continue to be met. The conclusions of the [Updated Final Safety Analysis Report] remain valid. Thus, since the operating parameters and system performance will remain within design requirements and safety analysis, 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.

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

Date of amendment requests: August 23, 2002.

Description of amendment requests: The proposed amendments would revise Facility Operating Licenses (OLs) DPR-58 and DPR-74, for Unit 1 and Unit 2, respectively, and Technical Specifications (TS) for Unit 1 and Unit 2. The licensee proposes to delete obsolete and/or expired license conditions from the Unit 1 and Unit 2 OLs, and make editorial changes to the Unit 1 and Unit 2 OLs. Administrative changes to specific TS for Unit 1 and Unit 2 are also proposed.

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.

The proposed deletion of obsolete and/or expired license conditions from the Unit 1 and Unit 2 OLs is administrative in nature. The deletion of these license conditions has no impact on plant operations since these requirements are no longer applicable. The proposed TS changes, the renumbering of the Unit 2 OL pages, and the correction of a typographical error in the Unit 1 OL are also administrative in nature and do not impact CNP's current design and licensing basis. Since the proposed changes are administrative and do not impact plant operations or design, the changes do not involve any significant increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated.

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

Response: No.

The proposed deletion of obsolete and/or expired license conditions from the Unit 1 and Unit 2 OLs is administrative in nature. The proposed TS changes, the renumbering of the Unit 2 OL pages, and the correction of a typographical error in the Unit 1 OL are also administrative in nature. These proposed changes do not impact plant operations or plant equipment in any manner or involve a physical alteration to the plant, nor a change in the methods used to respond to plant transients that has not been previously analyzed. No new or different equipment is being installed and no installed equipment is being removed or operated in a different manner. Consequently, no new failure modes are introduced and the proposed administrative changes to the Unit 1 and Unit 2 OL do not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated. Therefore, 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 proposed deletion of obsolete and/or expired license conditions from the Unit 1 and Unit 2 OLs does not affect alarm or trip setpoints. The proposed TS changes, the renumbering of the Unit 2 OL pages, and the correction of a typographical error in the Unit 1 OL are administrative in nature and do not impact the condition, design, or performance of any plant structure, system or component. Thus, the results of the accident analyses will not be affected as any input assumptions are protected. The format changes improve readability and appearance and do not alter any requirements. Thus, the proposed changes do not involve a significant reduction in a margin of safety.

In summary, based upon the above evaluation, [Indiana Michigan Power Company] I&M has concluded that the proposed changes involve no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

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

NRC Section Chief: L. Raghavan.

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

Date of amendment requests: August 30, 2002.

Description of amendment requests: The proposed amendments would revise the reactor trip system (RTS) and engineered safety features actuation system (ESFAS) Technical Specification (TS) Surveillance Requirements in TS 3/4.3.1 and TS 3/4.3.2, respectively, by increasing (1) the channel operational test surveillance intervals for analog channels, logic cabinets, and reactor trip breakers (RTBs), and (2) the completion time (CT) and bypass time (BT) for the RTBs in accordance with the evaluation and justifications presented in the referenced document, WCAP-15376, Revision 0, “Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times,” dated October 2000. Additionally, the proposed amendments would remove Mode 2 applicability for the RTS low pressurizer pressure and high pressurizer water level trips and to add a note to TS Table 4.3-1 clarifying that channel functional testing requirements for the reactor trip bypass breakers are only applicable when they are racked in Start Printed Page 63696and closed for bypassing an RTB. The proposed amendments would also make format and capitalization changes to the affected TS pages that improve the appearance of the TS pages, but do not affect any requirements.

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.

The proposed changes to the STIs [surveillance test intervals] and RTB CT and BT reduce the potential for inadvertent reactor trips and spurious actuations, and therefore do not increase the probability of any accident previously evaluated. The proposed changes do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the RTS and ESFAS signals. These changes satisfy the acceptance criteria specified in the NRC's regulatory guidance for evaluating risk-informed changes in RG 1.174 [“An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” dated July 1998] and RG 1.177 [“An Approach for Plant-Specific Risk-Informed Decisionmaking: Technical Specifications,” dated August 1998]. The RTS and ESFAS will continue to perform their functions with high reliability as originally assumed in the safety analysis, and the increase in risk is within the acceptance criteria of existing regulatory guidance; therefore, there will not be a significant increase in the consequences of any accidents.

The RTS and ESFAS are not accident initiators or precursors in the safety analysis. No new initiators are created by this activity. The proposed changes do not change any RTS or ESFAS setpoints, nor do they alter the accident mitigation function of any system, structure or component, design assumptions, conditions or configuration of the facility, or the manner in which the plant is operated and maintained. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

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.

No new system interfaces or interactions are created. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed changes do not result in a change in the manner in which the RTS and ESFAS provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. The proposed changes to STI, CT, and BT do not change any existing accident scenarios, do not alter assumptions made in the safety analysis, nor create any new or different accident scenarios.

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

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174 and RG 1.177.

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: March 1, 2002.

Description of amendment request: The proposed amendment would revise the containment spray nozzle inspection frequency contained in Technical Specification Surveillance Requirement (SR) 3.6.6.9. Specifically, the inspection frequency would be conducted “[f]ollowing maintenance which could result in nozzle blockage,” rather than at the currently specified 10-year frequency. Maintenance which could result in nozzle blockage is controlled by procedures which establish foreign material exclusion (FME) controls. The FME controls require post-maintenance verification of system cleanliness and freedom from foreign materials.

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

The following evaluation supports the finding that operation of the facility in accordance with the proposed change would not:

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

The proposed change revises the surveillance frequency for containment spray nozzle inspections from every ten years to following maintenance which could result in nozzle blockage. Analyzed events are initiated by the failure of plant structures, systems or components. The containment spray system is not considered as an initiator of any analyzed event. The proposed change does not have a detrimental impact on the integrity of any plant structure, system or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. As a result, the probability of any accident previously evaluated, is not significantly increased.

This change does not affect the plant design. Due to the plant design, the spray headers are maintained dry at the level of the nozzles. Formation of corrosion products is unlikely due to the corrosion resistant materials used in spray header construction. Due to their location at the top of the containment, introduction of foreign material from sources external to the spray nozzles is unlikely. Since loss of foreign material control when working within the affected boundary is the most likely cause for obstruction, testing or inspection following such an occurrence would verify nozzle condition, and the system would be capable of performing its safety function. As a result, the consequences of any accident previously evaluated are not significantly affected.

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

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

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

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

The margin of safety for this system is based on the capacity of the spray headers. Since the system is not susceptible to corrosion induced obstruction or obstruction Start Printed Page 63697from sources external to the spray nozzles, and performance of maintenance on the system would require evaluation of the potential for nozzle blockage and the possible need for a test or inspection, the likelihood that the spray nozzles might be blocked would not be affected by the reduction in surveillance frequency. Therefore, the capacity of the system would remain unaffected. Hence, this change does not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: July 31, 2002.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS). The proposed amendments represent a full implementation of an alternative source term (AST) for the Units 1, 2, and 3 operating licenses. The amendments adopt the AST methodology by revising the current accident source term and replacing it with an accident source term as prescribed in 10 CFR 50.67.

The AST analyses were performed using the guidance provided by Regulatory Guide 1.183, “Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” dated July 2000, and Standard Review Plan Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.” The four limiting design basis accidents (DBAs) considered were the Control Rod Drop Accident, the Refueling Accident, the Loss of Coolant Accident, and the Main Steam Line Break Accident.

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:

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

The AST and those plant systems affected by implementing AST do not initiate DBAs. The AST does not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences. The implementation of the AST has been evaluated in the analyses for the limiting DBAs at BFN. The equipment affected by the proposed change is mitigative in nature and relied upon following an accident. The proposed changes to the TS do revise certain performance requirements. However, these changes will not involve a revision to the parameters or conditions that could contribute to the initiation of a design basis accident discussed in Chapter 14 of the BFN Updated Final Safety Analysis Report.

Plant specific radiological analyses have been performed and, based on the results of these analyses, it has been demonstrated that the dose consequences of the limiting events considered in the analyses are within the regulatory guidance provided by the NRC for use with the AST. This guidance is presented in 10 CFR 50.67, Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1. Therefore, the proposed amendment does not result in a significant increase in the consequences or a significant increase in the probability of any previously evaluated accident.

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

Implementation of AST does not alter any design basis accident initiators. These changes do not affect the design function or mode of operations of systems, structures, or components in the facility prior to a postulated accident. Since systems, structures, and components are operated essentially no differently after the AST implementation, no new failure modes are created by this proposed change. Therefore, the proposed license amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The changes proposed are associated with a revision to the licensing basis for BFN. The results of accident analyses revised in support of the proposed change are subject to the acceptance criteria in 10 CFR 50.67. The analyzed events have been carefully selected, and the analyses supporting this submittal have been performed using approved methodologies. The dose consequences of these limiting events are within the acceptance criteria provided by the regulatory guidance as presented in 10 CFR 50.67, Regulatory Guide 1.183, and SRP 15.0.1.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the changes are considered to not result in 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 are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: August 1, 2002.

Description of amendment request: The proposed amendment would revise the Browns Ferry design and licensing basis as described in section 14.5.2.8 of the Updated Final Safety Analysis Report (UFSAR) to eliminate consideration of a pressure regulator downscale failure event as an abnormal operational transient.

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:

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

The proposed amendment involves a change in transient analysis assumptions and does not change the plant or the manner in which it is operated. Therefore, the amendment has no affect on the probability of an accident. The proposed amendment is based upon upgrades and reliability improvements made to the main turbine generator electro-hydraulic control system, which render the analysis of a Pressure Regulator Downscale Failure event and consideration of the associated consequences unnecessary. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed amendment involves a change in transient analysis assumptions and does not change the plant or the manner in which it is operated. The only event affected, the Pressure Regulator Failure Downscale transient, is of a type already considered. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed amendment eliminates the consideration of the Pressure Regulator Downscale Failure event as an abnormal Start Printed Page 63698operational transient based on the low likelihood of occurrence of such an event due to improvements in the system design of the main turbine electro-hydraulic control system. Other abnormal operational pressurization transients as described in the UFSAR will continue to be analyzed and ensure required margins of safety to fuel thermal limits are maintained. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. In conclusion, the proposed amendment does not adversely affect the public health and safety, and does not involve any significant safety hazards.

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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 7902.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: September 3, 2002.

Description of amendment request: The proposed amendment would revise 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 would be 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 would be 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.”

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 following NSHC in its application dated September 3, 2002.

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

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

The proposed change 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.

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

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: August 20, 2002.

Description of amendment request: The proposed amendment would revise Technical specifications (TSs) Table 3.3.6.1-1, Function 5.a, Reactor Water Cleanup (RWCU) System Isolation, Main Steam Valve Vault Area Temperature—High, to extend the frequency of the channel calibration Surveillance Requirement (SR) from 122 days to 24 months.

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

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

No. The proposed amendment changes the channel calibration surveillance frequency from 122 days to 24 months. Under certain circumstances, TS SR would allow a maximum surveillance interval of 30 months for the SR. An instrumentation calculation in Start Printed Page 63699accordance with the guidelines of Generic Letter 91-04 has shown that the reliability of protective instrumentation will be preserved for the maximum allowable surveillance interval. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed change simply extends the channel calibration interval of instrumentation from 122 days to 24 months and does not affect plant modes of operation. Hence, the change does not create the possibility of any new failure mechanisms. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. The proposed amendment changes the instrument channel calibration surveillance interval from 122 days to 24 months. An instrumentation calculation in accordance with the guidelines of Generic Letter 91-04 has shown safety margins are preserved with the extended surveillance interval and that the TS allowable values are not changed. Therefore, it is concluded that the proposed amendment does not involve a significant reduction in a margin of safety.

The proposed amendment changes the instrument channel calibration surveillance internal from 122 days to 24 months. An instrumentation calculation in accordance with the guidelines of Generic Letter 91-04 has shown safety margins are preserved with the extended surveillance interval and that the TS allowable values are changed. Therefore, it is concluded that the Proposed amendment 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: September 3, 2002.

Description of amendment request: The proposed amendment would revise technical specifications 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 would be extended from the current limit of up to 24 hours, or up to the limit of the surveillance frequency interval, whichever is “less,” to up to 24 hours, or up to the limit of the surveillance frequency interval, whichever is “greater.” In addition, the following requirement would be 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.”

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). TVA reviewed the following proposed NSHC determination published in the Federal Register as part of the Consolidated Line Item Improvement Process for Technical Specification Task Force item 358, and concluded in its application of September 3, 2002, that the proposed NSHC determination applied to Watts Bar.

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

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

The proposed change 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.

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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902. Start Printed Page 63700

NRC Section Chief: Allen G. Howe.

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.

Nuclear Management Company, LLC, Docket No. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin

Date of amendment request: April 30, 2002.

Brief description of amendment request: The proposed amendment would increase the licensed reactor core power level by 1.4 percent from 1518.5 MWt to 1540 MWt.

Date of publication of individual notice in Federal Register: September 11, 2002 (67 FR 57630).

Expiration date of individual notice: October 11, 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, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of application for amendment: August 1, 2001, and as supplemented by letters dated June 19 and September 9, 2002.

Brief description of amendment: The amendment revises technical specification requirements that have been superseded based on the licensed operator training program being accredited by the Institute of Nuclear Power Operations, promulgation of the revised 10 CFR Part 55, and adoption of a systems approach to training as required by 10 CFR 50.120.

Date of issuance: September 24, 2002.

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

Amendment No.: 154.

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

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55009). The supplemental letters dated June 19 and September 9, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 24, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 26, 2002, as supplemented on August 1, 2002.

Brief description of amendment: The amendment revised the safety limit minimum critical power ratio values for Cycle 19 in Section 2.1.A of the Technical Specifications, and made several editorial or administrative corrections.

Date of Issuance: September 26, 2002.

Effective date: September 26, 2002, and shall be implemented before Cycle 19 startup.

Amendment No.: 233.

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

Date of initial notice in Federal Register: August 6, 2002 (67 FR 50949). The August 1, 2002, letter provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated September 26, 2002.

No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of application for amendment: August 1, 2001 as supplemented by letters dated June 19, July 19, and September 9, 2002.

Brief description of amendment: The amendment revised, clarified, and deleted, as appropriate, requirements regarding Facility Staff Qualifications and licensed operator and non-licensed personnel training programs. The changes revised requirements that have been superseded based on licensed operator training programs being accredited by the Institute for Nuclear Power Operations, promulgation of the revised 10 CFR Part 55, “Operator's Licenses,” which became effective on May 26, 1987, and adoption of a systems Start Printed Page 63701approach to training as required by 10 CFR 50.120, “Training and qualification of nuclear power plant personnel.”

Date of issuance: September 23, 2002.

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

Amendment No.: 241.

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

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55009). Exelon's June 19, July 19, and September 9, 2002, letters provided clarifying information within the scope of the original application and did not change the NRC staff's proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 23, 2002, as supplemented July 16, 2002.

Brief description of amendment: The amendment eliminates the requirement to perform response time testing for two reactor protection system functions and two primary containment isolation functions.

Date of issuance: October 2, 2002.

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

Amendment No.: 151.

Facility Operating License No. NPF-43: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42818). The July 16, 2002, supplemental letter provided additional clarifying information that was within the scope of the original application and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2002.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-423 Millstone Power Station, Unit Nos. 1, 2, and 3 New London County, Connecticut

Date of application for amendment: November 8, 2001, as supplemented August 14, 2002.

Brief description of amendment: The amendments incorporate administrative and editorial changes into the Millstone Unit No. 1 Permanently Defueled Technical Specifications (PDTS) and the Millstone Unit Nos. 2 and 3 Technical Specifications (TSs).

Date of issuance: September 17, 2002.

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

Amendment Nos.: 111, 270 and 212.

Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: This amendment revises the Unit No. 1 PDTS and the Units 2 and 3 TSs.

Date of initial notice in Federal Register: December 12, 2001 (66 FR 64290). The August 14, 2002, letter provided clarifying information that did not change the scope of the proposed action or the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 7, 2001, as supplemented by letter dated August 7, 2002.

Brief description of amendments: The amendments revised the Technical Specifications 5.6.5.a by adding a few parameter limits currently included in the Core Operating Limits Report.

Date of issuance: October 1, 2002.

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

Amendment Nos.: 202 and 195.

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

Date of initial notice in Federal Register: August 23, 2002 (67 FR 54680). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 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: October 7, 2001, as supplemented by letter dated August 7, 2002.

Brief description of amendments: The amendments revised the Technical Specifications 5.6.5.a by adding a few parameter limits currently included in the Core Operating Limits Report.

Date of issuance: October 1, 2002.

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

Amendment Nos.: 208 & 189.

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

Date of initial notice in Federal Register: August 23, 2002 (67 FR 54680). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 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: January 10, 2002.

Brief description of amendment: The amendment revised the technical specifications to extend the surveillance test interval of certain instrument channels from the current 18 months to 24 months.

Date of issuance: September 20, 2002.

Effective date: September 20, 2002, to be implemented within 30 days from the date of issuance.

Amendment No.: 179.

Facility Operating License No. NPF-21: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 6, 2002 (67 FR 50951). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2002.

No significant hazards consideration comments received: No.

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: September 24, 2001, as supplemented by letters dated April 22 and July 29, 2002.

Brief description of amendment: The amendment extends the allowed outage time for a Division I or Division II Start Printed Page 63702Emergency Diesel Generator (EDG) from 72 hours to 14 days. The changes are intended to provide flexibility in scheduling EDG maintenance activities, reduce refueling outage duration, and improve EDG availability during plant shutdowns.

Date of issuance: September 25, 2002.

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

Amendment No.: 125.

Facility Operating License No. NPF-47: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 12, 2001 (66 FR 64292). The April 22 and July 29, 2002, supplemental letters provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 25, 2002.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: January 31, 2002, as supplemented by letter dated September 9, 2002.

Brief description of amendment: The amendment changed administrative Technical Specification 5.5.16 regarding the Containment Integrated Leak Rate Testing (ILRT) to allow a one-time extension of the interval (to 15 years) for performance of the next ILRT.

Date of issuance: September 24, 2002.

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

Amendment No.: 219.

Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7417). The September 09, 2002, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 24, 2002.

No significant hazards consideration comments received: No.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: August 1, 2001, as supplemented by letters dated June 19 and September 9, 2002.

Brief description of amendments: The amendments would revise requirements that have been superseded based on licensed operator training programs being accredited by the Institute for Nuclear Power Operations, promulgation of the revised 10 CFR Part 55, “Operators” Licenses,” which became effective on May 26, 1987, and adoption of a systems approach to training as required by 10 CFR 50.120, “Training and qualification of nuclear power plant personnel.”

Date of issuance: September 24, 2002.

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

Amendment Nos.: 130 and 125.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55018). The supplements dated June 19 and September 09, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 24, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 15, 2002, as supplemented July 8, 2002.

Brief description of amendments: The amendments change Technical Specification surveillance requirements and allowable values for reactor vessel steam dome pressure—high instrumentation to reflect replacement of pressure switches with analog units.

Date of issuance: October 2, 2002.

Effective date: As of the date of issuance and shall be implemented within 90 days for Unit 3 and prior to startup from the next refueling outage for Unit 2.

Amendment Nos.: 195 & 188.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: May 28, 2002 (67 FR 36930). The supplement dated July 8, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 2, 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: May 14, 2002, as supplemented by letter dated September 5, 2002.

Brief description of amendments: These amendments revised technical specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. A TS Bases Control Program is added to the TSs. Additionally, two administrative changes affecting TS Section 6.2.2, “Unit Staff,” and Section 6.5.1.2, “Composition,” were incorporated.

Date of issuance: October 2, 2002.

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

Amendment Nos.: 162 and 124.

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

Date of initial notice in Federal Register: August 27, 2002 (67 FR 55041). The supplement dated September 5, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 27, 2002 (67 FR 55041).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 2, 2002.

No significant hazards consideration comments received: No. Start Printed Page 63703

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 2, York County, Pennsylvania

Date of application for amendment: June 10, 2002, as supplemented August 2, 2002.

Brief description of amendment: This amendment revises the Technical Specifications (TSs) for the safety limit for the minimum critical power ratio from its current value of 1.09 to 1.07 for two recirculation-loop operation, and from 1.10 to 1.09 for single recirculation-loop operation.

Date of issuance: September 23, 2002.

Effective date: As of the date of issuance, to be implemented prior to startup for cycle 15 operations, scheduled for September 2002.

Amendment No.: 246.

Facility Operating License No. DPR-44: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 6, 2002 (67 FR 50953). The August 2, 2002, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 23, 2002.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 21, 2002.

Brief description of amendment: The amendment modifies the Technical Specification Surveillance Requirement 3.7.3.1 to improve consistency with Cooper Nuclear Station (CNS) Amendment No. 185, approved on March 13, 2001, and eliminate unnecessary restrictions regarding how the reactor equipment cooling system surge tank level is monitored.

Date of issuance: September 18, 2002.

Effective date: September 18, 2002.

Amendment No.: 194

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 2, 2002 (67 FR 15624). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 18, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 24, 2002.

Brief description of amendment: The amendment revised Surveillance Requirement 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period was 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 was 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: October 2, 2002.

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

Amendment No.: 107.

Facility Operating License No. NPF-69: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 20, 2002 (67 FR 53987). The staff's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 19, 2001, as supplemented April 19, 2002.

Brief description of amendment: The amendment extends the time for completing required action A.1 of TS 3.8.4, “Electrical Power Systems—DC Sources—Operating,” for restoring the 125 volt direct current (VDC) electrical power subsystem to operable status. The change, in effect, provides for replacement of 125 VDC batteries 1D1 and 1D2 while the plant is at power. The time is extended on a one-time basis, and for each battery division separately, from 8 hours to 10 days. The one-time change also requires that required features be declared inoperable when the associated 125 VDC source is inoperable and the redundant required features are also inoperable for at least 4 hours.

Date of issuance: October 1, 2002.

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

Amendment No.: 247.

Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 5, 2002 (67 FR 5329). The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2002.

No significant hazards consideration comments received: No.

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

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

Brief description of amendment: Amendment changes Technical Specification 5.0 to be consistent with Technical Specifications Task Force Change No. 258, Revision 4, “Changes to Section 5.0, Administrative Controls.”

Date of issuance: October 2, 2002.

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

Amendment No.: 248.

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

Date of initial notice in Federal Register: April 2, 2002 (67 FR 15625). The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 7, 2002, as supplemented August 20 and 29, 2002.

Brief description of amendment: The amendment would revise the Kewaunee Nuclear Power Plant Technical Specification (TS) Sections for administrative changes:

(1) Section 1-“Definitions,” (2) Section 2-“Safety Limits and Limiting Safety System Settings,” (3) Section 5-“Design Features,” and (4) Section 6-“Administrative Controls.”Start Printed Page 63704

The administrative changes include capitalizing defined words, formatting section titles, renumbering pages and correcting miscellaneous grammar and punctuation errors.

Date of issuance: September 19, 2002.

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

Amendment No.: 162.

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

Date of initial notice in Federal Register: July 23, 2002 (67 FR 48220). The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 19, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 7, 2002, as supplemented July 19 and September 11, 2002.

Brief description of amendment: The amendment revised the Kewaunee Nuclear Power Plant technical specification (TS) requirements for meeting surveillances in TS 4.0.a, TS requirements for missed surveillances in TS 4.0.c, and TS requirements for a Bases control program consistent with TS Bases Control Program described in Section 5.5 of NUREG-1431, Standard TS for Westinghouse Plants, Revision 2.

Date of issuance: September 24, 2002.

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

Amendment No.: 163.

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

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42829). The supplements dated July 19 and September 11, 2002, provided clarifying information that did not change the scope of the May 7, 2002, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 24, 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: December 21, 2001, as supplemented April 26, 2002.

Brief description of amendment: The amendment revises Technical Specifiction (TS) Sections 3.7/4.7, “Containment Systems,” to (1) clarify existing requirements, (2) make editorial changes, (3) revise limiting conditions for operation (LCOs) and surveillance requirements, and (4) add certain LCOs.

Date of issuance: September 23, 2002.

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

Amendment No.: 130.

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

Date of initial notice in Federal Register: May 14, 2002 (67 FR 34490). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 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: October 17, 2001, as supplemented June 25, 2002.

Brief description of amendment: The amendment revises the multiplier values for the single-loop operation average planar linear heat generation rate to account for the use of General Electric (GE)14 fuel.

Date of issuance: October 2, 2002.

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

Amendment No.: 131.

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

Date of initial notice in Federal Register: November 14, 2001 (66 FR 57122). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 13, 2001, as supplemented by letter dated February 27, 2002.

Brief description of amendments: The amendment revises Technical Specification (TS) 3.7.16, “Spent Fuel Pool Boron Concentration”; TS 3.7.17, “Spent Fuel Assembly Storage—Region 1/Region 2”; and TS 4.3, “Fuel Storage,” for Diablo Canyon Nuclear Power Plant Units 1 and 2, to allow the use of credit for soluble boron in the spent fuel pool criticality analysis.

Date of issuance: September 25, 2002.

Effective date: September 25, 2002, and shall be implemented within 30 days from the date of issuance.

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

Facility Operating License Nos. DPR-80 and DPR-82: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55020). The February 27, 2002, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 25, 2002.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

Date of amendment request: March 4, 2002, as supplemented by letter dated July 11, 2002.

Brief Description of amendment: The proposed amendment revises Technical Specifications (TS) 5.5.9.3.a, “Steam Generator Tube Surveillance Program, Inspection Frequencies.” Specifically, the proposed changes revise the Farley Nuclear Plant, Unit 1 TS to allow a 40-month inspection interval after its first (post-replacement) inservice inspection, rather than after two consecutive inspections resulting in C-1 classification.

Date of issuance: September 20, 2002.

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

Amendment No.: 157.

Facility Operating License No. NPF-2: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 20, 2002 (67 FR 53991). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2002. Start Printed Page 63705

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 20, 2001, as supplemented by letters dated January 24, April 25, July 3, and July 16, 2002.

Brief description of amendments: The amendments revise the Technical Specifications to support extension of certain surveillance requirements from “92 days” to “92 days on an alternate test basis.”

Date of issuance: September 26, 2002.

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

Amendment Nos.: 234 and 176.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 28, 2001 (66 FR 59514). The supplements dated January 24, April 25, July 3, and July 16, 2002, provided clarifying information that did not change the scope of the September 20, 2001, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 26, 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: August 2, 2001, as supplemented by letters dated March 6, April 2, and June 25, 2002. The supplemental information provided clarification that did not change the scope or the initial no significant hazards consideration determination.

Brief description of amendments: The amendments revise the Technical Specification permitting a one time extension of Title 10 of the Code of Federal Regulations, Part 50, Appendix J, Option B, Performance-Based Leakage-Test Requirements.

Date of issuance: September 17, 2002.

Effective date: September 17, 2002.

Amendment Nos.: Unit 1-143; Unit 2-131.

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

Date of initial notice in Federal Register: August 6, 2002 (67 FR 50959). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: March 4, 2002 (TSC 00-04).

Description of amendment request: The amendments relocated certain Technical Specification (TS) surveillance requirements to the Sequoyah Technical Requirements Manual.

Date of issuance: September 5, 2002.

Effective date: September 5, 2002.

Amendment Nos: 277 and 268.

Facility Operating License No. DPR-79: Amendments revise the TSs.

Date of initial notice in Federal Register:April 16, 2002 (67 FR 18648). 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.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: September 21, 2001, as supplemented by letters dated June 11, July 19, August 9 and 30, and September 5 and 12, 2002 (TS 00-06).

Brief description of amendments: The amendments revised the Technical Specifications (TSs) to allow irradiation of up to 2256 tritium-producing burnable absorber rods.

Date of issuance: September 30, 2002.

Effective date: As of the date of issuance and shall be implemented prior to irradition of TPBARs.

Amendment Nos.: Unit 1-278, Unit 2-269.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the TSs.

Date of initial notice in Federal Register: December 17, 2001 (66 FR 65000). The supplemental letters provided clarifying information that did not expand the application beyond the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in an Environmental Assessment dated September 23, 2002 (67 FR 59581) and in a Safety Evaluation dated September 30, 2002.

No significant hazards consideration comments received: Comments were received in response to the staff's proposed no significant hazards consideration determination that was published in the December 17, 2001, Federal Register, from Dr. Kenneth D. Bergeron and The Blue Ridge Environmental Defense League (BREDL). BREDL's comments incorporated Mr. Bergeron's comments by reference. These comments were addressed by the staff in a letter from Dr. Brian Sheron to Mr. Bergeron dated September 6, 2002, with a copy to BREDL (Accession No. ML022410310).

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: September 12, 2001, as supplemented September 17, 2002 (TS 01-04).

Brief description of amendments: The proposed amendments would change the Sequoyah Nuclear Plant, Units 1 and 2 Technical Specification (TS) 3/4 6.5.1 and associated Bases to reflect an increase in the ice condenser basket weight from 1071 pounds to 1145 pounds and the total ice condenser ice weight from 2,082,024 pounds to 2,225,880 pounds. This change is being made in response to a reanalysis by Westinghouse Electric Company that identified a modeling input error used in the original analysis.

Date of issuance: September 30, 2002.

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

Amendment Nos.: 279 & 270.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the TS.

Date of initial notice in Federal Register: January 8, 2002 (67 FR 934). The September 17, 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 amendment is contained in a Safety Evaluation dated September 30, 2002.

No significant hazards consideration comments received: No. Start Printed Page 63706

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

Date of application for amendment: August 20, 2001, as supplemented by letters of October 29, November 14, November 21, December 7, December 19, 2001, and January 14, February 19, February 21, May 21, May 23, and July 30, 2002.

Brief description of amendment: The amendment allows Watts Bar Nuclear Plant, Unit 1, to irradiate up to 2304 tritium-producing burnable absorber rods in the reactor core each fuel cycle.

Date of issuance: September 23, 2002.

Effective date: As of the date of issuance and shall be implemented prior to starting up from the outage where TVA inserts tritium-producing burnable absorber rods in the core.

Amendment No.: 40.

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

Date of initial notice in Federal Register : December 17, 2001 (66 FR 65005). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in an Environmental Assessment dated August 20, 2002 (ADAMS Accession No. ML022320905) and in a Safety Evaluation dated September 23, 2002.

No significant hazards consideration comments received: Comments were received in response to the staff's proposed no significant hazards consideration determination (66 FR 65005) from Dr. Kenneth D. Bergeron and The Blue Ridge Environmental Defense League (BREDL). BREDL's comments incorporated Dr. Bergeron's comments by reference. These comments were addressed by the staff in a letter from Dr. Brian Sheron to Dr. Bergeron dated September 6, 2002, with a copy to BREDL (Accession No. ML022410310). The staff made a final determination that the amendment involves no significant hazards consideration, which is contained in the Safety Evaluation dated September 23, 2002.

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

Date of amendment request: October 23, 2001, as supplemented by letters dated July 23, August 29, and September 6, 2002.

Brief description of amendments: The amendments revise TS 5.5.9, “Steam Generator Tube Inspection Report,” to permit installation of leak-tight sleeves in the Comanche Peak Steam Electric Station, Unit 1, steam generators as an alternative to plugging defective steam generator tubes.

Date of issuance: September 25, 2002.

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

Amendment Nos.: 101.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register : December 26, 2001 (66 FR 66473). The July 23, August 29, and September 6, 2002, supplemental letters provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 25, 2002.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

Date of application for amendment: June 17, 2002.

Brief description of amendment: The amendment revises Limiting Conditions for Operation (LCOs), Required Actions for LCOs, Surveillance Requirements, and Tables specifying requirements on instrumentation in the following Technical Specifications: (1) TS 3.3.6, “Containment Purge Isolation Instrumentation”; (2) TS 3.3.7, “Control Room Emergency Ventilation System (CREVS) Instrumentation”; (3) TS 3.3.8, “Emergency Exhaust System (EES) Actuation Instrumentation”; and (4) TS 3.9.4, “Containment Penetrations.” The revisions allow the equipment hatch and the emergency air lock to be open in refueling outages during core alterations and/or movement of irradiated fuel within containment.

Date of issuance: September 9, 2002.

Effective date: September 9, 2002, and shall be implemented, including the incorporation of the changes to the Bases of the Technical Specifications and to the Final Safety Analysis Report for Callaway, as described in the licensee's letter of June 17, 2002, prior to entry into Mode 6 during Refueling Outage 12 that is scheduled for the Fall of 2002.

Amendment No.: 152.

Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register : July 23, 2002 (67 FR 48222). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 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: February 26, 2002, as supplemented by letter dated July 15, 2002.

Brief description of amendment: These amendments revise the surveillance frequency of the quench and recirculation spray system nozzles, from a time period of every 10 years to whenever maintenance is conducted that could contribute to nozzle blockage.

Date of issuance: October 1, 2002.

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

Amendment Nos.: 233 and 215.

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

Date of initial notice in Federal Register : April 30, 2002 (67 FR 21296). The supplemental letter dated July 15, 2002, provided clarifying information that did not change the scope of the February 26, 2002, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2002.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 4th day of October 2002.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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

End Signature End Preamble

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 paragraphs (d)(1) and (d)(2) regarding petitions to intervene and contentions. For the complete, corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April 29, 2002.

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[FR Doc. 02-25990 Filed 10-11-02; 8:45 am]

BILLING CODE 7590-01-P