02-11871. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 April 19, 2002 through May 2, 2002. The last biweekly notice was published on April 30, 2002 (67 FR 21283).

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 NRC'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 June 13, 2002, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the NRC'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 Systems (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. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and 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 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 PDR Reference staff at 1-800-397-4209, 304-415-4737 or by e-mail to [e-mail].

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: April 16, 2002.

Description of amendment request: The proposed amendments 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 determination in its application dated April 16, 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 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.

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

Date of amendment request: April 16, 2002.

Description of amendment request: The proposed amendments 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 determination in its application dated April 16, 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 proposes to determine that the amendments request involves no significant hazards consideration.

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

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: April 16, 2002.

Description of amendment request: The proposed amendments 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 determination in its application dated April 16, 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 proposes to determine that the amendments request involves no significant hazards consideration.

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

NRC Section Chief: John A. Nakoski.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: March 28, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) Section 3.7, “Auxiliary Electrical Systems,” and Section 4.6, “Emergency Power System Periodic Tests,” to relocate the requirements for the gas turbine generators to the Updated Final Safety Analysis Report and the plans, programs and procedures that document and control the credited functions of these systems, structures, and components. The proposed amendment would also delete TS 3.7.B.2.b to remove the option that allows power operation for up to 72 hours with a gas turbine as the only available 13.8 kilovolt power source.

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

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

The Gas Turbine Generators only provide a Licensing Basis Event mitigating function. There is no previously evaluated accident or event that is initiated by the Gas Turbine Generators or the associated fuel storage system. The ability of the Gas Turbine Generators to provide power, as a backup to the Emergency Diesel Generators, is not affected by the location of the description of their licensing basis.

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

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

There is no physical change to the plant. The currently existing gas turbine generators and associated fuel oil storage facilities will still be used. The only change is to relocate the limiting conditions for operations, surveillance requirements and associated bases from the Technical Specifications to other licensee controlled documents.

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

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in [a] margin of safety.

The deletion of the limiting conditions for operation and surveillance requirements for the gas turbine generators from the Technical Specifications does not alter the method of operation, the design requirements or the current licensing basis that the gas turbine generators be able to power all the loads required by 10 CFR Part 50, Appendix R to place the plant into a safe shutdown condition following a fire and maintain safe shutdown for three days. It also does not remove the licensing basis requirement of 10 CFR Part 50, Section 50.63, that the unit must have the capacity to withstand and recover from a station blackout. The current licensing basis will continue to credit the gas turbine generators as the alternate ac (AAC) power source in the event of a station blackout unless modified under the control of 10 CFR Part 50, Section 50.59.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in [a] margin of safety.

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

Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: April 11, 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 (LCO), 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 U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400), on possible amendments concerning missed surveillances, including a model safety evaluation 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 determination in its application dated April 11, 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 [a] Margin of Safety

The extended time allowed to perform a missed surveillance does not result in a significant reduction in [a] margin of safety. As supported by the historical data, the likely outcome of any surveillance is verification that the LCO 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 [a] 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: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Section Chief: Richard J. Laufer.

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 amendment request: March 8, 2002.

Description of amendment request: The proposed amendment is consistent with Technical Specifications Task Force (TSTF) Standard Technical Specification (TS) Change Traveler TSTF-360, Revision 1 and TSTF-204, Revision 3 and proposes to revise TS 3.8.4, “DC Sources—Operating,” TS 3.8.5, “DC Sources—Shutdown,” TS 3.8.6, “Battery Cell Parameters,” and TS 3.8.8, “Inverters—Shutdown.” The changes associated with TSTF-360, Revision 1, add new Required Actions and extend the Completion Times in TS 3.8.4 and TS 3.8.5 and also include the relocation to a licensee-controlled program of a number of Surveillance Requirements (SRs) in TS 3.8.4 and TS 3.8.6. The changes associated with TSTF-204, Revision 3, revise TS 3.8.5 and TS 3.8.8 to change requirements for DC electrical power subsystem and inverters.

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

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

The proposed changes revise TS 3.8.4, “DC Sources—Operating,” TS 3.8.5, “DC Sources—Shutdown,” TS 3.8.6, “Battery Cell Parameters,” and TS 3.8.8, “Inverters—Shutdown.”

TS 3.8.4, TS 3.8.5, and TS 3.8.6 have been revised to 1) add new Required Actions and extend the Completion Time for an inoperable battery charger, 2) provide alternate battery charger testing criteria for TS 3.8.4 and TS 3.8.5, 3) relocate to a licensee-controlled program a number of Surveillance Requirements (SRs) in TS 3.8.4 that perform preventive maintenance on the safety-related batteries, 4) relocate TS Table 3.8.6-1, “Battery Cell Parameters Requirements,” to a licensee-controlled program, 5) add to TS 3.8.6 specific Required Actions associated with out-of-limits conditions for battery cell float voltage, float current, electrolyte level, and electrolyte temperature, and 6) add a new administrative TS program for the maintenance and monitoring of station batteries based on the recommendations of Institute of Electrical and Electronics Engineers (IEEE) Standard 450-1995, “IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications.” In addition, TS 3.8.5 and TS 3.8.8 have been revised to require only one DC electrical power subsystem and two inverters, respectively, during shutdown conditions.

The DC Sources, Battery Cell Parameters, and Inverters are not initiators of any accident sequence analyzed in the Byron/Braidwood Stations' Updated Final Safety Analysis Report (UFSAR). As such, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

The initial conditions of Design Basis Accident (DBA) and transient analyses in the Byron/Braidwood Stations' UFSAR assume Engineered Safety Feature (ESF) systems are operable. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. The operability of the AC and DC electrical power distribution systems in accordance with the proposed TS is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed changes do not involve any physical alteration of the units. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints at which protective or mitigative actions are initiated that are affected by the proposed changes. The operability of the AC and DC electrical power distribution systems in accordance with the proposed TS is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant. These proposed changes will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures, which ensure the unit remains within analyzed limits, is proposed, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced. The proposed changes do not alter assumptions made in the safety analyses.

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

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

The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new administrative TS program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the AC and DC electrical power distribution systems will continue to provide adequate power to safety-related loads in accordance with analyses assumptions.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Anthony J. Mendiola.

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

Date of application for amendments: March 19, 2002.

Description of amendment request: The proposed amendment would allow plant operation to continue if the temperature of the Normal Heat Sink (NHS) exceeds the Technical Specification (TS) limit of 90 °F provided the water temperature, averaged over the previous 24-hour period, is at or below 90 °F. The proposed operational flexibility would only apply if the NHS temperature is between 90 °F and 92 °F. The current action time requirements would still apply if the NHS temperature exceeds 92 °F, or if the 24-hour averaged value exceeds 90 °F. The current TS Limiting Condition for Operation (LCO) limit of 90 °F would not be changed. In addition, an administrative change would remove references to a temporary TS change which had expired on May 31, 2000. The Bases for the associated TS would also be modified.

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

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes will allow plant operation to continue if the temperature of the NHS exceeds the TS limit of 90 °F provided that: (1) The water temperature, averaged over the previous 24 hour period, is at or below 90 °F, and (2) the NHS temperature is less than or equal to 92 °F. This increase in NHS temperature will not affect the normal operation of the plant to the extent that it would make any accident more likely to occur. In addition, there exists adequate margin in the safety systems and safety-related heat exchangers to assure the design safety functions are met at the higher temperature.

The proposed administrative change to remove an expired, temporary license amendment removes information which is no longer valid.

Thus, the proposed changes will have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design-basis accidents will not change. In addition, the proposed changes can not cause an accident. Therefore, there will be no increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes will allow plant operation to continue if the temperature of the NHS exceeds the TS limit of 90 °F provided that: (1) The water temperature, averaged over the previous 24-hour period, is at or below 90 °F, and (2) the NHS temperature is less than or equal to 92 °F. This will 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 will not alter the way any structure, system, or component functions and will not significantly alter the manner in which the plant is operated. There will be no adverse effect on plant operation or accident mitigation equipment. The proposed changes do not introduce any new failure modes. Also, the response of the plant and the operators following a design-basis accident is unaffected by the changes. In addition, the NHS is not an accident initiator and the design-basis heat removal capability of the affected safety-related components is maintained at the increased NHS temperature limit. The proposed administrative change to remove an expired, temporary license amendment removes information which is no longer valid. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any previously analyzed.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

The proposed changes will allow plant operation to continue if the temperature of the NHS exceeds the TS limit of 90 °F provided that: (1) The water temperature, averaged over the previous 24-hour period, is at or below 90 °F, and (2) the NHS temperature is less than or equal to 92 °F. The licensee performed an evaluation of the safety systems to ensure their safety functions can be met with a NHS water temperature of 92 °F. The higher NHS temperature represents a slight reduction in the margins of safety in terms of these systems' abilities to remove accident heat loads. As part of its evaluation, however, the licensee verified that these safety systems will still be able to perform their design-basis functions.

The proposed administrative change to remove an expired, temporary license amendment removes information which is no longer valid.

The proposed changes will have no adverse effect on plant operation or equipment important to safety. The plant response to the design-basis accidents will not change and the accident mitigation equipment will continue to function as assumed in the design-basis accident analysis. Therefore, there will be no significant reduction in a margin of safety.

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC, Docket Nos. 50-254, Quad Cities Nuclear Power Station, Unit 1, Rock Island County, Illinois

Date of amendment request: April 8, 2002.

Description of amendment request: The amendment would revise the safety limit minimum critical power ratio for two-loop and single-loop 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:

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

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established consistent with NRC approved methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio (SLMCPR) for Quad Cities Nuclear Power Station (QCNPS), Unit 1 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The proposed change revised the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. Operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9 percent of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.

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

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

Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for QCNPS, Unit 1.

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

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

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection. Additionally, operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9 percent of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the 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, Vice President, General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Anthony J. Mendiola.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: April 18, 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 determination in its application dated April 18, 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: R. Alexander Glenn, Associate General Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. Petersburg, Florida 33733-4042.

NRC Acting Section Chief: Thomas Koshy.

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: April 11, 2002.

Description of amendment requests: The proposed amendments would revise the Surveillance Requirements for containment leakage rate testing in Technical Specification (TS) 4.6.1.2 to allow a one-time extension of the interval between integrated leakage rate tests (ILRTs) from 10 to 15 years.

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

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

Response: No.

Probability of Occurrence of an Accident Previously Evaluated—

The proposed change to extend the ILRT interval from 10 to 15 years does not affect any accident initiators or precursors. The containment liner function is purely mitigative. There is no design basis accident that is initiated by a failure of the containment leakage mitigation function. The extension of the ILRT will not create any adverse interactions with other systems that could result in initiation of a design basis accident. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated—

The potential consequences of the proposed change have been quantified by analyzing the changes in risk that would result from extending the ILRT interval from 10 to 15 years. The increase in risk in terms of person rem per year within 50 miles resulting from design basis accidents was estimated to be of a magnitude that NUREG-1493 indicates is imperceptible. I&M has also analyzed the increase in risk in terms of the frequency of large early releases from accidents. The increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in Regulatory Guide 1.174. Additionally, the proposed change maintains defense in depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. I&M has determined that the increase in conditional containment failure probability from reducing the ILRT frequency from 1 test per 10 years to 1 test per 15 years would be small. Continued containment integrity is also assured by the history of successful ILRTs, and the established programs for local leakage rate testing and inservice inspections which are unaffected by the proposed change. Therefore, the consequences of an accident previously analyzed are not significantly increased.

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

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

Response: No.

The proposed change to extend the ILRT interval from 10 to 15 years does not create any new or different accident initiators or precursors. The length of the ILRT interval does not affect the manner in which any accident begins. The proposed change does not create any new failure modes for the containment and does not affect the interaction between the containment and any other system. Thus, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The risk-based margins of safety associated with the containment ILRT are those associated with the estimated person-rem per year, the large early release frequency, and the conditional containment failure probability. I&M has quantified the potential effect of the proposed change on these parameters and determined that the effect is not significant. The non-risk-based margins of safety associated with the containment ILRT are those involved with its structural integrity and leak tightness. The proposed change to extend the ILRT interval from 10 to 15 years does not adversely affect either of these attributes. The proposed change only affects the frequency at which these attributes are verified. Therefore, the proposed changes do not involve a significant reduction in margin of safety.

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

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

NRC Section Chief: L. Raghavan.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: March 22, 2002.

Description of amendment request: The proposed amendments change Seabrook Station Technical Specification (TS) 3/4.9.13, Spent Fuel Assembly Storage, and associated TS Figures and Index. The licensee will also revise the Bases to reflect the license amendment. The proposed changes reflect a revised criticality safety analysis supporting a two-zone spent fuel pool, consisting of BORAFLEX® and Boral® fuel assembly storage racks.

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

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

The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and TS Figure 3.9-2 do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility. In addition, the proposed changes do not affect the manner in which the plant responds in normal operation, transient, or accident conditions. The changes reflect the design capability of the BORAL® storage racks to safely store spent fuel.

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 in the Seabrook Station Updated Final Safety Analysis Report (UFSAR). Furthermore, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. Therefore, the proposed changes do not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and TS Figure 3.9-2 do not change the operation or the design basis of any plant system or component during normal or accident conditions. The proposed changes do not include any physical changes to the plant. In addition, the proposed changes do not change the function or operation of plant equipment or introduce any new failure mechanisms. The plant equipment will continue to respond per the design and analyses and there will not be a malfunction of a new or different type introduced by the proposed changes. The proposed changes do not modify the facility nor do they affect the plant's response to normal, transient, or accident conditions. The changes do not introduce a new mode of plant operation. The changes reflect the design capability of the BORAL® storage racks to safely store spent fuel. The plant's design and design basis are not revised and the current safety analyses remains in effect. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and TS Figure 3.9-2 do not adversely affect the safety margins established through Limiting Conditions for Operation, Limiting Safety System Settings, and Safety Limits as specified in the Technical Specifications nor is the plant design revised by the proposed changes. The safety margins established through Limiting Conditions for Operation, Limiting Safety System Settings, and Safety Limits as specified in the Technical Specifications are not revised nor is the plant design or its method of operation revised by the proposed changes. The changes reflect the design capability of the BORAL® storage racks to safely store spent fuel. Administrative control measures (e.g., procedures) will continue to be in place to ensure the safe placement of fuel assemblies within the spent fuel pool so as to remain less than or equal to 0.95 K_eff as required by TS 5.6.1.1 for spent fuel storage. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William J. Quinlan, Esq., Assistant General Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: December 21, 2001.

Description of amendment request: The proposed amendment would revise the Containment Systems Section of the Technical Specification (TS) to clarify existing requirements, make wording improvements, revise existing limiting condition for operations (LCO) and surveillance requirements (SR), and add an additional TS LCO to the Monticello TS.

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

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

The proposed TS changes do not introduce new equipment or new equipment operating modes, nor do the proposed changes alter existing system relationships. Providing additional time to correct a situation in which suppression pool water level may be outside the established limits, deleting an unnecessary TS regarding suppression pool water level instrumentation, adding a time limit in which to restore oxygen concentration in the containment to within limits, and clarifying specific use and actions for Primary Containment Isolation Valves, are not initiators of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment referenced in the proposed changes is still required to be operable and capable of performing its accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected.

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

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

The proposed changes do not involve physical alterations of the plant, no new or different type of equipment will be installed. Nor, are there significant changes in the methods governing normal plant operation. Providing additional time to correct a situation in which suppression pool water level may be outside the established limits, deleting an unnecessary TS regarding suppression pool water level instrumentation, restructuring the TS to provide clear Action Statements where needed; adding a time limit in which to restore oxygen concentration in the containment to within limits; and clarifying specific use and actions for Primary Containment Isolation Valves will not lead to an accident beyond those previously evaluated.

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

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

Providing additional time to correct a situation in which suppression pool water level may be outside the established limits, deleting an unnecessary TS regarding suppression pool water level instrumentation, restructuring the TS to provide clear Action Statements where needed; adding a time limit in which to restore oxygen concentration in the containment to within limits; and clarifying specific use and actions for Primary Containment Isolation Valves does not result in a significant reduction in the margin of safety. Allowing up to 2 hours to restore level, is acceptable because the suppression pool water level does not change rapidly during normal operation, and during operations that do create changes to the suppression pool water level, the level of the pool is closely monitored. The changes that provide specific LCO action statements for allowed time to place the reactor in a condition in which the LCO is no longer applicable are acceptable based on industry practices and engineering judgements. Adding an additional LCO which places a specified time limit on oxygen concentration greater than or equal to 4% by volume is acceptable because it provides a TS requirement which limits additional oxygen in the containment. Providing a revision to the LCO for inoperable primary containment isolation valves is acceptable because it clarifies what is specifically required for this method of isolation, and changing the interval at which deactivated and isolated valves must be recorded from daily to monthly is acceptable because the devices are operated under administrative controls and the probability of their misalignment is low. Relocating TS requirements is acceptable because it places the requirement for limiting the use of the purge and vent valves in a more appropriate TS and rewording the LCO is acceptable because it provides clarification for use of the purge and vent valves.

Therefore, these proposed changes will not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: March 20, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.7.8, “Service Water (SW) System,” which is applicable in Modes 1, 2, and 3, to allow the SW system to be operable with five operable SW pumps, provided one Unit is in Mode 5 or Mode 6, or defueled, and the SW system is capable of providing required cooling water flow to required equipment. The proposed amendment would change the existing TS requirement which now requires that both units be in Mode 5 (cold shutdown) within 36 hours if five of the total of six SW pumps are operable.

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

1. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a significant increase in the probability or consequences of any accident previously evaluated.

The SW System is primarily a support system for systems required to be operable for accident mitigation. Failures within the SW System are not an initiating condition for any analyzed accident.

The SW System removes the required heat from the containment fan coolers and residual heat removal heat exchangers ensuring containment pressure and temperature profiles following an accident are as evaluated in the [Final Safety Analysis Report] FSAR. This in turn ensures that environmental qualification of equipment inside containment is maintained and thus function as required post-accident. Single Unit operation with five operable SW pumps will continue to be capable of supplying the required cooling water flow to systems required for accident mitigation.

Therefore, the consequences of an accident previously evaluated will not be significantly increased as a result of the proposed change.

2. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a new or different kind of accident from any accident previously evaluated.

The possibility for a new or different type of accident from any accident previously evaluated is not created as a result of this amendment. The evaluation of the effects of the proposed changes indicate that the SW System will be able to perform all of its design basis functions within the design limits of the system. These changes do not introduce any new or different normal operation or accident initiators. Therefore, operation of the SW System as proposed will not create any new failure mechanisms.

Equipment important to safety will continue to operate as designed. The changes do not result in any event previously deemed incredible being made credible. The changes do not result in more adverse conditions or result in any increase in the challenges to safety systems. Therefore, operation of the Point Beach Nuclear Plant in accordance with the proposed amendment will not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not result in a significant reduction in a margin of safety.

The SW System functions to mitigate the effects of accidents. There are no new or significant changes to the initial conditions contributing to accident severity or consequences. The proposed amendment will not otherwise affect the plant protective boundaries, will not cause a release of fission products to the public, nor will it degrade the performance of any other SSCs [structure, system and components] important to safety. Therefore, reducing the required number of operable SW pumps from six to five with one Unit in Mode 5 or 6, or defueled, while maintaining the capability of required flow to required equipment, will not result in 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: John H. O'Neill, Jr., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: L. Raghavan.

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

Date of amendment request: March 27, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 1.3.1, “Limiting Safety Systems Settings, Reactor Protective System,” to change the high power trip setpoint from 107.0% to 109.0%. This complies with the regulatory requirements in 10 CFR part 50 Appendix A, Criterion 10 and 20 by continuing to protect the fuel from exceeding the design basis 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 change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The change does not result in a high power trip setpoint that will cause the analysis value of 112.0% to be exceeded. There is no change in the analysis value of 112.0% for the high power trip setpoint used in the evaluation of the transients and accidents. All of the evaluated transients and accidents currently show acceptable results and will not be affected by this change. Changing the high power trip setpoint will not affect the probability of an accident, since that circuit is not a transient or accident initiator. The change to the setpoint will not change the failure possibilities for this circuit. The effect of the proposed change is the reduction in the probability of an undesired safety system challenge initiated by an erroneous high power trip during a flow streaming event.

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

The change to the RPS [reactor power system] high power trip setpoint does not provide the possibility of the creation of a new or different type of accident. Changing the setpoint does not change the method of operation of the high power trip circuit or its expected response once the setpoint is reached. The trip will occur within previously analyzed limits.

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

The proposed setpoint change does not constitute a significant reduction in the margin of safety due to the fact that the transient and accident analyses contained in the Updated Safety Analysis Report have been evaluated using an analysis trip setpoint of 112.0% with the event initiated from the appropriate power level and have been shown to produce acceptable results.

The acceptance criteria used in the analysis have been developed for the purpose of use in design basis accident analyses such that meeting these limits demonstrates adequate protection of public health and safety. An acceptable margin of safety is inherent in these licensing limits. Therefore, the proposed changes do not involve a reduction in a margin of safety.

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

Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 29, 2002.

Description of amendment request: The proposed amendment would modify the Technical Specifications to allow the use of the pressure-temperature curves approved in Amendment No. 131 for an additional cycle.

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

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

The proposed amendment to revise the technical specifications to extend the use of the pressure-temperature (P-T) limits does not affect the operation or configuration of any plant equipment. Thus, no new accident initiators are created by this change. The proposed change extends the use of the P-T limits for an additional cycle. The P-T limits are based on the projected reactor vessel neutron fluence at 32 effective full power years (EFPY) of operation. At the end of cycle 10, Hope Creek Generating Station (HCGS) was at approximately 12.2 EFPY of operation (38.1% of the 32 EFPY). At the end of cycle 12 there will remain sufficient margin to ensure that the current 32 EFPY fluence projections will not be exceeded. This ensures that the basis for proposed applicability of the current P-T limits is conservative for use until the end of cycle 12 ensuring that the reactor vessel integrity is protected under all operating conditions. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed amendment revises the technical specifications to extend the use of the pressure-temperature (P-T) limits. It does not change the design function or operation of any systems, structures, or components. Plant operation will not be affected by the proposed amendments and no new failure mechanisms, malfunctions or accident initiators will be created. The current P-T limits will remain valid and conservative during the proposed extension period. The proposed change, therefore, 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?

The proposed change extends the use of the current P-T limits for an additional cycle of operation. The P-T limits are based on the projected reactor vessel neutron fluence at 32 EFPY of operation. At the end of cycle 10 in April 2000, HCGS was at approximately 12.2 EFPY of operation (38.1% of the 32 EFPY). At the end of cycle 12, HCGS will have obtained less than 50% of the 32 EFPY operating time which provides significant margin to ensure that the current 32 EFPY fluence projection will not be exceeded. The current margin of safety for plant operations is established by the P-T curves analyzed at 32 EFPY. Because the proposed change will not exceed this fluence, the current margin of safety is maintained. The proposed change, therefore, does not involve a significant reduction in a margin of safety.

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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 3, 2002.

Description of amendment request: The proposed amendment would relocate parts of Technical Specification (TS) 3/4.4.4, “Reactor Coolant System—Chemistry,” from the TS to the Updated Final Safety Analysis Report (UFSAR).

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 is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. Conductivity, chloride, and pH limits are not assumed to be an initiator of any analyzed event, nor are these limits assumed in the mitigation of consequences of accidents.

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 does not involve the modification of any plant equipment and does not change the method by which any safety-related system performs its function. The current safety analysis assumptions are not altered as a result of this change.

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 change represents the relocation of current TS requirements to the UFSAR based on regulatory guidance and previously approved changes for other stations. The proposed change is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. Margins of safety are unaffected by requirements that are retained but relocated from the TS to the UFSAR.

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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: March 25, 2002, as supplemented by the letter dated April 23, 2002.

Brief description of amendments: The proposed change would revise the current Technical Specification (TS) 3.7.3 to adopt the version of the same TS in NUREG-1431, “Standard Technical Specifications for Westinghouse Plants,” Revision 2, to add, among other things, operability requirements for Feedwater Control Valves (FCV) and Associated Bypass Valves, and would allow for the extended out-of-service time for one or more Feedwater Isolation valves (FIVs). In addition, a footnote, which allowed a one-time extension for Condition A Completion Time, is being deleted because it is no longer applicable.

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

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

Response: No.

The proposed change extends the Completion Time for one or more Feedwater Isolation Valves (FIVs) inoperable from 4 hours to 72 hours. Extending the Completion Time is not an accident initiator and thus does not change the probability that an accident will occur. However, it could potentially affect the consequences of an accident if an accident occurred during the extended unavailability of the inoperable FIV. The increase in time that the FIV is unavailable is small and the probability of an event occurring during this time period, which would require isolation of the Main Feedwater flow paths, is low. Moreover, the redundancy provided by the Feedwater Control Valves, which have [the] same actuation signals and closure time requirements as the FIVs, provides adequate assurance that automatic feedwater isolation will occur if called upon.

The deletion of the footnote, which is no longer applicable, is an administrative change and does not affect the probability or consequences of an accident previously evaluated.

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

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

Response: No.

Closure of the FIVs is required to mitigate the consequences of a Main Steam Line Break and Main Feedwater Line Break accidents. The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

The deletion of the footnote, which is no longer applicable, is an administrative change and does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes do not change any Technical Specification Limit or accident analysis assumption. Therefore they do not involve a reduction in a margin of safety.

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: March 27, 2002.

Brief description of amendments: The proposed change would revise Technical Specification (TS) 5.3.1 to require that each member of the unit staff, with the exception of Licensed Reactor Operators (RO) and Licensed Senior Reactor Operators (SRO), shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 2, 1987. Also, a new TS 5.3.2 would be added to require that the Licensed RO and Licensed SRO shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 3, May 2000 and the current TS 5.3.2 would be renumbered to TS 5.3.3.

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

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

Response: No.

The proposed TS change is an administrative change to clarify the current requirements for licensed operator qualifications and licensed operator training program. These changes conform to the current requirements of 10 CFR [Part] 55. The TS requirements for all other unit staff qualifications remain unchanged.

Although licensed operator qualifications and training may have an indirect impact on accidents previously evaluated, the NRC [Nuclear Regulatory Commission] considered this impact during the rulemaking process, and by promulgation of the revised 10 CFR [Part] 55 rule, concluded that this impact remains acceptable as long as the licensed operator training program is certified to be accredited and is based on a systems approach to training. TXU Energy's [TXU Generation Company LP] licensed operator training program is accredited by INPO [Institute of Nuclear Power Operations] and is based on a systematic approach to training.

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

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

Response: No.

The proposed TS change is an administrative change to clarify the current requirements for licensed operator qualifications and [the] licensed operator training program, and to conform to the revised 10 CFR [Part] 55. The TS requirements for all other unit staff qualifications remain unchanged.

As noted above, although licensed operator qualifications and training may have an indirect impact on the possibility of a new or different kind of accident from any accident previously evaluated, the NRC considered this impact during the rulemaking process, and by promulgation of the revised rule, concluded that this impact remains acceptable as long as the licensed operator training program is certified to be accredited and based on a systems approach to training. As previously noted, TXU Energy's licensed operator training program is accredited by INPO and is based on a systems approach to training.

Additionally, the proposed TS change does not affect plant design, hardware, system operation, or procedures. 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 TS change is an administrative change to clarify the current requirements applicable to licensed operator qualifications and licensed operator training program. This change is consistent with the requirements of 10 CFR [Part] 55. The TS qualification requirements for all other unit staff remain unchanged.

Licensed operator qualifications and training can have an indirect impact on a margin of safety. However, the NRC considered this impact during the rulemaking process, and by promulgation of the revised 10 CFR [Part] 55, determined that this impact remains acceptable when licensees maintain a licensed operator training program that is accredited and based on a systems approach to training. As noted previously, TXU Energy's licensed operator training program is accredited by INPO and is based on a systems approach to training.

The NRC has concluded, as stated in NUREG-1262, “Answers to Questions at Public Meetings Regarding Implementation of Title 10, Code of Federal Regulations, Part 55 on Operators' Licenses,” that the standards and guidelines applied by INPO in their training accreditation program are equivalent to those put forth or endorsed by the NRC. As a result, maintaining an INPO-accredited, systems approach-based licensed operator training program is equivalent to maintaining [an] NRC-approved licensed operator training program which conform[s] with applicable NRC Regulatory Guides or NRC-endorsed industry standards. The margin of safety is maintained by virtue of maintaining an INPO-accredited licensed operator training program.

In addition, the NRC has recently published NRC Regulatory Issue Summary 2001-01, “Eligibility of Operator License Applicants,” dated January 18, 2001, “* * * to familiarize addressees with the NRC's current guidelines for the qualification and training of reactor operator (RO) and senior operator (SO) license applicants.” This document again acknowledges that the INPO National Academy for Nuclear Training (NANT) guidelines for education and experience, outline acceptable methods for implementing the NRC's regulations in this area.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: April 1, 2002.

Brief description of amendments: This proposed amendment would include topical report ERX-2001-005, “ZIRLO^TM Cladding and Boron Coating Models for TXU Electric's Loss of Coolant Accident Analysis Methodologies,” in the list of approved methodologies for use in generating the Core Operating Limits Report in Technical Specification (TS) 5.6.5, “Core Operating Limits Report (COLR).” In addition, the proposed change would include ZIRLO^TM clad in the description of the fuel assemblies in TS 4.2.1, “Fuel Assemblies.”

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

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

Response: No.

Administrative changes to the Technical Specifications that do not affect the accident analyses cannot change the probability of an accident previously evaluated, nor will it increase radiological consequences predicted by the analyses of record. Controlling the use of fuel assemblies within limitations previously approved by the NRC [U.S. Nuclear Regulatory Commission] constrains fuel performance to within limits bounded by existing design basis accident and transient analyses.

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

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

Response: No.

Use of ZIRLO^TM clad fuel assemblies in accordance with NRC approved methodologies and of a design approved by the NRC ensures that their effect on core performance remains within existing design limits. Use of fuel assemblies whose design has been previously approved by the NRC is consistent with current plant design bases, does not adversely affect any fission product barrier, and does not alter the safety function of safety significant systems, structures and components or their roles in accident prevention or mitigation. Currently licensed design basis accident and transient analyses of record remain valid.

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

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

Response: No.

The proposed change does not alter the manner in which Safety Limits, Limiting Safety System Setpoints, or Limiting Conditions for Operation are determined. This proposed change to TSs 4.2 and 5.6.5 is bounded by existing limits on reactor operation. It leaves current limitations for use of fuel assemblies in place, conforms to plant design bases, is consistent with the safety analyses as accepted in the topical report, and limits actual plant operation within analyzed and NRC approved boundaries.

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

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of application request: February 15 and November 7, 2001, and March 1, 2002.

Description of amendment request: The proposed amendment would revise paragraph d.1.j (2) in Technical Specification (TS) 5.5.9, “Steam Generator (SG) Tube Surveillance Program.” The revision would (1) delete the requirement that all SG tubes containing an Electrosleeve, a Framatome proprietary process, be removed from service within two operating cycles following installation of the first Electrosleeve; (2) add the requirement that Electrosleeves will not be installed in the outermost periphery tubes of the SG bundles where potentially locked tubes would cause high axial loads; (3) revise the references describing electrosleeving; and (4) add the requirement that all sleeves with detected inside diameter (ID) flaw indications will be removed from service upon detection. The requirement to remove SG tubes containing electrosleeves in two operating cycles was incorporated in TS 5.5.9 in Amendment No. 132 issued May 21, 1999. The first Electrosleeve tube was installed in the fall of 1999 and the two-cycle allowance will expire in the fall of 2002.

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

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

The proposed change would remove the restriction that requires all steam generator tubes repaired with Electrosleeves to be removed from service at the end of two operating cycles following installation of the first Electrosleeve. This would allow all steam generator tubes repaired with Electrosleeves to remain in service. Reference 2 [licensee's letter dated October 27, 1998] concluded that there was no significant increase in the probability or consequences of an accident previously evaluated when using the Electrosleeve repair method. The two operating cycle restriction was invoked because the NRC staff concluded that the UT [ultrasonic] methods used to perform NDE [nondestructive examination] for inservice inspections of the Electrosleeved tubes could not reliably depth size stress corrosion cracks to ensure that structural limits are maintained.

Revision 4 to topical report BAW-10219P [nonproprietary version is attached to the application] has addressed the concerns that resulted in the restriction of two operating cycles and consequently, the probability of an accident previously evaluated is not significantly increased. As a result, the consequences of any accident previously evaluated are not affected.

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

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing plant operation. Reference 2 concluded that the use of the Electrosleeve repair method did not create the possibility of a new or different kind of accident from any accident previously evaluated when using this method to repair steam generator tubes. This proposed change removes the two operating cycle limit for the Electrosleeved tubes based on the evaluations and justifications of the NDE techniques used to perform inservice examinations of the Electrosleeved steam generator tubes provided in Revision 4 of the topical report.

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

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

The proposed change does not affect the acceptance criteria for an analyzed event. The margin of safety presently provided by the structural integrity of the steam generator tubes remains unchanged. Reference 2 concluded that the use of the Electrosleeve repair method did not involve a significant reduction in a margin of safety when using this method to repair steam generator tubes. The proposed change removes the two operating cycle limit based on the evaluations and justifications presented in Revision 4 of the topical report.

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

The reference to “Reference 2” in the criteria above is a reference to the licensee's letter dated October 27, 1998, and the no significant hazards consideration (NHSC) in that letter, which was published in the Federal Register (63 FR 66604) on December 2, 1998. This NHSC is applicable to the current application letters because it applies to the use of Electrosleeved steam generator tubes, the subject of the current application letters.

The NRC staff published an earlier Notice of Consideration for the application dated February 15, 2001, in the Federal Register on March 21, 2001 (66 FR 15931).

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: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Stephen Dembek.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: February 26, 2002.

Description of amendment request: Revise the definition of Operable in Technical Specification (TS) 1.0.K with respect to support system requirements for AC power sources. Conforming changes are made to specific support system TSs in Sections 3/4.5, “Core and Containment Cooling Systems,” 3/4.7, “Station Containment Systems,” and 3/4.10, “Auxiliary Electrical Power Systems,” and associated Bases.

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

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

The revised definition of “Operable” redefines the AC power source requirements to allow either normal or emergency power available for equipment requiring AC power to be considered operable and provides conforming changes to specific supported system Technical Specifications. None of the proposed changes affects any parameters or conditions that could contribute to the initiation of any accident. The proposed change does not affect the ability of the AC power sources to perform their required safety functions nor does the proposed change affect the ability of the systems requiring AC power to perform their respective safety functions. As a result, the ability of these systems to mitigate accident consequences is unchanged. As such, these changes do not impact initiators of analyzed events, nor the analyzed mitigation of design basis accident or transient events.

More stringent requirements for the inoperable AC power source action provisions that ensure availability of all TS required systems, subsystems, trains, components, and devices and the purely administrative changes do not affect the initiation of any event, nor do they negatively impact the mitigation of any event.

The elimination of some explicit requirements to verify the operability of remaining equipment (i.e., to verify which TS action is required to be entered and taken) does not affect the initiation of any event, nor does it negatively impact the mitigation of any event.

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

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

The proposed changes do not involve any physical modification to the plant, change in Technical Specification setpoints, change in plant design basis, or a change in the manner in which the plant is operated. No new or different type of equipment will be installed. No safety-related equipment or safety functions are altered as a result of these changes. In addition, there are no changes in methods governing normal plant operation. No new accident modes are created since plant operation is unchanged. None of the proposed changes affects any parameters or conditions that could contribute to the initiation of any accident. The changes do not introduce any new accident or malfunction mechanism that could create a new or different kind of accident, thus, no new failure mode is created. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The manner in which plant systems relied upon in the safety analyses to provide plant protection is not changed. Plant safety margins continue to be maintained through the limitations established in the Technical Specifications Limiting Conditions for Operation and Actions. These changes do not impact plant equipment design or operation, and there are no changes being made to safety limits or safety system settings that would adversely affect the ability of the plant to respond as assumed in the accident analyses as a result of the proposed changes. Since the changes have no effect on any safety analysis assumptions or initial conditions, the margins of safety in the safety analyses are maintained.

In addition, administrative changes that do not change technical requirements or meaning, and the imposition of more stringent requirements to ensure operability, have no negative impact on margins of safety.

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

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: March 19, 2002.

Description of amendment request: The proposed Technical Specification changes involve the removal of the existing scram function and Group 1 isolation valve closure functions of the Main Steam Line Radiation Monitors (MSLRM). An explicit requirement for periodic functional test and calibration of the MSLRM is added to maintain operability of the mechanical vacuum pump trip function.

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

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

The scram and Group 1 isolation functions of the MSLRMs do not serve as initiators for any of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR). The MSLRM scram function is not credited in the UFSAR, and the Group 1 isolation trip function of the MSLRMs was only assumed in one design-basis event which was the control rod drop accident. Because these functions are not initiators of accidents, their removal does not increase the probability of occurrence of previously evaluated accidents.

There is no accident analysis that relies on the high radiation scram of the reactor protection system and its removal has no impact on the consequences of accidents previously evaluated. The results of the control rod drop accident analysis remain within approved guidelines.

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

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

The proposed changes to the plant involve limited changes to protective circuitry, but do not involve any plant hardware changes that could introduce any new failure modes. The changes will not affect non-MSLRM scram and isolation functions. In addition, the MSLRMs will remain active for other trip/isolation functions, and these monitors will still alarm in the control room to alert operators to off-normal conditions. The reconstituted design-basis control rod drop accident analysis does not rely upon the trip functions that are being eliminated.

Therefore, the removal of the Group 1 isolation valve closure and scram functions of the MSLRMs does not create the possibility of a new or different kind of accident than those previously evaluated.

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

The proposed change involves the elimination of the scram and Group I isolation signal from the MSLRMs. Operation under the proposed change will not change any plant operation parameters, nor any protective system setpoints other than removal of these functions. The effects of the control rod drop accident without the MSLRM scram and isolation signal results in doses which remain well within 10 CFR Part 100, “Reactor Site Criteria,” limits. The proposed changes will reduce the chances of unnecessary plant trips occurring as a result of an inadvertent MSLRM scram or Group I isolation.

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

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendments request: March 28, 2002.

Description of amendments request: This requested amendment would permit Virginia Electric and Power Company (VEPCO) to replace the existing Westinghouse fuel with Framatome ANP Advanced Mark-BW fuel at North Anna Power Station, Units 1 and 2. The accompanying requested exemptions from 10 CFR 50.44 and 10 CFR 50.46 will be processed separately.

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 probability of occurrence or the consequences of an accident previously evaluated is not significantly increased. The Advanced Mark-BW fuel is very similar in design to the Westinghouse fuel that is being replaced in the core. The reload core designs for North Anna cycle will meet all applicable design criteria. [VEPCO] will use the NRC-approved standard reload design models and methods to demonstrate that all applicable design criteria and all pertinent licensing basis criteria will be met. Evaluations will be performed as part of the cycle specific reload safety analysis to confirm that the existing safety analyses remain applicable for operation of the Framatome Advanced Mark-BW fuel.

Operation of the Advanced Mark-BW fuel will not result in a measurable impact on normal operating plant releases, and will not increase the predicted radiological consequences of accidents postulated in the UFSAR [Updated Final Safety Analysis Report]. Therefore, neither the probability of occurrence nor the consequences of any accident previously evaluated is significantly increased.

2. The possibility for a new or different type of accident from any accident previously evaluated is not created. The Framatome Advanced Mark-BW fuel is very similar in design (both mechanical and composition of materials) to the resident Westinghouse fuel. The North Anna core in which the fuel operates will be designed to meet all applicable design criteria and ensure that all pertinent licensing basis criteria are met. Demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of accident. North Anna safety analyses have demonstrated in Section 6.0 of [the March 28, 2002 submittal] that the use of Advanced Mark-BW fuel is acceptable. All design and performance criteria will continue to be met and no new single failure mechanisms will be created. The use of the Advanced Mark-BW fuel does not involve any alteration to plant equipment or procedures which would introduce any new or unique operational modes or accident precursors. Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created.

3. The margin of safety is not significantly reduced. The operation of Advanced Mark-BW fuel does not change the performance requirements on any system or component such that any design criteria will be exceeded. The normal limits on core operation defined in the North Anna Technical Specifications will remain applicable for the use of Advanced Mark-BW fuel. The reload core designs for the cycles in which the Advanced Mark-BW fuel will operate will specifically evaluate any pertinent differences between the Advanced Mark-BW fuel product and the current Westinghouse fuel product, including both the mechanical design differences and the past irradiation history. The use of Advanced Mark-BW fuel will be specifically evaluated during the reload design process using [VEPCO's] reload design models and methods approved by the NRC. North Anna safety analyses have demonstrated in Section 6.0 of [the March 28, 2002 submittal] that the use of Advanced Mark-BW fuel is acceptable. Therefore, the margin of safety as defined in the Bases to the North Anna Units 1 and 2 Technical Specifications is not significantly reduced.

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

Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Section Chief: John A. Nakoski.

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 [e-mail].

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of application for amendment: January 31, 2002, as supplemented on March 27, 2002.

Brief description of amendments: The amendment allows a one-time 5-year extension, for a total of 15 years, for the performance of the next Unit 1 integrated leak rate test (ILRT). The amendment also exempts Unit 1 from the requirement to perform a post-modification containment ILRT associated with the steam generator replacement.

Date of issuance: May 1, 2002.

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

Amendment No.: 252.

Renewed Facility Operating License No. DPR-53: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7413). The March 27, 2002, supplemental letter provided clarifying information that did not change the scope of the original notice or the initial proposed no significant hazards consideration. The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated May 1, 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: May 25, 2001, as supplemented by letter dated January 24, 2002.

Brief description of amendments: The amendments eliminated response time testing requirements for selected sensors and specified instrumentation loops for the Engineered Safety Features and the Reactor Trip System.

Date of issuance: April 22, 2002.

Effective date: As of the date of issuance and shall be impl

emented within 30 days from the date of issuance.

Amendment Nos.: 197, 190.

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

Date of initial notice in Federal Register: December 12, 2001 (66 FR 64290). The supplement dated January 24, 2002, provided clarifying information that did not change the scope of the May 25, 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 April 22, 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: December 20, 2001, as supplemented by letters dated February 14, and March 26, 2002.

Brief description of amendments: The amendments revised the Technical Specifications to incorporate NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-51, “Revise containment requirements during handling irradiated fuel and core alterations,” Revision 2. The amendments selectively adopted the Alternate Source Term specifically for a fuel handling accident and a weir gate drop accident at Catawba Nuclear Station, Units 1 and 2.

Date of issuance: April 23, 2002.

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

Amendment Nos.: 198/191.

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

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7415). The supplements dated February 14, and March 26, 2002, provided clarifying information that did not change the scope of the December 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 April 23, 2002.

No significant hazards consideration comments received: No.

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

Date of application of amendments: December 28, 2000, as supplemented by letters dated February 15, April 26, June 26, and October 31, 2001, and March 4, 2002.

Brief description of amendments: The amendments revised the Technical Specifications related to controls to ensure acceptable margins of subcriticality in the spent fuel pools to account for Boraflex degradation.

Date of Issuance: April 22, 2002.

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

Amendment Nos.: 323, 323, 324.

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

Date of initial notice in Federal Register: February 7, 2001 (66 FR 9382). The supplements dated February 15, April 26, June 26, and October 31, 2001, and March 4, 2002, provided clarifying information that did not change the scope of the December 28, 2000, 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 April 22, 2002.

No significant hazards consideration comments received: No.

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

Date of application of amendments: December 20, 2001.

Brief description of amendments: The amendments revise the Technical Specifications (TS) to eliminate the use of the term “unreviewed safety question,” and replace the word “involve” with the word “require” as it applies to changes made to the updated Final Safety Analysis Report and the TS Bases.

Date of Issuance: April 22, 2002.

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

Amendment Nos.: 324, 325.

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

Date of initial notice in Federal Register: January 22, 2002 (67 FR 2923). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 22, 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: March 22, 2002, as supplemented by letter dated March 28, 2002.

Brief description of amendment: The amendment modifies Technical Specification Surveillance Requirement (SR) 3.6.1.3.6 to add a footnote specifying that the isolation time of each main steam isolation valve (MSIV) include circuit response time and valve motion time until the next outage greater than 72 hours.

Date of issuance: April 25, 2002.

Effective date: April 25, 2002.

Amendment No.: 175.

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

Public comments requested as to proposed no significant hazards consideration: Yes (67 FR 16767 dated April 8, 2002). The notice provided an opportunity to submit comments on the Commission's proposed no significant hazards consideration determination. No comments have been received. The notice also provided for an opportunity to request a hearing by May 8, 2002, but indicated that if the Commission makes a final no significant hazards consideration determination any such hearing would take place after issuance of the amendment. The Commission's related evaluation of the amendment, finding of exigent circumstances, consultation with the State of Washington and final determination of no significant hazards consideration are contained in a Safety Evaluation dated April 25, 2002.

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

NRC Section Chief: Stephen Dembek.

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

Date of application for amendment: September 7, 2001 as revised December 17, 2001.

Brief description of amendment: The amendment revised the Post Accident Monitoring Instrumentation Technical Specifications to ensure that licensee commitments to Regulatory Guide 1.97 are properly reflected.

Date of issuance: April 25, 2002.

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

Amendment No.: 211.

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

Date of initial notice in Federal Register: February 5, 2002 (67 FR 5328). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 25, 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: September 21, 2001, as supplemented by letter dated January 31, 2002.

Brief description of amendments: The amendments revise the reactor core safety limit for peak fuel centerline temperature from less than or equal to 4700 °F (i.e., the current technical specifications limit) to the design-basis fuel centerline melt temperature of less than 5080 °F, for unirradiated fuel, decreasing by 58 °F per 10,000 Megawatt-Days per MetricTonne Uranium (MWD/MTU) burnup. Additionally, the licensee is allowed to irradiate four ZIRLO clad rods to 69,000 MWD/MTU that are currently in Byron Unit 2 reactor. The staff denied a portion of the amendment request regarding extending burnup limit up to 75,000 MWD/MTU for future lead test assembly (LTA) campaigns. A separate Notice of Partial Denial of Amendment to Facility Operating License and Opportunity for Hearing has been published in the Federal Register.

Date of issuance: April 19, 2002.

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

Amendment Nos.: 127 and 122.

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: November 28, 2001 (66 FR 59505). The supplemental letter dated January 31, 2002, contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: November 30, 2001.

Brief description of amendments: The amendments 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 is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less,” to, “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”

Date of issuance: April 19, 2002.

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

Amendment Nos.: 205 and 201.

Facility Operating License Nos. DPR-29 and DPR-30: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7417). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 19, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 25, 2002.

Brief description of amendments: These amendments revised the Technical Specifications requirement for pressure testing diesel fuel oil system piping. The elevated pressure test will be replaced by a test at normal system operating conditions in accordance with the inservice inspection program.

Date of Issuance: April 23, 2002.

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

Amendment Nos.: 181 and 124.

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

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7419). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 23, 2002.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: November 16, 2001, as supplemented March 12, 2002.

Brief description of amendment: The amendment revises TS Table 3.3-4, “Engineered Safety Feature Actuation System Instrumentation Trip Setpoints.” The changes are required as part of a planned design change to replace the existing 4kV offsite power transformers, loss of voltage relays, and degraded voltage relays with components of an improved design to increase the reliability of offsite power for safety-related equipment.

Date of issuance: April 19, 2002.

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

Amendment No.: 268.

Facility Operating License No. DPR-58: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: December 12, 2001, (66 FR 64298). 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 April 19, 2002.

No significant hazards consideration comments received: No.

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 request: April 9, 2002, as supplemented April 25, 2002.

Description of amendment request: The amendment revises Technical Specification Surveillance Requirement 4.8.2.3.c.1 for the Train AB and CD batteries. The amendment modifies the requirement to verify that the Train AB and CD battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deterioration. The amendment allows batteries exhibiting damage or deterioration to be determined operable by an evaluation. The amendment is consistent with an NRC-approved change to the Standard Technical Specifications for Westinghouse plants (NUREG 1431, Revision 1), as documented in Technical Specification Task Force Standard Technical Specification Change Traveler-38, “Revise visual surveillance of batteries to specify inspection is for performance degradation.”

Date of issuance: April 26, 2002.

Effective date: As of the date of issuance, to be implemented immediately.

Amendment No.: 249.

Facility Operating License No. DPR-74: Amendment revise the technical specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. April 25, 2002 (67 FR 20552).

The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated April 26, 2002.

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: April 1, 2002, as supplemented by letters dated April 10 and April 15, 2002.

Brief description of amendment: This amendment adds an exception to the technical specifications to perform the surveillance test of Table 3-2, Item 20 (Recirculation Actuation Logic Channel Functional Test) under administrative controls while components in excess of those allowed by Conditions a, b, d, and e of TS 2.3(2) are maintained operable by dedicated operator action and are required to be returned to operable status within one hour. This exception will apply only to the remainder of Cycle 20 and the entirety of Cycle 21.

Date of issuance: April 19, 2002.

Effective date: April 19, 2002, to be implemented within 30 days from the date of issuance.

Amendment No.: 206.

Facility Operating License No. DPR-40: Amendment revised the Technical Specifications. Public comments requested as to proposed no significant hazards consideration: Yes (67 FR 16130 dated April 4, 2002). The notice provided an opportunity to submit comments on the Commission's proposed no significant hazards consideration determination. No comments have been received. The notice also provided for an opportunity to request a hearing by May 6, 2002, but indicated that if the Commission makes a final no significant hazards consideration determination any such hearing would take place after issuance of the amendment. The Commission's related evaluation of the amendment, finding of exigent circumstances, consultation with the State of Nebraska and final determination of no significant hazards consideration are contained in a Safety Evaluation dated April 19, 2002.

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

NRC Section Chief: Stephen Dembek.

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

Date of amendment request: December 14, 2001, as supplemented by letter dated February 13, 2002.

Brief description of amendment: The amendment deletes technical specification (TS) Figures 2-1A (Reactor Coolant System (RCS)—Temperature Limits for Heatup) and 2-1B (RCS Pressure—Temperature Limits for Cooldown) and replaces them with a single Figure 2-1. Additionally, the amendment changes the lowest service temperature from 182 ° F to 164 ° F to be in compliance with Reference 4, American Society of Mechanical Engineers (ASME) Section III, NB-2332 and the basis for the minimum boltup temperature to be in compliance with Reference 5, ASME Section XI, Appendix G. The Bases for TS 2.1 is being updated to reflect the use of ASME Code Case N-640 and the Westinghouse Electric Company/Combustion Engineering (W/CE) pressure temperature (P-T) limit curve methodology as applicable. Finally, based on the replacement of Figures 2-1A and 2-1B with a single Figure 2-1, the following TS are changed: 2.1.1(8), 2.1.2(1), 2.1.2(2), 2.1.2(6), 2.1.2(6)(a), 2.1.2(6)(c), 2.1.2(6)(d), and 2.1.6(4) as they reference the deleted curves.

Date of issuance: April 22, 2002.

Effective date: April 22, 2002, to be implemented within 30 days from the date of issuance.

Amendment No.: 207.

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

Date of initial notice in Federal Register: January 22, 2002 (67 FR 2928). The February 13, 2002, supplemental letter 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 April 22, 2002.

No significant hazards consideration comments received: No.

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

Date of amendment request: November 21, 2001, as supplemented by letter dated February 13, 2002.

Brief description of amendment: The amendment reformats and revises Technical Specifications (TSs) 2.15(5) and (6), “Instrumentation and Control Systems.” The new TSs clarify the scope of the alternate shutdown panels (ASPs). The change resulted from a corrective action needed to address the regulatory requirements for the ASPs and the associated auxiliary feedwater panel, as documented in Licensee Event Report 97-002, Revision 0, dated May 14, 1997.

Date of issuance: April 25, 2002.

Effective date: April 25, 2002, to be implemented within 60 days from the date of issuance.

Amendment No.: 208.

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

Date of initial notice in Federal Register: December 26, 2001 (66 FR 66470). The February 13, 2002, supplemental letter 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 April 25, 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, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: November 16, 2001.

Brief description of amendments: The amendments revised Technical Specification Section 5.5.16, “Containment Leakage Rate Testing Program,” to allow a one-time extension of the 10 CFR Part 50, Appendix J, Type A integrated leak rate test interval from the required 10 years to a test interval of 15 years.

Date of issuance: April 22, 2002.

Effective date: April 22, 2002, to be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-150; Unit 2-150.

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

Date of initial notice in Federal Register: January 8, 2002 (67 FR 930). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 22, 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, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: September 13, 2001, and supplemental letter dated March 14, 2002.

Brief description of amendments: The amendments revise TS Section 5.5.9, “Steam Generator Tube Surveillance Program,” to allow the extension of the steam generator tube W star (W*) alternate repair criteria (ARC) through Cycles 12 and 13. This extension will allow the licensee additional time to validate the W* leak rate model through performance of additional in-situ pressure testing of W* indications.

Date of issuance: April 29, 2002.

Effective date: April 29, 2002, to be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-151; Unit 2-151.

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

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55021). The March 14, 2002, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 29, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 13, 2002.

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

Date of issuance: April 23, 2002.

Effective date: April 23, 2002, to be implemented within 60 days of issuance.

Amendment Nos.: Unit 2-186; Unit 3-177.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 19, 2002 (67 FR 12605). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 23, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 21, 2001, as supplemented by letters dated October 24, 2001 and March 14, 2002.

Brief description of amendments: The amendments revise TS 5.5.2.12, “Ventilation Filter Testing Program.” Specifically, the reference to the American Society of Mechanical Engineers (ASME) Code N510-1989 was changed to the American National Standards Institute Standard N510-1975. This change was requested to ensure the clarity of the methodology used to test the Control Room Emergency Air Cleanup System and Post-Accident Cleanup Filter System High Efficiency Particulate Air (HEPA) filters. Although the test methodology is slightly different than that in N510-1989, the acceptance criteria are the same. Also, in Subsection 5.5.2.12.d the references to Regulatory Guide (RG) 1.52, Revision 2, and ASME N510-1989 were deleted. This section is concerned with pressure drop testing across HEPA filters.

Date of issuance: April 30, 2002.

Effective date: April 30, 2002, to be implemented within 30 days of issuance.

Amendment Nos.: Unit 2—187; Unit 3—178.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7421). The March 14, 2002, supplemental letter 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 no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 30, 2002.

No significant hazards consideration comments received: No.

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

Date of amendments request: June 5, 2001.

Brief Description of amendments: The amendments revise Technical Specifications (TS) Surveillance Requirement 3.4.14.1 to clarify the frequency of performance with regard to Reactor Coolant System Pressure Isolation Valves in the Residual Heat Removal System flow path. Also, related TS Bases and editorial changes are part of this TS change.

Date of issuance: April 22, 2002.

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

Amendment Nos.: 155/147.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55025). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 22, 2002.

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 19, 2001, as supplemented by letter dated March 11, 2002.

Brief description of amendments: The amendments revised the Technical Specifications to state that a representative sample of reactor instrumentation excess flow check valves (EFCVs) will be tested every 18 months such that each EFCV will be tested at least once every 10 years. Prior to issuance of these amendments; the EFCVs were required to be tested every 18 months.

Date of issuance: April 11, 2002.

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

Amendment Nos.: 230/171.

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

Date of initial notice in Federal Register: November 4, 2001 (66 FR 57125). The supplement dated March 11, 2002, provided clarifying information that did not change the scope of the September 19, 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 April 11, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 27, 2001, as supplemented by letter dated January 23, 2002.

Brief description of amendments: The amendments revise the frequency for Surveillance Requirement (SR) 3.8.1.13 from once every 18 months (with a maximum of 22.5 months including the 25% grace period of SR 3.0.2) to once every 24 months (for a maximum of 30 months including the 25% grace period of SR 3.0.2). The change allows this SR to be performed following the diesel generator inspection/maintenance, which is performed at a 24-month interval in accordance with the manufacturer's recommendations.

Date of issuance: April 22, 2002.

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

Amendment Nos.: 126, 104.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 25, 2001 (66 FR 38767). The supplement dated January 23, 2002, provided clarifying information and reduced the scope of the June 27, 2001, application, but did not change the initial proposed no significant hazards consideration determination for this approval.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 22, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 17, 2001 (TS-366).

Brief description of amendments: The amendments removed the low-scram pilot air header pressure switches.

Date of issuance: April 8, 2002.

Effective date: As of date of issuance, to be implemented within 120 days following completion of the Unit 2 Cycle 12 refueling outage scheduled for the spring 2003, and the Unit 3 Cycle 10 refueling outage scheduled for the spring 2002.

Amendment Nos.: 276 and 235.

Facility Operating License Nos. DPR-52 and DPR-68: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 14, 2001 (66 FR 57126). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 8, 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 amendment: November 15, 2001, as supplemented March 11, 2002.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) and the facility operating licenses (FOLs) to reflect an increase in the authorized maximum steady-state core power levels at the Sequoyah Nuclear Plant, Units 1 and 2, from 3411 megawatts thermal (MWt) to 3455 MWt, an increase of approximately 1.3 percent.

Date of issuance: April 30, 2002.

Effective date: As of the date of issuance and shall be implemented within 45 days for Unit 1 and 120 days for Region 2.

Amendment Nos.: 275 and 264.

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

Date of initial notice in Federal Register: December 12, 2001 (66 FR 64303). The supplemental letter 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 a Safety Evaluation dated April 30, 2002.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 7th day of May 2002.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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