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Notice

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

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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 January 19, 2007, to February 1, 2007. The last biweekly notice was published on January 30, 2007 (72 FR 4304).

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 Start Printed Page 6781following 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, Rulemaking, Directives and Editing 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 O1F21, 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: Start Printed Page 6782(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)7ndash;(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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 26, 2006.

Description of amendment request: The proposed change deletes reference to the containment fan cooler (CFC) condensate flow switch from Technical Specification (TS) 3.4.5.1, “Reactor Coolant System Leakage—Leakage Detection Instrumentation,” and to modify or delete associated actions.

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 Reactor Coolant System (RCS) leakage detection systems are passive monitoring systems therefore the proposed changes do not affect reactor operations or accident analyses and have no radiological consequences. The proposed change continues to require diverse methods of monitoring leakage. The gaseous radioactivity monitor, although not included in the TSs and the CFC condensate flow switches, which are proposed for removal from the TSs, will be maintained functional and available.

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

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

Response: No.

The proposed change introduces no new mode of plant operation or any plant modification. The RCS leakage detection instrumentation is used solely for monitoring purposes and is not part of plant control instruments or engineered safety feature actuation circuits. The change does not vary or affect any plant operating condition or parameter.

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

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

Response: No.

The proposed change does not modify any of the RCS leakage detection instrumentation. The proposed change continues to require diverse methods of monitoring leakage. In addition, although not required by TS, multiple means of diverse monitoring RCS leakage will remain functional and available.

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: Terence A. Burke, Associate General Council—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: David Terao.

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

Date of amendment request: January 18, 2007.

Description of amendment request: The proposed change will revise the description of Grand Gulf Nuclear Station Technical Specification 4.2.2, “Control Rod Assemblies,” to allow to the use of hafnium as an additional type of control material.

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 NRC has specifically approved the use of hafnium as neutron absorbing material for use in BWR [boiling-water reactor] control rod assemblies. The use of hafnium in control rods as a neutron absorber material does not significantly alter the neutronic or mechanical functional characteristics of the control rods. Control rod designs using hafnium have been successfully used in other BWRs. Since control rods that utilize hafnium have a longer lifetime, the probability of some accidents involving the handling, on-site storage, and shipping of irradiated rods will actually be reduced. The proposed change does not alter the required number of control rods nor does it affect any of the specifications related to the control rods (e.g., the shutdown margin and scram timing requirements are unaffected).

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

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

Response: No.

The application of a control rod design using hafnium as an absorber material does not produce any new mode of plant operation or alter the control rods in such a way as to affect their function or operability since the new control rods are designed to be compatible with the existing control rods.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Start Printed Page 6783

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

Response: No.

The proposed change does not significantly affect the neutronic or mechanical characteristics of the control rods since the hafnium containing controls rods are designed to be compatible with the existing design and reload licensing criteria; therefore, there is no significant change in the margin of safety. It does not change the required number of existing control rods. It does not affect the existing Technical Specifications related to control rods (e.g., required shutdown margin and scram time, etc.).

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: Terence A. Burke, Associate General Council—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: David Terao.

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

Date of amendment request: October 11, 2006.

Description of amendment request: The proposed amendment would modify the plant Improved Technical Specifications (ITSs) to implement a more conservative requirement in ITS 3.7.7, “Nuclear Services Closed Cycle Cooling Water (SW) System.” The current Action A allows the plant to operate for up to 72 hours before initiating a shutdown when one required SW heat exchanger is inoperable. The proposed revision will only allow operation to continue for 8 hours before initiating a shutdown when one required SW heat exchanger is inoperable.

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

The limiting design basis accident for CR-3 includes, as an assumption, adequate heat removal capability by the SW system. The amendment is being proposed to ensure the SW system performs its design basis function. Adequate heat removal is provided by three OPERABLE SW heat exchangers. The 8 hour completion time will reduce the window that the plant can operate with only two SW heat exchangers before a shutdown is required. The proposed change does not increase the probability of an accident previously evaluated since the amendment is not a modification to plant systems, nor a change to plant operation that could initiate an accident. Therefore, granting the LAR [license amendment request] does not involve a significant increase in the probability or consequences of an accident previously evaluated. The dose consequences of all design basis accidents are unchanged by this proposed amendment.

(2) Does not create the possibility of a new or different kind of accident from any accident previously evaluated?

The function of the SW system considered in the design basis is to remove process and operating heat from safety-related components during normal as well as transient conditions. The proposed amendment to limit the allowed ACTION Completion Time to 8 hours will ensure the function of the SW system is consistent with the design basis and will not result in changes to the design, physical configuration of the plant or the assumptions made in the safety analysis. The requirement does not change the function of the system nor its ability to perform its design function. No alteration to plant configuration or operation is proposed. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Does not involve a significant reduction in a margin of safety?

CR-3's design basis considers adequate heat removal by the SW system to cool the containment fan assembly cooling coils and fan motors, spent fuel pool, SW pump motors and other equipment which must function following an accident. This proposed amendment will not alter the current design basis. By limiting the allowed ACTION Completion Time to 8 hours, the proposed amendment to ITS 3.7.7 will limit the time the safety function of the SW system can be compromised. Therefore, the amendment does not result in a reduction of the margin of safety.

The NRC staff has reviewed the analysis provided for Florida Power Corporation 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief (Acting): Margaret H. Chernoff.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear Station, Unit 2, Dauphin County, Pennsylvania

Date of amendment request: December 13, 2006.

Description of amendment requests: The amendment application proposes to delete Technical Specification (TS) 6.8.1.3, which provides the requirement for submittal of the annual occupational radiation exposure report.

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

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? No

The proposed change eliminates the Technical Specification reporting requirement for occupational radiation exposure information, which is in excess to that required to be submitted by 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. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? No

The 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? No

This change is an administrative change to reporting requirements of occupational radiation exposure data and will not reduce a margin of safety because it has no effect on any safety analyses assumptions. Hence, this change is administrative in nature. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

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

NRC Branch Chief: Claudia Craig. Start Printed Page 6784

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

Date of amendment request: December 21, 2006.

Description of amendment request: The proposed amendment revises the licensing basis to reflect a revision to the spent fuel pool criticality analysis methodology and a new criticality analysis. In addition, associated changes are proposed to Technical Specifications 3.7.12, “Spent Fuel Storage,” and 4.3.1, “Criticality,” to reflect the results of the new criticality analysis.

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

Operation of the facility in accordance with the proposed amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated. The presence of soluble boron in the Spent Fuel Pool (SFP) water being used for criticality control does not increase the probability of a dropped fuel assembly accident within the pool. The handling of the fuel assemblies in the SFP has always been performed and will continue to be performed in borated water.

There is no increase in the probability of the accidental misloading of fuel assemblies into the SFP fuel storage racks when considering the presence of soluble boron in the pool water for criticality control. Fuel assembly placement will continue to be controlled pursuant to approved fuel handling procedures and in accordance with the spent fuel storage rack limitations specified in the Technical Specifications (TS). There is no increase in the consequences for an accidental misloading of fuel assemblies in the SFP fuel storage racks because the criticality analyses demonstrate that the pool will remain subcritical following an accidental misloading.

Soluble boron credit is used to provide margin to offset uncertainties, tolerances, and off-normal/accident conditions, and to provide subcritical margin such that the SFP keff [effective neutron multiplication constant] is maintained less than or equal to 0.95. The plant-specific criticality analysis results demonstrate that the spent fuel rack keff will remain<1.0 (at a 95/95 percent probability and confidence level) even with the SFP flooded with unborated water.

There is no increase in the probability of the loss of normal cooling to the SFP water when considering the presence of soluble boron in the pool water for subcriticality control since a high concentration of soluble boron has always been maintained in the SFP water.

A loss of normal cooling to the SFP water causes an increase in the temperature of the water passing through the stored fuel assemblies. This causes a decrease in water density, which would result in a net increase in reactivity when soluble boron is present in the water. However, the additional negative reactivity provided by the 2100 ppm [parts per million] boron concentration limit, above that provided by the concentration required (805 ppm) to maintain keff less than or equal to 0.95, will compensate for the increased reactivity which could result from a loss of SFP cooling event. Because adequate soluble boron will be maintained in the SFP water the consequences of a loss of normal cooling to the SFP will not be increased.

Therefore, the proposed changes do 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

Under the proposed amendment, no changes are being made to the fuel storage racks themselves, to any other systems, or to the physical structures of the Primary Auxiliary Building. Therefore, there are no changes proposed to the plant configuration, equipment design, or installed equipment.

Criticality accidents in the SFP are not new or different types of accidents. They have been analyzed in the FSAR [Final Safety Analysis Report] and in fuel storage criticality analysis reports associated with specific licensing amendments. The proposed new SFP storage limitations are consistent with the assumptions made in the new criticality analysis, and will not have any significant effect on normal SFP operations and maintenance, and do not create the possibility of a new or different kind of accident. Verifications will continue to be performed to ensure that the SFP loading configuration meets specified requirements.

The current TS includes a SFP boron concentration limit that conservatively bounds the boration assumption of the new criticality analysis. Since soluble boron has always been maintained in the SFP water, implementation of this requirement for SFP criticality control purposes has have no effect on normal pool operations and maintenance. Also, since soluble boron has always been present in the SFP, a dilution event has always been a possibility. The loss of substantial amounts of soluble boron from the SFP that could lead to keff exceeding 0.95 was evaluated as part of the analyses in support of this license amendment request. The evaluation demonstrates that a dilution of the SFP boron concentration from the minimum TS concentration of 2100 to 805 ppm is not credible.

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

(3) Does the proposed amendment result in a significant reduction in a margin of safety?

Response: No

The proposed Technical Specification changes providing the resulting spent fuel storage operation limits provide adequate safety margin to ensure that the stored fuel assembly array always remains subcritical. These limits are based on a plant-specific criticality analysis performed in accordance with the present Westinghouse spent fuel rack criticality analysis methodology which allows credit for soluble boron.

The criticality analysis takes credit for soluble boron to ensure that keff will be less than or equal to 0.95 under normal circumstances. While the criticality analysis used credit for soluble boron, storage configurations have been defined using 95/95 keff calculations to ensure that the spent fuel rack keff is less than unity (0.995) with no soluble boron. Soluble boron credit is used to provide safety margin to offset uncertainties, tolerances, and off-normal/accident conditions, and to provide subcritical margin such that the SFP keff is maintained less than or equal to 0.95.

The loss of substantial amounts of soluble boron from the SFP that could lead to keff exceeding 0.95 was evaluated as part of the analyses in support of this license amendment request. The evaluation demonstrates that a dilution of the SFP boron concentration from the minimum TS concentration of 2100 to 805 ppm is not credible. Also, the plant-specific criticality analysis results demonstrate that even if a complete dilution were to occur the spent fuel rack keff would remain <1.0 (at a 95/95 percent probability and confidence level) with the SFP flooded with unborated water. The plant-specific criticality analysis performed in accordance with the conservative analysis methodology of the Westinghouse licensing topical report demonstrates that the requirements of 10 CFR 50.68 and 10 CFR 50, Appendix A, General Design Criterion 62 will be satisfied. 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: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Acting Branch Chief: Patrick D. Milano.

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

Date of amendment request: December 29, 2006.

Description of amendment request: The proposed amendments would Start Printed Page 6785revise Technical Specification (TS) 5.5.8 to indicate that the Inservice Testing Program shall include testing frequencies applicable to the American Society of Mechanical Engineers Code for Operations and Maintenance (ASME OM Code), and to indicate that there may be some non-standard frequencies specified as 2 years or less in the Inservice Testing Program to which the provisions of Surveillance Requirement (SR) 3.0.2 are applicable. The proposed changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-479, Revision 0, “Changes to Reflect Revision of 10 CFR 50.55a,” and TSTF-497, Revision 0, “Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less.”

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 proposed changes revise TS 5.5.8, “lnservice Testing Program,” for consistency with 10 CFR 50.55a(f)(4) requirements regarding inservice testing of pumps and valves. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

2. [Do] the proposed change[s] create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

The proposed changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure. Therefore, this proposed change does not create the possibility of an accident of a different kind than previously evaluated.

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No

The proposed changes revise TS 5.5.8, “lnservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: David Terao.

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

Date of amendment requests: December 29, 2006.

Description of amendment requests: The proposed amendments will revise Technical Specification (TS) 5.5.16 for consistency with the requirements of 10 CFR 50.55a(g)(4) for components classified as Code Class CC. This regulation requires licensees to update their containment inservice inspection requirements in accordance with Subsections IWE and IWL of Section XI, Division I of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code as limited by 10 CFR 50.55a(b)(2)(vi) and modified by 10 CFR 50.55a(b)(2)(viii) and 10 CFR 50.55a(b)(2)(ix). This license amendment request is consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Traveler number TSTF-343, “Containment Structural Integrity.”

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 revises the Technical Specification (TS) administrative controls programs for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The proposed change affects the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The frequency of visual examinations of the concrete surfaces of the containment and the mode of operation during which those examinations are performed has no relationship to or adverse impact on the probability of any of the initiating events assumed in the accident analyses. The proposed change would allow visual examinations that are performed pursuant to NRC-approved ASME [Code,] Section XI requirements (except where relief has been granted by the NRC) to meet the intent of visual examinations required by Regulatory Guide 1.163, without requiring additional visual examinations pursuant to the Regulatory Guide. The intent of early detection of deterioration will continue to be met by the more rigorous requirements of the Code-required visual examinations. As such, the safety function of the containment as a fission product barrier is maintained.

The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. It does not involve the addition or removal of any equipment, or any design changes to the facility.

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 accident from any accident previously evaluated?

Response: No.

The proposed change revises the TS Administrative Controls programs for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The change affects the frequency of visual examinations that will be performed for the concrete surfaces of the containments. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or a malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.

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? Start Printed Page 6786

Response: No.

The proposed change revises the TS Administrative Controls programs for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The change affects the frequency of visual examinations that will be performed for the concrete surfaces of the containments. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The safety function of the containment as a fission product barrier will be maintained.

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

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

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

NRC Branch Chief: David Terao.

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

Date of amendment requests: December 29, 2006.

Description of amendment requests: The proposed amendments will revise Technical Specification (TS) 3.4.1, “RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” and TS 5.6.5, “CORE OPERATING LIMITS REPORT (COLR). This license amendment request proposes to relocate the RCS DNB parameters for pressurizer pressure and RCS average temperature to the COLR. This relocation is consistent with Technical Specification Task Force Traveler TSTF-339, Revision 2, “Relocate TS Parameters to COLR.” TS 5.6.5 is revised to add topical reports WCAP-8567-P-A, “Improved Thermal Design Procedure,” and WCAP-11596-P-A, “Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores,” by name and title only. These changes are consistent with TSTF-363, Revision 0, “Revise Topical Report References in ITS 5.6.5, COLR.”

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

1. [Do] the proposed change[s] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes are programmatic and administrative in nature, and do not physically alter safety-related systems or affect the way in which safety-related systems perform their functions. The proposed changes relocate cycle-specific parameters from Technical Specification (TS) 3.4.1 to the Core Operating Limits Report (COLR). This does not change plant design or affect system operating parameters. The proposed changes do not, by themselves, alter any of the parameters. Removal of the cycle-specific parameters from the TS does not eliminate existing requirements to comply with the parameters. Also, TS 5.6.5 is revised to add topical reports WCAP-8567-P-A, “Improved Thermal Design Procedure,” and WCAP-11596-P-A, “Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores,” as they are approved analytical methods for determining core operating limits.

Although relocation of the cycle-specific parameters to the COLR would allow revision of the affected parameters without prior NRC approval, there is no significant effect on the probability or consequences of an accident previously evaluated. Future changes to the COLR parameters could result in event consequences that are either slightly less or slightly more severe than the consequences for the same event using the present parameters. The differences would not be significant and would be bounded by the existing requirement of TS 5.6.5c to meet the applicable limits of the safety analyses.

The cycle-specific parameters being transferred from the TS to the COLR will continue to be controlled under existing programs and procedures. The Final Safety Analysis Report Update (FSARU) accident analyses will continue to be examined with respect to changes in the cycle-dependent parameters obtained using NRC reviewed and approved reload design methodologies to ensure that the transient evaluation of new reload designs are bounded by previously accepted analyses. This examination will continue to be performed pursuant to 10 CFR 50.59 requirements, ensuring that future reload designs use NRC-approved methodologies and do not involve more than a minimal increase in the probability or consequences of an accident previously evaluated in the FSARU.

The proposed changes do not allow for an increase in plant power levels, do not increase the production, and do not alter the flow path or method of disposal of radioactive waste or byproducts. Therefore, the proposed changes do not change the type or increase the amount of effluents released offsite.

The proposed changes to TS 5.6.5b to reference only the topical report number and title for five of the topical reports do not alter the analytical methods that have been previously reviewed and approved by the NRC. This method of referencing topical reports would allow the use of current topical reports to support limits in the COLR without having to submit a request for an amendment to the operating license. Implementation of revisions to these topical reports would still be reviewed in accordance with 10 CFR 50.59 and, where required, revisions would be submitted to the NRC for approval prior to implementation.

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

2. [Do] the proposed change[s] create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

The proposed changes that relocate cycle-specific parameters from the TS to the COLR, thus removing the requirement for prior NRC approval of revisions to those parameters, do not involve a physical change to the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. No changes are being made to the parameters within which the plant is operated, other than their relocation to the COLR. No protective or mitigative action setpoints are affected by the proposed changes. The proposed changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No change to procedures that ensure the plant remains within analyzed limits are being proposed, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.

Relocation of cycle-specific parameters does not influence, impact, or contribute in any way to the possibility of a new or different kind of accident. The relocated cycle-specific parameters will continue to be calculated using the NRC-reviewed and approved methodology. The proposed changes do not alter assumptions made in the safety analysis, and operation within the core operating limits will continue.

The proposed changes to reference only the topical report number and title do not alter the use of the analytical methods that have been previously reviewed and approved by the NRC. This method of referencing topical reports would allow the use of current topical reports to support limits in the COLR without having to submit a request for an amendment to the operating license. Implementation of revisions to topical reports would still be reviewed in accordance with 10 CFR 50.59 and, where required, would receive NRC review and approval.

The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is a clarification to provide a complete listing of approved analytical methods used for determining core operating limits.

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

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are Start Printed Page 6787initiated. The proposed changes do not physically alter safety-related systems, nor do they affect the way in which safety-related systems perform their functions. No protective or mitigative action setpoints are affected by the proposed changes. Therefore, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. As the proposed changes to relocate cycle-specific parameters to the COLR will not affect plant design or system operating parameters, there is no detrimental impact on any equipment design parameter, and the plant will continue to be operated within prescribed limits.

The development of cycle-specific parameters for future reload designs will continue to conform to NRC-reviewed and approved methodologies, and will be performed pursuant to 10 CFR 50.59 to assure that the plant operates within cycle-specific parameters.

The proposed changes to reference only the topical report number and title do not alter the use of the analytical methods used to determine core operating limits that have been reviewed and approved by the NRC. This method of referencing topical reports would allow the use of current NRC-approved topical reports to support limits in the COLR without having to submit a request for an amendment to the operating license. Implementation of revisions to topical reports would still be reviewed in accordance with 10 CFR 50.59 and, where required, receive NRC review and approval.

The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is a clarification to provide a complete listing of approved analytical methods used for determining core operating limits.

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

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

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

NRC Branch Chief: David Terao.

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

Date of amendment requests: January 11, 2007.

Description of amendment requests: The proposed amendments would revise the Technical Specifications (TSs) to support replacement of the steam generators (SGs) at Diablo Canyon Power Plant, Unit Nos. 1 and 2. Revisions are proposed to TS 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” TS 5.5.9, “Steam Generator (SG) Program,” and TS 5.6.10, “Steam Generator (SG) Tube Inspection Report.” The replacement SGs are to be installed during the Diablo Canyon Power Plant, Unit No. 2, 14th refueling outage (2R14), currently scheduled for February 2008, and the Unit No. 1, 15th refueling outage (1R15), currently scheduled for January 2009.

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

1. [Do] the proposed change[s] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The revised engineered safety feature actuation system (ESFAS) steam generator (SG) Water Level-High High feedwater isolation Nominal Trip Setpoint and Allowable Value have been determined using the existing setpoint methodology approved for Diablo Canyon Power Plant. The setpoint analysis for the replacement steam generators (RSGs) accounts for the setpoint uncertainties specific to the RSG design. The revised Feedwater Isolation SG Water Level-High High (P-14) Nominal Trip Setpoint and Allowable Value are applied using a conservative surveillance requirement methodology. The function of the ESFAS instrumentation is unchanged. The Feedwater Isolation SG Water Level-High High (P-14) ESFAS instrumentation will continue to function in a manner consistent with the plant design basis and satisfy all the requirements of the safety analyses.

The probability and consequences of accidents previously evaluated in the Final Safety Analysis Report (FSAR) Update are not adversely affected because the revised Feedwater Isolation SG Water Level-High High (P-14) Nominal Trip Setpoint and Allowable Value continue to assure a conservative plant response to high SG level, consistent with the safety analyses and licensing basis.

The proposed changes revise and clarify the surveillance requirements for ESFAS Function 5.b, Feedwater Isolation SG Water Level-High High (P-14). These changes ensure that this function will actuate as assumed in the safety analyses.

The proposed changes to TS 5.5.9 delete the alternate repair criteria (ARC) for the existing SGs, incorporate tube inspection periods applicable to Alloy 690 thermally treated tubes, and delete the TS 5.6.10 reporting requirements for ARC. The TS 5.5.9 SG structural integrity, accident induced leakage, and operational leakage performance criteria will continue to be met for the RSGs. Meeting the SG performance criteria provides reasonable assurance that the SG tubes will remain capable of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. Removal of the ARC for the existing SGs will ensure that all tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness will be plugged as required by TS 5.5.9.c. With the revised SG tube inspection period, the SGs will continue to meet the SG program defined by NEI [Nuclear Energy Institute] 97-06, “Steam Generator Program Guidelines,” which incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring.

Removal of the ARC will reduce the allowable accident induced leakage following a main steamline break accident. The proposed changes do not have any impact on the accident induced leakage assumed in the other design basis accidents. The changes do not have any impact on the allowable SG operational leakage, allowable reactor coolant system activity, or the allowable SG secondary activity.

The proposed changes will not affect the probability of any accident initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety-related equipment assumed to function during an accident. There will be no change to accident mitigation performance. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the FSAR Update.

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

2. [Do] the proposed change[s] create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

The proposed changes will not affect the normal method of plant operation or create new methods of plant operation related to the Feedwater Isolation SG Water Level-High High (P-14) ESFAS setpoints. The proposed changes to the Feedwater Isolation SG Water Level-High High (P-14) instrumentation surveillance requirements will provide assurance that the plant will operate within the limits assumed in the safety analyses. The assumptions made in the setpoint analyses for the Feedwater Isolation SG Water Level-High High (P-14) ESFAS instrument do not create any new accidents, accident initiators, or failure mechanisms.

The proposed changes, which delete the TS 5.5.9 ARC for the existing SGs, incorporate tube inspection periods for Alloy 690 thermally-treated tubes in TS 5.5.9, and delete the ARC reporting requirements in TS 5.6.10, will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The primary-to-secondary leakage that may be experienced during all plant conditions will be monitored to ensure it remains within current safety analysis assumptions. The proposed changes do not adversely affect the method of operation of the SGs or the primary or secondary coolant controls and do not impact other plant systems or components. Start Printed Page 6788

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

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The FSAR Update Excessive Heat Removal due to Feedwater System Malfunctions event credits the Feedwater Isolation SG Water Level-High High (P-14) ESFAS instrumentation. The safety analysis limit assumed for the Feedwater Isolation SG Water Level-High High (P-14) ESFAS instrumentation for this event has not changed for the safety analyses for the RSGs. None of the acceptance criteria for Excessive Heat Removal due to Feedwater System Malfunctions event are changed as a result of the revised Feedwater Isolation SG Water Level-High High (P-14) Nominal Trip Setpoint and Allowable Value. The instrument surveillance requirement changes for the Feedwater Isolation SG Water Level-High High (P-14) function ensure that the instrumentation will actuate as assumed in the safety analysis.

The safety function of the SGs is maintained by ensuring the integrity of the tubes. SG tube integrity is a function of the design, environment, and the physical condition of the SG tubes. The proposed changes, which delete the TS 5.5.9 ARCs for the existing SGs, incorporate tube inspection periods for Alloy 690 thermally treated tubes in TS 5.5.9, and delete the ARC reporting requirements in TS 5.6.10, do not adversely impact the SG tube design or operating environment. SG tube integrity will continue to be maintained by implementing the SG Program to manage SG tube inspection, assessment, and repair. The requirements established by the SG program are consistent with those in the applicable design codes and standards.

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 requests involve no significant hazards consideration.

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

NRC Branch Chief: David Terao.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power Plant (HBPP), Unit 3 Humboldt County, California

Date of amendment request: May 17, 2006.

Description of amendment request: The licensee has proposed to modify the Physical Security Plan (PSP) to allow leaving certain security posts temporarily under emergency conditions requiring personnel to evacuate occupied plant areas for their health and safety.

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

Response: No.

Allowing the security posts and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously maintained has no impact on the probability of an accident from occurring, especially acts of nature such as earthquakes and tsunamis.

The HBPP Defueled Safety Analysis Report, Appendix A, and NRC Safety Evaluation Report (SER), Section 10, dated April 29, 1987, evaluate various accidents at HBPP. Because all fuel has been removed from the reactor vessel and stored in the spent fuel pool, the majority of accidents analyzed pertain to events that could only affect spent fuel or the spent fuel pool. All accidents affecting spent fuel or the spent fuel pool do not require security personnel action to protect the public health and safety, or to maintain offsite radiological doses well within regulatory limits. In addition, NRC SER, Section 10.7, “Impact of Tsunami Flooding,” analyzes the impact of tsunami flooding. That analysis identifies a likely impact of the tsunami to be a release of the radwaste tank radionuclide contents to the bay and some damage to the reactor building. For both situations, no security personnel action is required to maintain offsite radiological doses well within regulatory limits.

Allowing the security posts and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously maintained temporarily, under emergency conditions, does not create problems that could increase the consequences of an accident. The primary function of the manning and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, is to monitor, detect and assess unauthorized intrusion into the protected area, and has nothing to do with the probability or consequences of plant accidents.

If security personnel evacuate PSP, Section 3.1.4 and Table 7-1, security posts during a tsunami, those security personnel will be able to return to the PSP, Section 3.1.4 and Table 7-1, security posts after the tsunami and assess damage or intrusion by observing alarms and/or physical conditions as well as resume implementation of security post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1. In addition, upon evacuation, security personnel notify offsite security backup personnel of the evacuation and the need for the offsite personnel to remotely monitor HBPP security system alarms. Conversely, if security personnel remain at the PSP, Section 3.1.4 and Table 7-1, security posts during a tsunami and become injured, those security personnel would be unable to assist in the resumption of implementation of security post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1. Therefore, not continually manning the PSP, Section 3.1.4 and Table 7-1, security posts during a tsunami does not increase the consequences of the tsunami.

(2) Does the change create the possibility of a new or different kind of accident from any accident evaluated?

Response: No.

As discussed in the response to Question 1 above, none of the analyzed accidents require security personnel action to keep offsite radiological doses well within regulatory limits. In addition, allowing security personnel to not continuously maintain security post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, after an emergency situation has occurred has no impact on the possibility of a new or different kind of accident from occurring. The primary function of the manning and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, is to monitor, detect, and assess unauthorized intrusion into the protected area, and has nothing to do with the possibility of a different kind of plant accident occurring.

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

Response: No.

NRC SER, Section 10.8, “Accident Analysis Conclusions,” summarizes the consequences from accidents in terms of offsite radiological doses. SER, Section 10.8, includes the statement, “The (NRC) staff has determined that offsite radiological consequences due to a tsunami are within acceptable dose guideline values.” As discussed in the response to Question 1 above, none of the analyzed accidents require security personnel action to keep offsite radiological doses well within regulatory limits. Therefore, allowing security personnel to not continuously maintain security post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, after an emergency situation has occurred has no impact on 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: Mr. Antonio Fernández, Esquire, Pacific Gas & Electric Company, Post Office Box 7442, San Francisco, CA 94120.

NRC Branch Chief: Claudia Craig.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power Plant (HBPP), Unit 3 Humboldt County, California

Date of amendment request: December 20, 2006.

Description of amendment request: The licensee has proposed to amend the Start Printed Page 6789Facility Operating License by deleting paragraph 2.B.3(c), and replacing it with a new paragraph 2.B.4 to read as follows: “Pursuant to the Act and Title 10, CFR, Chapter I, Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components.”

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

Response: No.

The proposed change eliminates a restriction regarding the type and limits of byproduct and special nuclear material to be received, possessed, and used onsite. However, in the proposed change, the type or amount of byproduct, source, or special nuclear material to be received, possessed, or used would not change plant systems or accident analysis, and as such, would not affect initiators of analyzed events or assumed mitigation of accidents. Therefore, the proposed change does not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change eliminates a restriction regarding the limits and type of byproduct and special nuclear material to be received, possessed, and used onsite. The proposed change does not involve a physical alteration to the plant or require existing equipment to be operated in a manner different from the present design. Temporary equipment brought onsite for decommissioning activities would still be required to be operated in accordance with plant procedures and licensing bases documents, regardless of the byproduct material content. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident evaluated.

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

Response: No.

The proposed change eliminates a restriction regarding the limit and type of byproduct and special nuclear material to be received, possessed, and used onsite. The proposed change has no effect on existing plant equipment, operating practices, or safety analysis assumptions. Temporary equipment brought onsite for decommissioning activities would still be required to be operated in accordance with plant procedures and licensing bases documents, regardless of the byproduct material content. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The U.S. Nuclear Regulatory Commission 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. Antonio Fernández, Esquire, Pacific Gas & Electric Company, Post Office Box 7442, San Francisco, CA 94120.

NRC Branch Chief: Claudia Craig.

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

Date of amendment request: November 15, 2006.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) Table 3.6.3-1, “Primary Containment Isolation Valves,” and relocate the information to the Technical Requirements Manual. The amendment would also revise other TS sections that reference TS Table 3.6.3-1. The proposed changes are based on the guidance in Generic Letter 91-08, “Removal of Component Lists from Technical Specifications.”

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

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

Response: No.

The proposed relocation of Technical Specification component lists of primary containment isolation valves does not alter the requirements for component operability or surveillance currently in the Technical Specifications. The proposed change to remove the component lists from TS and relocate the information to an administratively controlled document will have no impact on any safety related structures, systems or components.

The probability of occurrence of a previously evaluated accident is not increased because this change does not introduce any new potential accident initiating conditions. The consequences of accidents previously evaluated in the UFSAR [Updated Final Safety Analysis Report] are not affected because the ability of the components to perform their required function is not affected.

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

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

Response: No.

The proposed changes are administrative in nature, conform to the guidance in Generic Letter 91-08 and do not result in physical alterations or changes in the method by which any safety related system performs its intended function. The proposed changes do not affect any safety analysis assumptions. The proposed changes do not create any new accident initiators or involve an activity that could be an initiator of an accident of a different type.

All components will continue to be tested to the same requirements as defined in the Technical Specification Surveillance Requirements. The proposed revision does not make changes in any method of testing or how any safety related system performs its safety functions.

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

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

Response: No.

The proposed change to remove Technical Specification Table 3.6.3-1 from the Technical Specifications and relocate it to the Technical Requirements Manual does not alter the Technical Specification requirements for containment integrity and containment isolation and will not affect the containment isolation capability. Future revisions to the Technical Requirements Manual Table will be subject to evaluation pursuant to 10 CFR 50.59 [Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59].

The proposed change will not affect the current Technical Specification requirements or the components to which they apply.

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

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

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

NRC Branch Chief: Harold K. Chernoff.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco Nuclear Generating Station, Sacramento County, California

Date of amendment request: April 12, 2006, and supplemented November 21, 2006.

Description of amendment request: The licensee has proposed to amend its Start Printed Page 6790license to incorporate a new license condition addressing the license termination plan (LTP). This amendment will document the approval of the LTP, document the criteria for making changes to the LTP which will and will not require pre-approval by the NRC, and will document any conditions imposed with the approval of the LTP.

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 license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed change is administrative. The change allows for the approval of the LTP and provides the criteria for when changes to the LTP require prior U.S. Nuclear Regulatory Commission (NRC) approval. This change does not affect possible initiating events for accidents previously evaluated or alter the configuration or operation of the facility. Safety limits, limiting safety system settings, and limiting control systems are no longer applicable to Rancho Seco in the permanently defueled mode, and are therefore not relevant.

The proposed change does not affect the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and has no impact on plant operations. Therefore, the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. As described above, the proposed change is administrative and provides the criteria for when changes to the LTP require prior NRC approval. The safety analysis for the facility remains complete and accurate. There are no physical changes to the facility as a result of the proposed amendment and the plant conditions for which the design basis accidents have been evaluated are still valid.

The operating procedures and emergency procedures are not affected. The proposed changes do not affect the emergency planning zone, the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and have no impact on plant operations. Consequently, no new failure modes are introduced as the result of the proposed changes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. As described above, the proposed changes are administrative. There are no changes to the design or operation of the facility. The proposed changes do not affect the emergency planning zone, the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and have no impact on plant operations. Accordingly, neither the design basis nor the accident assumptions in the Defueled Safety Analysis Report, nor the Technical Specification Bases are affected. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's significant hazards 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: Arlen Orchard, Esq., General Counsel, Sacramento Municipal Utility District, 6201 S Street, P.O. Box 15830, Sacramento, CA 95817-1899.

NRC Branch Chief: Claudia M. Craig.

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

Date of amendment request: January 30, 2007.

Description of amendment request: The proposed amendment would revise the Farley Nuclear Plant, Units 1 and 2, Technical Specifications (TSs) to reflect a change to a site vice president organizational structure. The resulting structure places a vice president at the plant site. The proposed amendment describes changes in titles and administrative duties that accompany the reorganization.

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

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

The proposed change to [the] FNP TS involves SNC moving to a site vice president organizational structure. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. The proposed change also does not affect the operation, maintenance, or testing of the plant. Therefore, the response of the plant to previously analyzed accidents will not be affected. Consequently, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

As a result of the proposed change to the FNP TS, the qualification requirements for the unit staff position[s] will remain unchanged and the plant staff will continue to meet applicable regulatory requirements. Also, since no change is being made to the design, operation, maintenance, or testing of the plant, no new methods of operation or failure modes are introduced by the proposed change. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

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

The proposed change to the FNP TS will have no adverse impact on the onsite organizational features necessary to assure safe operation of the plant since the qualification requirements for the unit staff remains unchanged. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. Therefore, the proposed change does not involve a significant decrease 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: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: January 30, 2007.

Description of amendment request: The proposed amendments would revise the Hatch Nuclear Plant, Units 1 and 2, Technical Specifications (TSs) to reflect a change to a site vice president organizational structure. The resulting structure places a vice president at the plant site. The proposed amendment describes changes in titles and administrative duties that accompany the reorganization.

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:

Start Printed Page 6791

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

The proposed change to [the] HNP TS involves SNC moving to a site vice president organizational structure. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. The proposed change also does not affect the operation, maintenance, or testing of the plant. Therefore, the response of the plant to previously analyzed accidents will not be affected. Consequently, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

As a result of the proposed change to the HNP TS, the qualification requirements for the unit staff position[s] will remain unchanged and the plant staff will continue to meet applicable regulatory requirements. Also, since no change is being made to the design, operation, maintenance, or testing of the plant, no new methods of operation or failure modes are introduced by the proposed change. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

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

The proposed change to the HNP TS will have no adverse impact on the onsite organizational features necessary to assure safe operation of the plant since the qualification requirements for the unit staff remains unchanged. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. Therefore, the proposed change does not involve a significant decrease 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: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: January 30, 2007.

Description of amendment request: The proposed amendment would revise the Vogle Electric Generating Plant, Units 1 and 2, Technical Specifications (TSs) to reflect a change to a site vice president organizational structure. The resulting structure places a vice president at the plant site. The proposed amendment describes changes in titles and administrative duties that accompany the reorganization. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

The proposed change to [the] VEGP TS involves SNC moving to a site vice president organizational structure. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. The proposed change also does not affect the operation, maintenance, or testing of the plant. Therefore, the response of the plant to previously analyzed accidents will not be affected. Consequently, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

As a result of the proposed change to the VEGP TS, the qualification requirements for the unit staff position[s] will remain unchanged and the plant staff will continue to meet applicable regulatory requirements. Also, since no change is being made to the design, operation, maintenance, or testing of the plant, no new methods of operation or failure modes are introduced by the proposed change. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

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

The proposed change to the VEGP TS will have no adverse impact on the onsite organizational features necessary to assure safe operation of the plant since the qualification requirements for the unit staff remains unchanged. Since the proposed change is administrative in nature, it does not involve any physical changes to any structures, systems, or components, nor will their performance requirements be altered. Therefore, the proposed change does not involve a significant decrease 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: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: December 21, 2006 (TS-456).

Description of amendment request: The proposed amendment would revise Technical Specification (TS) Limiting Condition for Operation (LCO) 3.10.1 and the associated TS Bases to expand its scope to include provisions for temperature excursions greater than 212 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with inservice leak or hydrostatic testing, while considering operational conditions to be in Mode 4.

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 21, 2006 (71 FR 48561), on possible amendments to revise the plant-specific TS, to expand the scope of TS LCO 3.10.1, to include provisions for temperature excursions greater than 200 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in MODE 4, 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 October 27, 2006 (71 FR 63050). The licensee affirmed the applicability of the model NSHC determination in its application dated December 21, 2006.

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.

Technical Specifications currently allow for operation at greater than [200] °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not Start Printed Page 6792adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Technical Specifications currently allow for operation at greater than [200] °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different types of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. 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.

Technical Specifications currently allow for operation at greater than [200] °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: L. Raghavan.

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.

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

Date of amendment request: July 27, 2006, as supplemented by letters dated October 4 and October 9, 2006.

Brief description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.7.14, “Spent Fuel Pool Boron Concentration,” TS 3.7.15, “Spent Fuel Pool Storage,” and the associated Figure 3.7.15-1, and TS 4.3, “Fuel Storage,” and the associated Figure 4.3.1.2-1. In addition, this amendment would add TS 5.5.17, “Metamic Coupon Sampling Program,” and Surveillance Requirement 3.7.15.2 that directs the performance of the coupon sampling program. The proposed TS changes support a modification to the ANO-1 spent fuel pool (SFP) that would utilize Metamic® poison insert assemblies. In addition to the proposed plant modification, the licensee would increase the SFP boron concentration and credit boron to ensure that a 5-percent subcriticality margin is maintained during normal and accident conditions. This proposed amendment also would increase the allowable initial fuel assembly uranium-235 (U-235) enrichment from 4.1 weight percent (wt%) to a maximum U-235 enrichment of 4.95 wt%.

Date of publication of individual notice in Federal Register: December 26, 2006 (71 FR 77414).

Expiration date of individual notice: February 26, 2007.

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 (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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of application for amendment: January 26, 2006, as supplemented by letter dated December 20, 2006.

Brief description of amendment: The amendment revised the Millstone Power Station, Unit No. 2 Technical Specifications (TSs) to update the list of NRC-approved documents specified in the TSs that describe the analytical methods used to determine the core operating limits. The proposed change also corrects a typographical error in TS 5.3.1, “Reactor Core, Fuel Assembly,” which was introduced in the retyped pages provided to the NRC for issuance of Amendment No. 280, dated September, 25, 2003.

Date of issuance: January 23, 2007. Start Printed Page 6793

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

Amendment No.: 295.

Facility Operating License Nos. DPR-65: The Amendment revised the TSs.

Date of initial notice in Federal Register: May 9, 2006 (71 FR 26997). The supplement dated December 20, 2006, provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the Federal Register, and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 23, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 17, 2006.

Brief description of amendment: The amendment changed the Millstone Power Station, Unit No. 2, Technical Specifications by replacing the existing maximum and minimum pressurizer water volume and water level limits with a maximum water level limit. The associated TS bases were updated to address the proposed changes.

Date of issuance: January 30, 2007.

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

Amendment No.: 296.

Facility Operating License No. DPR-65: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 11, 2006 (71 FR 65141).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 30, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 27, 2006, as supplemented by letters dated October 4, October 9, and December 14, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.14, “Spent Fuel Pool Boron Concentration,” TS 3.7.15, “Spent Fuel Pool Storage,” and the associated Figure 3.7.15-1, and TS 4.3, “Fuel Storage,” and the associated Figure 4.3.1.2-1. In addition, this amendment added TS 5.5.17, “Metamic Coupon Sampling Program,” and Surveillance Requirement 3.7.15.2 that directs the performance of the coupon sampling program. The TS changes support a modification to the ANO-1 spent fuel pool (SFP) that utilize Metamic® poison insert assemblies. In addition to the proposed plant modification, the licensee increased the SFP boron concentration and credited boron to ensure that a 5-percent subcriticality margin is maintained during normal and accident conditions. This amendment also increased the allowable initial fuel assembly uranium-235 (U-235) enrichment from 4.1 weight percent (wt%) to a maximum U-235 enrichment of 4.95 wt%.

Date of issuance: January 26, 2007.

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

Amendment No.: 228.

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

Date of initial notice in Federal Register: December 26, 2006 (71 FR 77414). The supplement dated December 14, 2006, 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 January 26, 2007.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-374, LaSalle County Station, Unit 2, LaSalle County, Illinois

Date of application for amendments: April 21, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.13, “Primary Containment Leakage Testing Program,” to reflect a one-time extension of the LaSalle, Unit 2 primary containment Type A integrated leak rate test (ILRT) from the current requirement of no later than December 7, 2008, to prior to startup following the 12th LaSalle, Unit 2 refueling outage.

Date of issuance: January 24, 2007.

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

Amendment No.: 166.

Facility Operating License No. NPF-18: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: June 6, 2006 (71 FR 32605). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 24, 2007.

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

Date of application for amendments: April 13, 2005, as supplemented by letters dated December 22, 2005, June 12, 2006, and January 4, 2007.

Brief description of amendments: The proposed amendment would extend, on a one-time basis, the completion time (CT) for required action C.4, “Restore required Diesel Generators (DGs) OPERABLE status,” associated with Technical Specification (TS) Section 3.8.1 from 72 hours to 6 days. This proposed change would only be used during the upcoming Unit 2—spring 2007 refueling outage, and later during the Unit 1—spring 2008 refueling outage. The amendment would also extend the CT from 2 hours to 6 hours in TS Section 3.8.1, Required Action F.1, “Restore one required DG to OPERABLE status.” This proposed change to be used during the upcoming Unit 2—spring 2007 refueling outage, and later during the subsequent Unit 1—spring 2008 refueling outage.

Date of issuance: January 29, 2007.

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

Amendment Nos.: 180/167.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications/License.

Date of initial notice in Federal Register: June 7, 2005 (70 FR 33210). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 29, 2007.

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, Unit No. 1, St. Lucie County, Florida

Date of application for amendment: April 24, 2006, as supplemented September 14, 2006.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) consistent with the NRC-approved Revision 4 to TS Task Force (TSTF) Standard TS Change Start Printed Page 6794Traveler, TSTF-449, “Steam Generator Tube Integrity.”

Date of Issuance: January 30, 2007.

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

Amendment No.: 200.

Renewed Facility Operating License No. DPR-67: Amendment revised the TSs.

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40746). The September 14, 2006, supplement did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated: January 30, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 16, 2005, as supplemented by letter dated October 25, 2006.

Brief description of amendment: The amendment relocates Technical Specification Surveillance Requirement 4.1.4d for core spray header differential pressure instrumentation to the Updated Final Safety Analysis Report.

Date of issuance: January 31, 2007.

Effective date: January 31, 2007.

Amendment No.: 192.

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

Date of initial notice in Federal Register: March 28, 2006 (71 FR 15484). The supplemental letter dated October 25, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2007.

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: January 25, 2006.

Brief description of amendments: The amendments revised Technical Specification (TS) 1.1, “Definitions,” and TS 3.4.16, “RCS [Reactor Coolant System] Specific Activity.” The amendments replaced the current TS 3.4.16 limit on RCS gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit is based on a new dose equivalent Xe-133 definition that would replace the current E-Bar average disintegration energy definition. In addition, the current dose equivalent I-131 definition is revised to allow the use of alternate thyroid dose conversion factors.

Date of issuance: January 19, 2007.

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

Amendment Nos.: Unit 1-192; Unit 2-193.

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

Date of initial notice in Federal Register: March 14, 2006 (71 FR 13176). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 19, 2007.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of application for amendments: October 13, 2005, as supplemented on May 18, September 15 (PLA-6112 and PLA-6114), September 29, October 20, November 14, December 13, and December 14, 2006.

Brief description of amendments: The amendments revise the SSES 1 and 2 Technical Specifications (TSs) to incorporate a full-scope application of an alternate source term methodology in accordance with Title 10 of the Code of Federal Regulations, section 50.67.

Date of issuance: January 31, 2007.

Effective date: As of the date of issuance and to be implemented by October 30, 2007.

Amendment Nos.: 239 and 216.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the TSs and license.

Date of initial notice in Federal Register: August 29, 2006 (71 FR 51231). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 31, 2007.

The supplements dated September 15 (PLA-6112 and PLA-6114), September 29, October 20, November 14, December 13, and December 14, 2006, 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.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 26, 2006 (TS-457).

Brief description of amendments: The amendments revise Technical Specification (TS) Action 3.8.1.B.4 for Browns Ferry Nuclear Plant Units 2 and 3. The revision changes the restoration time of an inoperable Emergency Diesel Generator from 14 to 7 days.

Date of issuance: January 26, 2007.

Effective date: Within 60 days of NRC approval or prior to changing Unit 1 reactor mode to startup, whichever is earlier.

Amendment Nos.: 298 and 256.

Renewed Facility Operating License Nos. DPR-52 and DPR-68: Amendments revised the TSs.

Date of initial notice in Federal Register: November 21, 2006 (71 FR 67398).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 26, 2007.

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: May 25, 2006.

Brief description of amendment: The amendment revised TSs by adding Limiting Condition for Operation (LCO) 3.0.8. This change is consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-372, “Addition of LCO 3.0.8, Inoperability of Snubbers.”

Date of issuance: January 31, 2007.

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

Amendment No.: 179.

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

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40755).

The Commission's related evaluation of the amendment is contained in a Start Printed Page 6795Safety Evaluation dated January 31, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 22, 2006.

Brief description of amendment: These amendments revise the existing steam generator tube surveillance program to be consistent with the Technical Specification Task Force (TSTF) Standard TS Change Traveler, TSTF-449, “Steam Generator Tube Integrity.”

Date of issuance: October 16, 2006.

Effective date: As of the date of issuance and shall be implemented within 180 days from the date of issuance.

Amendment Nos.: 248, 228.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the licenses and the technical specifications.

Date of initial notice in Federal Register: August 1, 2006 (71 FR 43537)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 16, 2006.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 6th day of February 2007.

For the Nuclear Regulatory Commission.

John W. Lubinski,

Acting Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

End Signature End Preamble

[FR Doc. E7-2323 Filed 2-12-07; 8:45 am]

BILLING CODE 7590-01-P