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

I. Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and 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 October 1, 2004 through October 14, 2004. The last biweekly notice was published on October 12, 2004 (69 FR 60677).

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. Within 60 days after the date of publication of this notice, 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.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

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

Within 60 days after the date of publication of this notice, 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 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include 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/requestor to relief. A petitioner/requestor who fails to satisfy 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.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

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

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

Date of amendment request: April 30, 2004.

Description of amendment request: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of title 10 of the Code of Federal Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, several notes or specific exceptions are revised to reflect the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated April 30, 2004.

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

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

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

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

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

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

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

The NRC staff 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. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60666.

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: December 9, 2003.

Description of amendment request: The proposed amendment would revise Technical Specification 3.7.1, “Main Steam Safety Valves (MSSVs),” to increase the maximum allowable lift setting on two MSSVs on each unit. In addition, the proposed amendment would increase the completion time for reducing the Power Level-High Trip setpoint.

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

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

This license amendment request proposes to increase the upper range of the relief setting of the first two Main Steam Safety Valves (MSSVs) by 10 psi [pounds per square-inch]. The MSSVs are not accident initiators. They are credited with relieving secondary system pressure and act as a heat sink for the Reactor Coolant System (RCS) when the preferred heat sink is not available. Increasing the upper end of the setpoint for the first two MSSVs to lift does not affect the steam relieving capacity of the total or any combination of MSSVs that lift. This proposed amendment does not install any new components or change the physical characteristics of the MSSVs. Therefore, the change does not involve a significant increase in the probability of an evaluated accident.

The Updated Final Safety Analysis Report Chapter 14 safety analyses were reviewed considering the change to the upper end of the lift settings range of the first two MSSVs. The analyses show that increasing the upper end of the lift setting range does not exceed the pressure limits of the reactor coolant or main steam systems, nor the radiological consequences anticipated by the safety analyses. Therefore, the change will not involve a significant increase in the consequences of an evaluated accident.

This proposed amendment will also increase the Technical Specification Completion Time to reset the Power Level-High Trip from 12 hours to 36 hours. The purpose of the Power Level-High Trip is to trip the reactor if reactor power exceeds a set value, and is required by Technical Specifications to be reset according to the number of MSSVs remaining operable. The trip is not an accident initiator but is a signal that responds to an accident condition. Therefore, the change does not involve a significant increase in the probability of an evaluated accident.

Reducing the setpoint of the Power Level-High Trip within the time allotted by Technical Specifications provides additional assurance that the MSSVs will be able to perform their design function by keeping the reactor power within the ability of the MSSVs to relieve steam volume. There is low probability of a transient that could result in steam generator overpressure during the proposed 36 hours to reset the Power Level-High Trip. Therefore, this change does not involve a significant increase in the consequences of an evaluated accident.

Therefore, this proposed license amendment does not significantly increase the probability or consequences of an accident previously evaluated.

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

The proposed amendment will increase the upper end of the lift pressure for the first two MSSVs and increase the Technical Specification Completion Time to reset the Power Level-High Trip setpoint.

The proposed amendment does not involve a physical alteration of the plant or change the plant configuration. It does not require any new or unusual operator actions. The amendment does not alter the way any structure, system, or component functions and does not alter the manner in which the plant is operated. It does not introduce any new failure modes.

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

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

The margin of safety in this case is that the MSSVs release sufficient steam to relieve pressure in the secondary system and to act as a heat sink to prevent over-pressurization of the RCS when the preferred heat sink is not available. Increasing the upper end of the setpoint for the first two MSSVs to lift does not affect the steam relieving capacity of the total or any combination of MSSVs that lift. Potential delay in the opening of the first two MSSVs does not result in exceeding the pressure limits of the reactor coolant or main steam systems.

Reducing the Power Level-High Trip setpoint within the specified time limit provides additional assurance that the MSSVs will be able to perform their design function by keeping the reactor power within the ability of the MSSVs to relieve steam volume. A completion time of 36 hours to lower the Power Level-High Trip setpoint is based on a reasonable time to correct the MSSV inoperability, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

Therefore, this proposed license amendment does not involve a significant reduction in [a] margin of safety.

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

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

NRC Section Chief: Richard J. Laufer.

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

Date of amendments request: August 16, 2004.

Description of amendments request: The proposed change adds topical report NEDE-32906P-A, “TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses,” to the documents listed in Technical Specification (TS) 5.6.5 describing the approved methodologies used to determine the core operating limits. Unit 2 will be unable to resume power operation following Refueling Outage 16 without NRC approval for inclusion of the TRACG methodology in TS 5.6.5.b.

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:

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

Response: No.

The proposed change to TS 5.6.5.b will add General Electric Nuclear Energy topical report NEDE-32906P-A, “TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses,” to the list of documents describing approved methodologies for determining core operating limits. NRC review and acceptance of the TRACG methodology is documented in an October 22, 2001, letter and associated safety evaluation issued to General Electric Nuclear Energy (i.e., refer to ADAMS Accession Numbers ML012740390 and ML012740161). Analyzed events are assumed to be initiated by the failure of plant structures, systems, or components. The core operating limits, which are developed using the topical report being added, ensure that the integrity of the fuel will be maintained during normal operations and that design requirements will continue to be met. The proposed change does not involve physical changes to any plant structure, system, or component. Therefore, the probability of occurrence for a previously analyzed accident is not significantly increased.

The consequences of a previously analyzed accident are dependent on the initial conditions assumed for the analysis, the behavior of the fuel during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. Use of the analytical methodologies described in the topical report being added to TS 5.6.5.b will ensure that applicable design and safety analyses acceptance criteria are met. Use of these NRC-approved methodologies does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. As a result, no analysis assumptions are violated and there are no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident. Use of the approved methodologies described in the topical report being added to TS 5.6.5.b ensures that plant structures, systems, or components are maintained consistent with the safety analysis and licensing bases. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.

Therefore, the proposed change adding General Electric Nuclear Energy licensing topical report NEDE-32906P-A to the TS 5.6.5.b list of documents describing approved methodologies for determining core operating limits 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 adding licensing topical report NEDE-32906P-A to TS 5.6.5.b, and the use of the analytical methods described therein, does not involve any physical alteration of plant systems, structures, or components, other than allowing for fuel and core designs in accordance with NRC approved methodologies. The proposed methodology continues to meet applicable criteria for core operating limit analysis. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result no new failure modes are being introduced.

Therefore, the proposed change adding General Electric Nuclear Energy licensing topical report NEDE-32906P-A to the TS 5.6.5.b list of documents describing approved methodologies for determining core operating limits does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The margin of safety is established through the design of the plant structures, systems, and components, through the parameters within which the plant is operated, through the establishment of the setpoints for the actuation of equipment relied upon to respond to an event, and through margins contained within the safety analyses. The proposed change adding General Electric Nuclear Energy licensing topical report NEDE-32906P-A to the TS 5.6.5.b list of documents describing approved methodologies for determining core operating limits does not impact the condition or performance of structures, systems, setpoints, and components relied upon for accident mitigation. The proposed change does not significantly impact any safety analysis assumptions or results. Therefore, the proposed change adding topical report NEDE-32906P-A to the TS 5.6.5.b list of documents describing approved methodologies for determining core operating limits does 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: Steven R. Carr, Associate General Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief (Acting): Michael L. Marshall.

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

Date of amendment request: September 22, 2004.

Description of amendment request: The proposed change will revise Columbia Generating Station's licensing basis by replacing the current plant-specific reactor pressure vessel (RPV) material surveillance program with the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Specifically, the proposed amendment would revise Columbia's Final Safety Analysis Report (FSAR) to include participation in the ISP as described in the program document BWRVIP-86-A, “BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP) Implementation Plan,” dated October 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 implements an ISP program that meets the requirements of 10 CFR 50, Appendix H, Paragraph III.C, “Requirements for an Integrated Surveillance Program.” The proposed ISP program ensures the same level of RPV integrity as Columbia's current material surveillance program. Implementation of the proposed ISP is not a precursor or initiator of any previously evaluated accident. No physical changes to Columbia Generating Station are involved with the proposed change. The proposed change will not cause the RPV or interfacing systems to be operated outside of any design limit or testing limit, and will not alter any assumptions or initial conditions previously used in evaluating the radiological consequences of an accident.

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 revises the licensing basis for Columbia Generating Station to reflect participation in the BWRVIP ISP. The NRC has approved the ISP as an acceptable material surveillance program pursuant to 10 CFR 50, Appendix H, paragraph III.C. No physical changes to the plant are associated with the proposed change. No changes in design or operation of any system, structure, or component will be made as a result of the proposed change. The ISP is an alternative monitoring program and cannot create a new failure mode or a new or different kind of accident from any previously evaluated.

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.

Compliance with RPV material surveillance program requirements specified in 10 CFR 50, Appendix H and the fracture toughness requirements contained in 10 CFR 50, Appendix G ensure an adequate margin of safety exists in the fracture toughness of RPV beltline ferritic materials during any condition of normal operation, anticipated operational occurrence, and system hydrostatic tests. Implementation of the proposed ISP has been evaluated to meet the requirements of 10 CFR 50, Appendix H and this margin of safety is not impacted. Compliance with the requirements of 10 CFR 50, Appendix G will not be affected by this proposed change.

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

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: September 27, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.”

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated September 27, 2004.

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 eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed 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 Accident Previously Evaluated

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

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve significant hazards consideration.

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: September 30, 2004.

Description of amendment request: The proposed license amendment request would change the technical specifications and the Final Safety Analysis Report to revise the Columbia Generating Station's licensing and design bases to reflect the application of the alternative source term (AST) methodology with an exception. That exception is TID-14844, “Calculation of Distance Factors for Power and Test Reactor Sites,” which will continue to be used as the radiation dose basis for equipment qualification, and radiation zone maps/shielding calculations.

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

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

The alternative source term does not affect the design or operation of the facility in a manner that would impact the probability of an accident previously evaluated. Assumed performance requirements of the system structures and components are within existing design capability. The manner in which the systems are required to operate has not changed.

Once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences. The implementation of the alternative source term methodology has been evaluated in revisions to the analyses of the following limiting design basis accidents at Columbia Generating Station:

• Control Rod Drop Accident
• Fuel Handling Accident
• Main Steam Line Break Accident
• Loss of Coolant Accident

This amendment request includes changes to the Technical Specifications based on assumptions in the accident analyses. The results of these analyses demonstrate that, with the requested changes, the dose consequences of these limiting events are within the regulatory limits provided by the NRC for use with the alternative source term.

A new license and design basis analysis on secondary containment drawdown is provided to resolve a Justification for Continued Operation. The consequences, based on alternative source term methodology, remain within regulatory limits. This change to the licensing and design basis does not result in a significant increase in consequences.

Alternative source term methodology has been applied to resolve the Unresolved Safety Question on control room unfiltered air inleakage. The accident analyses results show, with the increased unfiltered air inleakage, the control room operator doses remain within regulatory limits.

Therefore, approval of the proposed amendment request 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 requested changes are based on accident analyses. System[,] structure and component performance assumptions included in the accident analyses result in doses within regulatory limits. Use of these performance assumptions does not:

• Require the installation of any new equipment,
• Require the modification of any existing equipment,
• Change the manner in which the equipment is required to be operated,
• Assume equipment performance outside existing design capabilities, or
• Require new operator actions.

Therefore Energy Northwest application of the alternative source term methodology does not create any new accident initiators or precursors of a new or different kind of accident.

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

The changes proposed are associated with the implementation of a new licensing basis for Columbia Generating Station. Approval of a basis change from the original source term developed in accordance with TID-14844 to a new alternative source term as described in RG [Regulatory Guide] 1.183 is requested. The results of the accident analyses revised in support of this submittal, and the requested Technical Specification changes, are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies.

Safety margins and analytical conservatisms have been evaluated and are satisfied. The analyzed accidents have been carefully selected and margin has been retained to ensure that the analyses adequately bound postulated event scenarios. The dose consequences of these limiting design basis accidents are within the acceptance criteria found in the applicable regulatory requirements and guidance. These requirements and guidance are presented in 10 CFR 50, App. A, 10 CFR 50.67, GDC [General Design Criterion] 19, and RG 1.183.

The proposed changes can be made while still satisfying regulatory requirements and review criteria, with margin. The changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Therefore, operation of Columbia Generating Station in accordance with the requested amendment does not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: April 14, 2004.

Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) to allow a one-time interval extension of no more than five years for the Type A, Integrated Leakage Rate Test (ILRT) of the primary containment. The proposed amendment would also correct the TSs to remove a reference to an obsolete alphanumeric identifier in TS 4.7.A.2.a, and reformat existing text on TS Pages 3/4.7-4 and 3/4.7-5 to improve consistency in its presentation.

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 Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below.

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

The proposed changes do not modify the design or operation of the containment. Therefore, the proposed changes, therefore, will not increase the probability of accidents previously evaluated.

The proposed extension to Type A, ILRT testing does not involve a significant increase in the consequences of an accident. Research documented in NUREG-1493 has found that Type A tests identify only a few potential containment leakage paths that also cannot be identified by Type B and C tests. The leaks that have been found by Type A tests have only been marginally above existing requirements. The NUREG then concluded that reducing the Type A testing frequency to once every 20 years was found to lead to an imperceptible increase in risk. These generic conclusions were confirmed by a plant-specific risk analysis performed using the current Pilgrim individual plant examination (IPE) internal events model that concluded the radiological consequences are low to negligible, and remain below regulatory limits. Therefore, any potential change in the radiological consequences is not considered significant.

The proposed correction to remove the alphanumeric identifier (i.e., definition 1.U), which is no longer used in the TSs, from the statements regarding the applicability of surveillance frequency to leak rate tests is editorial in nature. Likewise, the proposed formatting changes to existing information to improve its presentation are also editorial in nature. Since these changes are administrative in nature, they cannot increase the probability or consequences of previously analyzed accidents.

Therefore, since the radiological consequences are below the regulatory limits and the probability of an accident is unchanged, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

There are no new plant operation modes or physical modifications being proposed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.

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

The proposed revisions to the TSs add a one-time, 5-year extension to the current interval of 10 years from the last Type A test. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public. The NUREG also found that, generically, the design containment leakage rate contributes about 0.1 percent to the individual risk, and that the decrease in Type A testing frequency would have a minimal affect on this risk since 95 percent of the potential leakage paths are detected by Type C testing. This was further confirmed by a plant-specific risk assessment using the current Pilgrim IPE internal events model. Therefore, by meeting applicatory regulatory limits, any potential decrease in margin of safety would not be considered significant.

The proposed correction to remove the alphanumeric identifier (i.e., definition 1.U), which is no longer used in Pilgrim TSs, from the statements regarding the applicability of surveillance frequency to leak rate tests is editorial in nature. Likewise, the proposed formatting changes to existing information to improve its presentation are also editorial in nature. As these changes are administrative in nature, the proposed changes do 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: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

NRC Section Chief: Daniel Collins, Acting.

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

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

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

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

Date of amendment request: April 30, 2004.

Description of amendment request: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, several notes or specific exceptions are revised to reflect the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated April 30, 2004.

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

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

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

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

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

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

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

The NRC staff 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. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Gene Suh.

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

Date of amendment request: May 20, 2004.

Description of amendment request: The proposed changes would modify the Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TSs) to support activation of the trip outputs of the previously-installed Oscillation Power Range Monitor (OPRM) portion of the Power Range Neutron Monitoring (PRNM) system. Specifically, the proposed changes would revise LGS TS 2.2.1, “Reactor Protection System Instrumentation Setpoints,” TS 3/4.3.1, “Reactor Protection System Instrumentation,” TS 3/4.3.6, “Control Rod Block Instrumentation,” TS 3/4.4.1, “Recirculation System” and their associated Bases.

The proposed changes would also revise TS 6.9.1, “Routine Reports,” and delete interim corrective action requirements from the Recirculation System TS.

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

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

Response: No. This modification has no impact on any of the previously installed PRNM functions. Plant operation in portions of the former restricted zone may potentially cause a marginal increase in the probability of occurrence of an instability event. This potential increase in probability is acceptable because the OPRM Upscale Function will automatically detect the condition and initiate a reactor scram before the Minimum Critical Power Ratio (MCPR) Safety Limit is reached. Consequences of the potential instability event are reduced because of the more reliable automatic detection and suppression of an instability event, and elimination of dependence on the manual operator actions.

The change to align the operability requirements for the Intermediate Range Monitor (IRM) rod block function with those for the corresponding IRM Reactor Protection System (RPS) functions affects only the rod block function. The justification for the change to IRM RPS function (done with the original PRNM modification) concluded that the RPS change would not increase the probability of occurrence of an accident previously evaluated; therefore, changing the associated rod block to align with those requirements would not do so either.

Therefore, the proposed changes do not involve a significant increase in the probability of 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 modification replaces procedural actions that were established to avoid operating conditions where reactor instabilities might occur with an NRC approved automatic detect and suppress function.

Potential failures in the OPRM Upscale Function could result in either failure to take the required mitigating action or an unintended reactor scram. These are the same potential effects of failure of the operator to take the appropriate action under the current procedural directions. The net effect of the modification changes the method by which an instability event is detected and by which mitigating action is initiated, but does not change the type of stability event that could occur. The effects of failure of the OPRM equipment are limited to reduced or failed mitigation, but such failure cannot cause an instability event or other type of accident.

The change to align the operability requirements for the IRM rod block function with those for the corresponding IRM RPS functions affects only the rod block function. The justification for the change to IRM RPS function (done with the original PRNM modification) concluded that the RPS change could not create the possibility of a new type of accident; therefore, changing the associated rod block to align with those requirements would not do so either.

Therefore, since no radiological barrier will be challenged as a result of activating the OPRM Upscale Function, it is concluded that 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?

Response: No. The current safety analysis assumes that the existing procedural actions are adequate to prevent an instability event. As a result, there is currently no quantitative or qualitative assessment of an instability event with respect to its impact on the MCPR Safety Limit.

The OPRM Upscale function is being implemented to automate the detection (via direct measurement of neutron flux) and subsequent suppression (via scram) of an instability event prior to exceeding the MCPR Safety Limit. The OPRM Upscale function provides a trip output of the same type as currently used for the Average Power Range Monitor (APRM). Its failure modes and types are identical to those for the present APRM output. Currently, the MCPR Safety Limit is not challenged by an instability event since the event is “mitigated” by manual means via the procedural actions, which prevent plant operating conditions where an instability event is possible. In both methods of mitigation (manual and automated), the margin of safety associated with the MCPR Safety Limit is still maintained.

Therefore, based on the fact that the MCPR Safety Limit will still be enforced, implementation of the OPRM Upscale function in place of the existing manual actions does not reduce the margin of safety.

The IRM rod block function is not considered in any safety analysis. As a result, its failure will not affect the margin of safety.

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

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

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Daniel S. Collins, Acting.

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

Date of amendment request: June 1, 2004.

Description of amendment request: The proposed changes would relocate the operability and surveillance requirements for the reactor coolant system safety/relief valve position instrumentation from the Limerick Generating Station (LGS) Technical Specifications (TSs) to the LGS Technical Requirements Manual (TRM) and plant procedures. Specifically, the changes would relocate TSs 3.4.2.c, 4.4.2.1, and the associated footnotes to the TRM. Additionally, the “Safety/Relief Valve Position Indicators” instrumentation would be relocated from Tables 3.3.7.5-1 and 4.3.7.5-1 of TSs 3.3.7.5 and 4.3.7.5, respectively to the TRM.

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

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

Response: No. The failure of the safety/relief valve (SRV) position instrumentation is not assumed to be an initiator of any analyzed event in the [Updated Final Safety Analysis Report] UFSAR. The proposed changes do not alter the physical design of the SRVs or any other plant structure, system, or component. The changes would remove the [SRV] position indicator operability and surveillance requirements from the LGS [TSs], and incorporate requirements verbatim for this instrumentation into a licensee-controlled document under the control of 10 CFR 50.59.

The proposed changes conform to NRC regulatory guidance regarding the content of plant [TSs] as identified in regulation 10 CFR 50.36, and NRC publication NUREG-1433, “Standard Technical Specifications—General Electric Plants, BWR/4.”

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

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

Response: No. The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions, associated with the operation of the plant. Accordingly, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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

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

Response: No. This instrumentation is not needed for manual operator actions necessary for safety systems to accomplish their safety function for the design basis accident events. The instrumentation provides only alarm and SRV position indication, and does not provide an input to any automatic trip function. Several diverse means are available to monitor SRV position, and operability and surveillance requirements will be established in a licensee-controlled document to assure the reliability of SRV position monitoring capability. Changes to these requirements will be subject to the controls of regulation 10 CFR 50.59, providing the appropriate level of regulatory control.

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

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

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Daniel S. Collins, Acting.

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

Date of application for amendments: June 24, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to Operable status within 7 days. Other changes included in the application are addressed in a separate notice.

The NRC staff issued a notice of opportunity for comment in the Federal Register on February 24, 2003 (68 FR 8637), on possible amendments to revise the action for one or more SDV vent or drain lines with an inoperable valve, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 15, 2003 (68 FR 18294). The licensee affirmed the applicability of the model NSHC determination in its application dated June 24, 2004.

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

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

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

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

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

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

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

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

Attorney for Licensee: Thomas S. O'Neill, Associate and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Daniel Collins, Acting.

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

Date of amendment requests: September 21, 2004.

Description of amendment requests: The proposed amendments would extend the allowed outage times from 72 hours to 14 days for an inoperable emergency diesel generator, an inoperable component cooling water system loop, an inoperable essential service water system loop, or an inoperable alternate offsite power circuit (69 kilovolt circuit). The proposed amendments would also change formats of the affected technical specification pages to improve their appearance but not alter any requirements.

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

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

Response: No.

The proposed changes to the Technical Specifications (TS) will extend the allowed outage time (AOT) for a single inoperable emergency diesel generator (EDG), one inoperable component cooling water (CCW) or essential service water (ESW) loop, or an inoperable 69 kilovolt (kV) offsite circuit from the current limit of 72 hours to 14 days. An independent alternating current (AC) power source consisting of two supplemental diesel generators (SDGs) will be installed to support the extended AOTs for the EDGs and the CCW and ESW systems. The SDGs will supply power to required safe shutdown loads in the affected unit.

The EDGs are backup AC power sources designed to power safe shutdown systems in the event of a loss of offsite power. As such, the EDGs are not initiators for any accident previously evaluated. The CCW and ESW systems provide cooling water to safety-related components. This is a support function, and malfunctions of the CCW and ESW systems are not initiators of any accidents previously analyzed. The 69 kV circuit is an alternate offsite power supply that must be manually connected by the control room operators to provide power to safety-related buses upon loss of the preferred 34.5 kV offsite power source. As such, the 69 kV circuit is not an initiator for any accident previously evaluated. The AOT extension for an inoperable EDG, a CCW or ESW loop, or 69 kV circuit does not introduce any failure mechanisms that would initiate a previously analyzed accident. Therefore, the proposed change permitting extension of the AOTs for the EDG, ESW, CCW, and 69 kV systems do not result in a significant increase in the probability of a previously evaluated accident.

The potential effect of the proposed change on the consequences of a previously evaluated accident has been considered. There are two EDGs per unit, and only one EDG per unit is required to fulfill the onsite AC power system safety function. During the extended AOT, the redundant EDG will be available to provide AC power to safety-related components. There are two CCW loops per unit, and only one CCW loop per unit is required to fulfill the CCW system safety function. During the extended AOT, the redundant CCW loop will be available to provide cooling water to safety-related components. There are two ESW loops per unit, and only one ESW loop per unit is required to fulfill the ESW system safety function for the affected unit. During the extended AOT, the redundant ESW loop will be available to provide cooling water to the safety-related components. The 69 kV offsite circuit is the alternate offsite power source. Only one offsite power source is required to fulfill the offsite power system safety function. During an extended AOT, the preferred offsite source will be available. Thus, the systems affected by the proposed amendment will still be capable of performing the safety functions needed to mitigate the consequences of an accident as previously evaluated.

The format changes improve appearance, but do not affect any requirements.

Therefore, the proposed change will not involve a significant increase in the probability or consequences of any 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 consists of increasing the AOTs allowed by TS for the EDG, CCW, ESW, and 69 kV systems. Extending existing AOTs, does not result in operation of the plant in any new or different manner, nor does it create any new accident precursors. The format changes improve appearance, but do not affect any requirements.

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

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

Response: No.

The margins of safety are established through design parameters, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change does not adversely affect any design or operating parameter or any setpoint used in the deterministic accident analyses to establish the margin of safety. Probabilistic risk assessment methods were used to evaluate the risked-based margins of safety for the proposed change. The results of these evaluations indicated the proposed AOT extensions combined with installation of additional on-site electrical power supplies results in a net risk reduction. The format changes improve appearance, but do not affect any requirements.

Therefore, the proposed change will not create a significant reduction in a margin of safety.

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

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: September 7, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 5.9.1b, “Annual Occupational Exposure Report,” and TS 5.9.1c, “Monthly Operating Reports.”

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated September 7, 2004.

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 eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed 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 Accident Previously Evaluated

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

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve significant hazards consideration.

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

NRC Section Chief: Robert A. Gramm.

STP Nuclear Operating Company, Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

Date of amendment request: September 30, 2004.

Description of amendment request: The amendment would change Technical Specification 4.4.4.2 to expand the range of conditions under which quarterly testing of block valves for the pressurizer power operated relief valves would be unnecessary.

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 block valve for the pressurizer power operated relief valve is not a potential accident initiator. Therefore, not requiring a surveillance of the block valve while it is being used to isolate its associated power operated relief valve will not increase the probability of an accident previously evaluated. Not requiring the surveillance of the block valve may slightly reduce the probability of a loss of coolant accident from a stuck open power operated relief valve since it will eliminate the challenge to the power operated relief valve from the pressure transient that results from cycling the block valve.

If pressurizer spray is not available or is not effective, either one of the two pressurizer power operated relief valves may be manually actuated to depressurize the reactor coolant system to mitigate the consequences of a steam generator tube rupture. Not performing the surveillance on the block valve is not relevant to the primary system for depressurizing the reactor coolant system (pressurizer spray). The block valves have been demonstrated by operating experience to be reliable and are also subject to the motor-operated valve testing program. Consequently, the proposed change does not significantly reduce the confidence that the block valve can be opened to permit manual actuation of the power operated relief valve to depressurize the reactor coolant system to mitigate an accident. Therefore, the proposed change does 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 accident from any accident previously evaluated?

Response: No.

The proposed change only affects the performance of the surveillance test for the block valve and does not introduce any operating configurations not previously evaluated.

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

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

Response: No.

The proposed change to the surveillance requirement for the block valve for the pressurizer power operated relief valve does not affect the assumptions in any accident analyses. There are no changes in plant performance parameters associated with the proposed change to the surveillance requirement for the block valve. Therefore, STPNOC concludes the proposed changes do not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: Michael K. Webb, Acting.

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

Date of amendment request: September 27, 2004.

Description of amendment request: The proposed amendment would change the Technical Specifications (TS) to provide consistency between Surveillance Requirement (SR) 4.7.1.6 and TS 3.3.5.1 regarding atmospheric steam relief valve instrumentation controls. The proposed amendment would also correct editorial errors in TS 3.7.1.6.

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

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

Response: No.

The proposed change does not involve a significant increase in the probability or consequences of a previously evaluated accident. The first proposed change only clarifies when SR 4.7.1.6 for the automatic controls of the atmospheric steam relief valve is applicable. The applicability is already established in TS 3.3.5.1 and meets the safety analysis. The second proposed change is editorial.

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

Response: No.

The first proposed change does not create the possibility of a new or different accident from any previously evaluated. The proposed change only clarifies when SR 4.7.1.6 for the automatic controls of the atmospheric steam relief valve is applicable. The applicability is already established in TS 3.3.5.1 and meets the safety analysis. The second proposed change is editorial.

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

Response: No.

The first proposed change does not involve a significant reduction in the margin of safety. The proposed change only clarifies when SR 4.7.1.6 for the automatic controls of the atmospheric steam relief valve is applicable. The applicability is already established in TS 3.3.5.1 and meets the safety analysis. The second proposed change is editorial.

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

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

NRC Section Chief: Michael K. Webb, Acting.

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

Date of amendment request: September 30, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 6.9.1.2, “Occupational Radiation Exposure Report,” and TS 6.9.1.5, “Monthly Operating Reports.”

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated September 30, 2004.

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—Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

The proposed change eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—Does the Proposed Change 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, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3—Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve significant hazards consideration.

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

NRC Section Chief: Michael K. Webb, Acting.

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: September 9, 2004.

Brief description of amendments: The proposed change will revise the surveillance requirement (SR) 3.6.6.8 frequency of every 10 years. Instead, the proposed change to SR 3.6.6.8 will require verification that spray nozzles are unobstructed following maintenance that could result in nozzle blockage.

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

Response: No.

The Containment Spray System is not considered an initiator of any analyzed event. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that may initiate an analyzed event. The proposed change will not alter the operation or otherwise increase the failure probability of any plant equipment that can initiate an analyzed accident. This change does not affect the plant design. There is no increase in the likelihood of formation of significant corrosion products. Due to their location at the top of the containment, introduction of foreign material into the spray headers is unlikely. Foreign material introduced during maintenance activities would be the most likely source for obstruction, and verification following such maintenance would confirm the nozzles remain unobstructed.

Consequently, there is no significant increase in the probability of an accident previously evaluated.

The Containment Spray System is designed to address the consequences of a LOCA [loss-of-coolant accident]. The Containment Spray System is capable of performing its function effectively with the single failure of any active component in the system, any of its subsystems, or any of its support systems. A plugged nozzle would have negligible impact on the capability of the Containment Spray System to respond to a Loss of Coolant Accident.

Therefore, the consequences of an accident previously evaluated are not significantly affected by the proposed change.

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 change will not physically alter the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation.

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 system is not susceptible to corrosion-induced obstruction or obstruction from sources external to the system. Maintenance activities that could introduce foreign material into the system would require subsequent verification to ensure there is no nozzle blockage. The spray header nozzles are expected to remain unblocked and available in the event that the safety function is required. Therefore, the capacity of the system would remain unaffected.

Therefore the proposed change does not involve a 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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Mohan Thadani, Acting Chief.

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: September 10, 2004.

Brief description of amendments: The proposed amendment would delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports.”

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated September 10, 2004.

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:

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

The proposed change eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve significant hazards consideration.

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

NRC Section Chief: Michael Webb, Acting.

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

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

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

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

Date of amendment request: April 1, 2004.

Brief description of amendment request: The proposed amendment would allow entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, and Surveillance Requirement 3.0.4 revised to reflect the LCO 3.0.4 allowance.

Date of publication of individual notice in Federal Register: August 24, 2004 (69 FR 52037).

Expiration date of individual notice: October 23, 2004.

Florida Power and Light Company, et al., Docket Nos. 50-335, and 50-389, St. Lucie Plant, Unit No. 1, and Unit No. 2, St. Lucie County, Florida

Date of amendment request: November 21, 2003.

Description of amendment request: Revise Technical Specifications to eliminate certain pressure sensor response time testing requirements as discussed in the Combustion Engineering Owners Group Topical Report NPSD-1167, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements.”

Date of publication of individual notice in the Federal Register: September 28, 2004 (69 FR 57975).

Expiration date of individual notice: November 29, 2004.

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, Unit 2, Matagorda County, Texas

Date of amendment request: September 30, 2004.

Description of amendment request: The amendment changes TS 4.4.4.2 to expand the range of conditions under which quarterly testing of block valves for the pressurizer power operated relief valves would be unnecessary.

Date of publication of individual notice in Federal Register: October 6, 2004 (69 FR 59969).

Expiration date of individual notice: October 20, 2004 (Comment); December 6, 2004 (Hearing).

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of application for amendment: February 27, 2004, as supplemented by letter dated August 11, 2004.

Brief description of amendment: The amendment revised the Technical Specifications, relocating the average power range monitor flux scram setting and rod block setting from the to the Core Operating Limits Report.

Date of Issuance: October 4, 2004.

Effective date: October 4, 2004 and shall be implemented within 60 days of issuance.

Amendment No.: 248.

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

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19563).

The supplement dated August 11, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 19, 2004.

Brief description of amendment: The amendment revised the Technical Specifications, changing the surveillance requirements associated with control rod scram time testing. Specifically, the amendment modified the conditions under which scram time testing of control rods is required, and added a requirement to perform such testing on a defined portion of control rods at a specified frequency during the operating cycle.

Date of Issuance: October 4, 2004.

Effective date: October 4, 2004 and shall be implemented within 60 days of issuance.

Amendment No.: 249.

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

Date of initial notice in Federal Register: April 27, 2004 (69 FR 22878).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 5, 2004, as supplemented August 6, 2004.

Brief description of amendment: This amendment adds a reference to the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants in Technical Specification Surveillance Requirement 4.0.5.a for the snubbers.

Date of issuance: October 1, 2004.

Effective date: October 1, 2004.

Amendment No. 117.

Facility Operating License No. NPF-63. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34697). The August 6, 2004, supplement contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: February 13, 2003, as supplemented July 8, 2003, December 12, 2003, June 4, 2004, July 30, 2004, and September 16, 2004.

Brief description of amendment: The amendment approves the use of an alternative source term methodology in accordance with Title 10 of the Code of Federal Regulations, Part 50, Section 50.67, based on a reevaluation of the design-basis loss-of-coolant and fuel handling accidents. In addition to related design-basis changes, the amendment revises the Technical Specifications to (1) permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs), (2) increase the allowable secondary containment bypass leakage, (3) delete the MSIV leakage control system, and (4) increase the allowed secondary containment draw-down time.

Date of issuance: September 28, 2004.

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

Amendment No.: 160.

Facility Operating License No. NPF-43: Amendment revises the Technical Specifications and authorizes changes to the Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28847).

The July 8, 2003, December 12, 2003, June 4, 2004, July 30, 2004, and September 16, 2004, supplemental letters provided additional clarifying information that was within the scope of the original application and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 28, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 30, 2004.

Brief description of amendment: The amendment revises technical specification 3.3.6.2, “Secondary Containment Isolation Instrumentation, Condition C, to add the words “not met” to the end of the phrase, “Required Action and associated Completion Time.” The omission of the words “not met” was an oversight during the change to Improved Standard Technical Specifications, NUREG 1433.

Date of issuance: January 30, 2004.

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

Amendment No.: 161.

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

Date of initial notice in Federal Register: June 22, 2004, (69 FR 34698).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 7, 2004.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 4, 2004, as supplemented by letter dated June 16, 2004.

Brief description of amendment: The amendment revises the Technical Specification requirements by eliminating the requirements associated with hydrogen recombiners and hydrogen monitors. These changes support implementation of the revisions to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.44, “Standards for Combustible Gas Control System In Light-Water-Cooled Power Reactors.” A notice of availability of this TS improvement was published in the Federal Register on September 25, 2003 (68 FR 55416).

Date of issuance: October 4, 2004.

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

Amendment No.: 142.

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

Date of initial notice in Federal Register: May 11, 2004 (69 FR 26187). The supplement dated June 16, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: July 28, 2003, as supplemented on May 20, 2004.

Brief description of amendment: The amendment revised Technical Specification 5.5.6, “Primary Containment Leakage Rate Testing Program,” to allow a one-time extension of the interval between the Type A, integrated leakage rate tests, from 10 years to no more than 15 years. Therefore, the first Type A test performed after the March 7, 1995, test shall be performed no later than March 7, 2010.

Date of issuance: September 28, 2004.

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

Amendment No.: 279.

Facility Operating License No. DPR-59: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 27, 2004 (69 FR 44696).

The supplement dated May 20, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 28, 2004.

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: May 21, 2004.

Brief description of amendments: The amendments revise the technical specifications to add an additional reference as an acceptable method for determining the reactor pressure vessel pressure-temperature limits.

Date of issuance: October 4, 2004.

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

Amendment Nos.: 139/139, 132/132.

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: August 3, 2004.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 15, 2004, and supplemented on June 22, 2004.

Brief description of amendments: The amendments allow for a one-time deferral of the Dresden, Units 2 and 3, Appendix J, Type A, Integrated Leakage Rate Test (ILRT) to no later than February 27, 2011, and July 13, 2009, respectively.

Date of issuance: October 13, 2004.

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

Amendment Nos.: 210/202.

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

Date of initial notice in Federal Register: March 16, 2004.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 13, 2004.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; 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 3, 2003.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.4.1, “Recirculation Loops Operating,” by adding a limiting condition for operation requirement that the linear heat generation rate (LHGR) limits shall be modified for single recirculation loop operation as specified in the Core Operating Limits Report. The associated TS Bases are also revised to reflect the new LHGR limit requirement.

Date of issuance: October 4, 2004.

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

Amendment Nos.: 167, 153, 209, 201, 221, 216.

Facility Operating License Nos. NPF-11, NPF-18, DPR-19, DPR-25, DPR-29 and DPR-30: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 6, 2004 (69 FR 694).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 27, 2004, as supplemented May 27, 2004.

Brief description of amendment: The amendment revised TS 3.4.5 to allow a one-cycle use of Westinghouse leak-limiting Alloy 800 SG tube sleeves as an acceptable SG tube repair. Specifically, surveillance requirements 4.4.5.4.a.6 and 4.4.5.4.a.9 are revised to list the Westinghouse leak-limiting Alloy 800 sleeves as an acceptable SG tube sleeving method in addition to the currently approved Westinghouse laser welded sleeves and the former ABB Combustion Engineering tungsten inert gas welded sleeves.

Date of issuance: October 5, 2004.

Effective date: Within 60 days of the date of issuance and shall include the licensee commitments contained in the licensee letters of January 27 and May 27, 2004. The commitments shall remain in effect for the authorized period of sleeving with Westinghouse Alloy 800 tubes, i.e., Cycle 17.

Amendment No: 260.

Facility Operating License No. DPR-66: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12369).

The supplement dated May 27, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on March 16, 2004 (69 FR 12369).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 5, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 17, 2003.

Brief description of amendments: These amendments revised the action requirements in TS 3/4 6.3 to more clearly define the action requirements for inoperable containment isolation valves (CIVs). The amendments also allowed under administrative control, the intermittent unisolating of penetration flow paths which have previously been isolated per the action requirements. The amendments also allowed the use of check valves as an isolation device, and an increase in the allowed outage time to 72 hours for CIVs associated with closed systems inside containment. The amendments also deleted existing surveillance requirements (SRs) and provided new SRs similar to those in the Improved Standard Technical Specifications.

Date of issuance: October 5, 2004.

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

Amendment Nos.: 261 and 143.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 25, 2003 (68 FR 66136).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 5, 2004.

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: October 29, 2003.

Brief description of amendments: These amendments relocate specific pressure and flow values associated with the high pressure safety injection, low pressure safety injection, boric acid makeup, and containment spray pumps from the Technical Specification to the St. Lucie Units 1 and 2 Updated Final Safety Analysis Reports.

Date of Issuance: October 6, 2004.

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

Amendment Nos.: 194, 136.

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

Date of initial notice in Federal Register: January 6, 2004 (69 FR 697).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 6, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 6, 2004.

Brief description of amendment: The amendment revises technical specification 3.3.a.2.B, by extending the completion time from 1 hour to 24 hours for an accumulator that is inoperable for any reason other than failure to meet minimum boron concentration requirements.

Date of issuance: October 5, 2004.

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

Amendment No.: 178.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53111).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 5, 2004.

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 1, 2002, and its supplement dated April 2, 2004.

Brief description of amendments: The amendments (1) change the allowances for bypassing and tripping tested channels, (2) remove a surveillance requirement for reactor trip system (RTS) turbine trip-turbine stop valve closure, (3) revise the nominal trip setpoint for RTS turbine trip-turbine stop valve closure, (4) revise the allowable value and nominal trip setpoint for RTS interlock, (5) and remove and relocate the turbine trip function from engineered safety feature actuation system turbine trip and feedwater isolation to other licensee-controlled documents.

Date of issuance: September 24, 2004.

Effective date: September 24, 2004, and shall be implemented within 120 days from the date of issuance.

Amendment Nos.: Unit 1-173; Unit 2-175.

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

Date of initial notice in Federal Register: January 7, 2003 (68 FR 810).

The April 2, 2004, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 24, 2004.

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: May 29, 2003, as supplemented by letter dated December 23, 2003, and May 7, 2004.

Brief description of amendments: The amendments revise several surveillance requirements (SRs) in Technical Specification (TS) 3.8.1 on alternating current sources for plant operation. The revised SRs have notes deleted or modified to adopt in part Staff-approved TSTF-283, Revision 3, which will allow these revised SRs to be performed, or partially performed, in reactor modes that previously were not allowed by the TSs. The proposed changes to SRs 3.8.4.7 and 3.8.4.8 for direct current sources were withdrawn in the licensee's letter dated May 7, 2004.

Date of issuance: September 28, 2004.

Effective date: September 28, 2004, and shall be implemented within 60 days of the date of issuance including the incorporation of the changes to the Technical Specification Bases for Technical Specification 3.8.1 as described in the licensee's letters dated May 29 and December 23, 2003, and May 7, 2004, and the NRC safety evaluation attached to the amendments. This includes the revision of procedures to instruct operator action to be taken to manually reset the emergency diesel generator, as discussed in Section 4.3 of the licensee's May 29, 2003, letter.

Amendment Nos.: Unit 1-174; Unit 2-176.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40715).

The December 23, 2003, and May 7, 2004, supplemental letters provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 28, 2004.

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: October 22, 2003.

Brief description of amendments: The amendments revise Section 3.6.3 of the Diablo Canyon Power Plant Technical Specifications to extend the local leakage rate testing intervals for the containment purge and vent valves with resilient seals from 184 days to 24 months.

Date of issuance: October 6, 2004.

Effective date: October 6, 2004, and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: Unit 1-175; Unit 2-177.

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

Date of initial notice in Federal Register: November 25, 2003 (68 FR 66139).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 6, 2004.

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: December 19, 2003, and its supplement dated May 13, 2004.

Brief description of amendments: The amendments change Technical Specification 5.5.9, “Steam Generator (SG) Tube Surveillance Program,” to revise the wedge region exclusion zones for outside diameter stress corrosion cracking alternate repair criteria (ARC) at tube support plate (TSP) intersections and for primary water stress corrosion cracking ARC at dented TSP intersections. The new wedge region exclusion zones are based on new analyses of loss-of-coolant accident plus safe shutdown earthquake loads completed in 2003 using plant-specific accident loads.

Date of issuance: October 6, 2004.

Effective date: October 6, 2004, and shall be implemented within 60 days of issuance.

Amendment Nos.: Unit 1-176; Unit 2-178.

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

Date of initial notice in Federal Register: February 3, 2004 (69 FR 5205).

The May 13, 2004, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 6, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 17, 2003, as supplemented July 15, and August 23, 2004.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) to delete the primary containment isolation valves and instrumentation associated with the permanent removal of the reactor vessel head spray piping.

Date of issuance: October 5, 2004.

Effective date: As of the date of issuance, to be implemented prior to restart from the fall 2004 refueling outage.

Amendment No.: 152.

Facility Operating License No. NPF-57: This amendment revised the TSs.

Date of initial notice in Federal Register: January 20, 2004 (69 FR 2746). The supplements dated July 15, and August 23, 2004, contained clarifying information and did not change the staff's proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 5, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 24, 2003, as supplemented by letter dated June 29, 2004.

Brief description of amendment: The amendment revised the Surveillance Requirements (SRs) associated with reactor protection system instrumentation. Specifically, the amendment revised the SRs associated with the control rod block instrumentation, source range monitors, and power distribution limits by removing unnecessary testing requirements.

Date of issuance: October 13, 2004.

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

Amendment No.: 153.

Facility Operating License No. NPF-57: This amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68672). The June 29, 2004 letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 13, 2004.

No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: March 1, 2004.

Brief description of amendment: The amendment extends the completion time (CT) from 1 hour to 24 hours for Condition B of Technical Specification (TS) 3.5.1, which defines requirements for the emergency core cooling system accumulators. Condition B of TS 3.5.1 specifies a CT to restore an accumulator to operable status when it has been declared inoperable for a reason other than the boron concentration of the water in the accumulator not being within the required range.

Date of issuance: October 4, 2004.

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

Amendment No.: 86.

Renewed Facility Operating License No. DPR-18: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34706).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 4, 2004.

No significant hazards consideration comments received: No.

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

Date of amendments request: September 19, 2003, as supplemented by letters dated March 31, June 18, and August 6, 2004.

Brief Description of amendments: This amendment revised Technical Specifications (TS) Limiting Conditions for Operation 3.8.4, “DC Sources—Operating,” for the remainder of operating cycle 19. Specifically, the TS change increased the Completion Time for the 1B Auxiliary Building DC electrical power system inoperability due to an inoperable battery to allow for on-line replacement of individual cells.

Date of issuance: September 30, 2004.

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

Amendment Nos.: 164.

Facility Operating License Nos. NPF-2: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 12, 2003 (68 FR 64137).

The supplements dated March 31, June 18 and August 6, 2004, provided clarifying information that did not change the scope of the September 19, 2003, application, nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2004.

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: August 29, 2003, as supplemented by letters dated November 11, 2003, and May 5, June 10, August 5, August 25, and September 27, 2004.

Brief Description of amendments: The amendments revised Technical Specifications Limiting Condition of Operation 3.9.3, “Containment Penetrations.” The changes allow the equipment hatch to be open during core alterations and/or during movement of irradiated fuel assemblies within containment.

Date of issuance: September 30, 2004.

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

Amendment Nos.: 165 and 157.

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

Date of initial notice in Federal Register: November 12, 2003 (68 FR 64137).

The supplements dated November 11, 2003, and May 5, June 10, August 5, August 25, and September 27, 2004, provided clarifying information that did not change the scope of the August 29, 2003, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2004.

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: August 25, 2004, as supplemented by letter dated September 27, 2004.

Brief Description of amendments: The amendments address the control room habitability guidance of Regulatory Guide 1.196, “Control Room Habitability at Light-Water Nuclear Power Reactors,” by revising Limiting Condition for Operation 3.7.10, “Control Room Emergency Filtration/Pressurization System (CREFS)” and TS 5.5.11, “Ventilation Filter Testing Program. The amendments also add a new section, TS 5.5.18, “Control Room Integrity Program (CRIP).”

Date of issuance: September 30, 2004.

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

Amendment Nos.: 166, 158.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53095). The supplement dated September 27, 2004, provided clarifying information that did not change the scope of the August 25, 2004, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 8, 2004, as supplemented by letter dated September 24, 2004.

Brief description of amendment: The amendment revises requirements in the technical specifications to adopt the provisions of Industry/Technical Specification Task Force (TSTF) change TSTF-359, “Increase Flexibility in Mode Restraints.” The availability of TSTF-359 for adoption by licensees was announced in the Federal Register on April 4, 2003 (68 FR 16579).

Date of issuance: October 8, 2004.

Effective date: October 8, 2004, and shall be implemented within 90 days of the date of issuance.

Amendment No.: 164.

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

Date of initial notice in Federal Register: May 11, 2004 (69 FR 26194).

The additional information provided in the supplemental letter dated September 24, 2004, does not expand the scope of the application as noticed and does not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 8, 2004.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 9, 2004, as supplemented by the letter dated September 14, 2004.

Brief description of amendment: The amendment revises requirements in the technical specifications to adopt the provisions of Industry/Technical Specification Task Force (TSTF) change TSTF-359, “Increase Flexibility in Mode Restraints.”

Date of issuance: October 7, 2004.

Effective date: October 7, 2004, and shall be implemented within 90 days of the date of issuance.

Amendment No.: 155.

Facility Operating License No. NPF-42. The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12373).

The additional information provided in the supplemental letter dated September 14, 2004, does not expand the scope of the application as noticed and does not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 7, 2004.

No significant hazards consideration comments received: No.