E7-9523. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from April 27, 2007, to May 10, 2007. The last biweekly notice was published on May 8, 2007 (72 FR 26173).

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

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

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

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

Written comments may be submitted by mail to the Chief, 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: (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, [e-mail]; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at [phone], verification number is [phone]. 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 [phone] or by e-mail to [e-mail]. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the 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 [phone], [phone] or by e-mail to [e-mail].

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

Date of amendment request: December 12, 2006.

Description of amendment request: The proposed amendment would revise Surveillance Requirement (SR) 3.3.1.1.8 and SR 3.3.1.3.2 to increase the interval between local power range monitor (LPRM) calibrations from 1000 megawatt-days per ton (MWD/T) average core exposure to 2000 MWD/T average core exposure. The proposed increase in the interval between required LPRM calibrations is acceptable due to improvements in fuel analytical bases, core monitoring processes, and nuclear instrumentation.

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 amendment revises the surveillance interval for the LPRM calibration from 1000 MWD/T average core exposure to 2000 MWD/T average core exposure. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in fuel analytical bases, core monitoring processes, and nuclear instrumentation. Therefore, the revised surveillance interval continues to ensure that the LPRM detector signal will continue to be adequately calibrated.

This change will not alter the operation of process variables, structures, systems, or components as described in the CPS [Clinton Power Station] Updated Safety Analysis Report (USAR). The proposed change does not alter the initiation conditions or operational parameters for the LPRM subsystem and there is no new equipment introduced by the extension of the LPRM calibration interval. The performance of the Average Power Range Monitor (APRM) system, Oscillation Power Range Monitor (OPRM) system, Rod Control and Information System (RC&IS) and 3D MONICORE core monitoring system is not significantly affected by the proposed surveillance interval increase. The proposed LPRM calibration interval extension will have no significant effect on the Reactor Protection System (RPS) instrumentation accuracy during power maneuvers or transients and will therefore not significantly affect the performance of the RPS. As such, the probability of occurrence for a previously evaluated accident is not increased.

The radiological consequences of an accident can be affected by the thermal limits existing at the time of the postulated accident; however, LPRM chamber exposure has no significant affect on the calculated thermal limits since LPRM accuracy does not significantly deviate with exposure. For the LPRM extended calibration interval, the total nodal power uncertainty remains less than the uncertainty assumed in the General Electric BWR [boiling water reactor] Thermal Analysis Basis (GETAB) safety limit, maintaining the accuracy of the thermal limit calculation. Therefore, the thermal limit calculation is not significantly affected by LPRM calibration frequency, and thus the radiological consequences of any accident previously evaluated are not increased.

Based on the above information, 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 previously evaluated?

Response: No.

The performance of the APRM, OPRM, RC&IS and 3D MONICORE systems is not significantly affected by the proposed LPRM surveillance interval increase. The proposed change does not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. The proposed amendment does not change or introduce any new equipment, modes of system operation or failure mechanisms.

Therefore, based on the above information, 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 has no impact on equipment design or fundamental operation, and there are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the proposed LPRM surveillance interval increase. The performance of the APRM, OPRM, RC&IS and 3D MONICORE systems is not significantly affected by the proposed change. The proposed LPRM calibration interval extension will have no significant effect on RPS instrumentation accuracy during power maneuvers or transients and will therefore not significantly affect the performance of the RPS. The margin of safety can be affected by the thermal limits existing at the time of the postulated accident; however, uncertainties associated with LPRM chamber exposure have no significant effect on the calculated thermal limits. The thermal limit calculation is not significantly affected since LPRM sensitivity with exposure is well defined. LPRM accuracy remains within the total nodal power uncertainty assumed in the GETAB, therefore maintaining thermal limits and the safety margin. The proposed change does not affect safety analysis assumptions or initial conditions and therefore, the margin of safety in the original safety analyses is maintained.

Based on the above information, the proposed change does not involve a significant reduction in a margin of safety .

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

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: January 26, 2007.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,” Table 3.3.1.1-1, “Reactor Protection System Instrumentation,” Function 8, “Scram Discharge Volume Water Level—High,” item b, “Float Switches,” by replacing Surveillance Requirement (SR) 3.3.1.1.9 with SR 3.3.1.1.12. This change will effectively revise the surveillance frequency for the scram discharge volume (SDV) level float switch from every 92 days to every 24 months.

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

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

Response: No.

The proposed TS change involves a change in the surveillance frequency for the SDV water level float switch channel functional test. The proposed TS change does not physically impact the plant. The proposed change does not affect the design of the SDV water level instruments, the operational characteristics or function of the instruments, the interfaces between the instruments and the RPS, or the reliability of the SDV water level instruments. The proposed TS change does not degrade the performance of, or increase the challenges to, any safety systems assumed to function in the accident analysis. As noted in the Bases to TS 3.3.1.1, even though the two types of SDV Water Level—High Functions are an input to the RPS logic, no credit is taken for a scram initiated from these functions for any of the design basis accidents or transients evaluated in the CPS [Clinton Power Station] Updated Safety Analysis Report (USAR). An inoperable SDV water level instrument is not considered as an initiator of any analyzed event. The proposed TS change does not impact the usefulness of the SRs in evaluating the operability of required systems and components, or the way in which the surveillances are performed. In addition, the frequency of surveillance testing is not considered an initiator of any analyzed accident, nor does a revision to the frequency introduce any accident initiators. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of a previously analyzed event are dependent on the initial conditions assumed in the analysis, the availability and successful functioning of equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The consequences of a previously evaluated accident are not significantly increased by the proposed change. The proposed change does not affect the performance of any equipment credited to mitigate the radiological consequences of an accident. The risk assessment of the proposed changes has concluded that there is an insignificant increase in the core damage frequency as well as the total population dose rate. Historical review of surveillance test results and associated maintenance records did not find evidence of failures that would invalidate the above conclusions.

Therefore, the proposed change does not alter the ability to detect and mitigate events and, as such, does not involve a significant increase in the consequences of an accident previously evaluated.

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

Response: No.

The proposed TS change does not introduce any failure mechanisms of a different type than those previously evaluated, since there are no physical changes being made to the facility. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no change being made to the parameters within which CPS is operated. There are no setpoints at which protective or mitigative actions are initiated that are affected by this proposed action. The change does not alter assumptions made in the safety analysis. This proposed action will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures, which ensure the unit remains within analyzed limits, is proposed, and no change is being made to procedures relied upon to respond to an off-normal event. As a result, no new failure modes are being introduced. The way surveillance tests are performed remains unchanged. A historical review of surveillance test results and associated maintenance records indicated there was no evidence of any failures that would invalidate the above conclusions.

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.

Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. The proposed TS change involves a change in the surveillance frequency for the SDV water level float switch channel functional test. There is no change in the design of the affected systems, no alteration of the setpoints at which alarms or actions are initiated, and no change in plant configuration from original design. The proposed change does not significantly impact the condition or performance of structures, systems, and components relied upon for accident mitigation. The proposed change does not result in any hardware changes or in any changes to the analytical limits assumed in accident analyses. Existing operating margin between plant conditions and actual plant setpoints is not significantly reduced due to these changes. The proposed change does not significantly impact any safety analysis assumptions or results.

AmerGen has conducted a risk assessment to determine the impact of a change to the SDV water level instrument surveillance frequency from the current once every 92 days to once every 24 months for the risk measures of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). This assessment indicated that the proposed CPS surveillance frequency extension has a very small change in risk to the public and is an acceptable plant change from a risk perspective.

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

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

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: April 30, 2007.

Description of amendment request: The amendment will revise the technical specifications to use other narrow range containment sump water level instrumentation rather than the existing redundant instruments to allow installation of new emergency core cooling system recirculation sumps strainers as specified in the Nuclear Regulatory Commission Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors.

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 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, and it does not change an accident previously evaluated in the Final Safety Analysis Report (FSAR). The use of other narrow range containment sump water level instruments rather than the existing narrow range containment recirculation sump water level instruments, which have level elements located inside the emergency core cooling system (ECCS) recirculation sumps, will continue to ensure that acceptable narrow range containment sump water level monitoring is maintained during post-accident conditions. Operation of the containment spray and residual heat removal systems is unchanged as a result of the proposed amendment. The level elements associated with the existing narrow range containment recirculation sump water level instruments are not accident initiators, and the FSAR does not credit these level elements in the dose analyses for loss-of-coolant accidents. The proposed amendment does not adversely affect the ability of structures, systems, or components (SSCs) to perform their design function. SSCs required for post-accident recirculation remain capable of performing their design functions.

Therefore, this amendment 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 amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated, and it does not change an accident previously evaluated in the Final Safety Analysis Report (FSAR). The use of other narrow range containment sump water level instruments rather than the existing narrow range containment recirculation sump water level instruments supports the replacement of the existing containment recirculation sump screens with new strainers in accordance with the response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors. The proposed amendment does not change the design function or the operation of the containment spray and residual heat removal systems associated with the containment recirculation sumps. The proposed amendment does not create new failure mechanisms or malfunctions or accident initiators. The proposed amendment will continue to ensure that acceptable narrow range containment sump water level monitoring is maintained during post-accident conditions, and that SSCs required for post-accident recirculation remain capable of performing their design functions.

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

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

Response: No.

The proposed amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not adversely affect a plant safety limit or a limiting safety system setting, and does not alter a design basis limit for a parameter evaluated in the FSAR. The use of other narrow range containment sump water level instruments, which meet the requirements of the FSAR, rather than the existing narrow range containment recirculation sump water level instruments, will continue to ensure that acceptable narrow range containment sump water level monitoring is maintained during post-accident conditions. The proposed amendment does not adversely affect the ability of SSCs to perform their design functions or the reliability of equipment to mitigate accidents evaluated in the FSAR. The proposed amendment will continue to ensure that SSCs required for post-accident recirculation remain capable of performing their design functions.

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

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

Attorney for licensee: 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: Thomas H. Boyce.

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

Date of amendment request: April 24, 2007.

Description of amendment request: The proposed change will add Optimized ZIRLO^TM as an acceptable fuel rod cladding material in the Waterford Steam Electric Station, Unit 3 (Waterford 3), Technical Specification (TS) 5.3.1, “Fuel Assemblies.” TS 5.3.1 currently identifies, in part, Zircaloy or ZIRLO^TM fuel rod cladding as the allowable fuel rod cladding 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-approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO^TM,” prepared by Westinghouse Electric Company, LLC (Westinghouse), addresses Optimized ZIRLO^TM and demonstrates that Optimized ZIRLO^TM has essentially the same properties as currently licensed ZIRLO^TM. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO^TM fuel cladding has been shown to meet all 10 CFR 50.46 design criteria and, therefore, will not increase the consequences of an accident.

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

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

Response: No.

Use of Optimized ZIRLO^TM clad fuel will not result in changes in the operation or configuration of the facility. Topical report WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material properties of Optimized ZIRLO^TM are similar to those of standard ZIRLO^TM. Therefore, Optimized ZIRLO^TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO^TM, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.

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 will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO^TM are not significantly different from those of standard ZIRLO^TM. Optimized ZIRLO^TM is expected to perform similarly to standard ZIRLO^TM for all normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO^TM material properties relative to standard ZIRLO^TM could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO^TM properties will be performed prior to the use of fuel assemblies with fuel rods containing Optimized ZIRLO^TM. These LOCA analyses will demonstrate that the acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized ZIRLO^TM fuel rod cladding is implemented.

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: Thomas G. Hiltz.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

Date of amendment request: March 30, 2007.

Description of amendment request: The proposed amendment would change the NMP2 Technical Specifications to reflect an expanded operating domain resulting from implementation of Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Analysis (ARTS/MELLLA). The Average Power Range Monitor (APRM) flow-biased simulated thermal power Allowable Value would be revised to permit operation in the MELLLA region. The current flow-biased Rod Block Monitor (RBM) would be replaced by a power dependent RBM, which also would require new Allowable Values. The flow-biased APRM simulated thermal power setdown requirement would be replaced by more direct power and flow dependent thermal limits administration. The Surveillance Requirement for the standby liquid control (SLC) system would be revised to require each SLC pump to deliver required flow at a discharge pressure ≥1325 psig in lieu of ≥1320 psig; the SLC relief valve setpoint would be increased from 1394 psig to 1400 psig. Finally, the proposed amendment employs a new model for performing the anticipated transients without scram (ATWS) analysis for ARTS/MELLLA conditions.

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

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

Response: No.

The proposed change eliminates the APRM flow-biased simulated thermal power setdown requirement and substitutes power and flow dependent adjustments to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits will be determined using NRC [Nuclear Regulatory Commission] approved analytical methods. The proposed change will have no effect upon any accident initiating mechanism. The power and flow dependent adjustments will ensure that the MCPR safety limit will not be violated as a result of any Anticipated Operational Occurrence (AOO), and that the fuel thermal and mechanical design bases will be maintained.

The proposed change also expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased simulated thermal power Allowable Value and the replacement of the current flow-biased RBM with a new power dependent RBM. The APRM and RBM are not involved in the initiation of any accident, and the APRM flow-biased simulated thermal power function is not credited in any NMP2 safety analyses. The proposed change will not introduce any initial conditions that would result in NRC approved criteria being exceeded and the APRM and RBM will remain capable of performing their design functions.

The Standby Liquid Control (SLC) System is provided to mitigate anticipated transients without scram (ATWS) events and, as such, is not considered an initiator of an ATWS event or any other analyzed accident. The revised SLC discharge pump test pressure neither reduces the ability of the SLC system to respond to or mitigate an ATWS event nor increases the likelihood of a system malfunction that could increase the consequences of an accident.

Based on the above discussion, it is concluded that 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 eliminates the APRM flow-biased simulated thermal power setdown requirement and substitutes power and flow dependent adjustments to the MCPR and LHGR thermal limits. Because the thermal limits will continue to be met, no analyzed transient event will escalate into a new or different type of accident due to the initial starting conditions permitted by the adjusted thermal limits.

The proposed change also expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased simulated thermal power Allowable Value and the replacement of the current flow-biased RBM with a new power dependent RBM. Changing the formulation for the APRM flow-biased simulated thermal power Allowable Value and changing from a flow-biased RBM to a power dependent RBM does not change their respective functions and manner of operation. The change does not introduce a sequence of events or introduce a new failure mode that would create a new or different [kind] of accident. While not credited, the APRM flow-biased simulated thermal power Allowable Value and associated scram trip setpoint will continue to initiate a scram to protect the MCPR safety limit. The power dependent RBM will prevent rod withdrawal when the power dependent RBM rod block setpoint is reached. No new failure mechanisms, malfunctions, or accident initiators are being introduced by the proposed change. In addition, operating within the expanded power flow map will not require any systems, structures or components to function differently than previously evaluated and will not create initial conditions that would result in a new or different kind of accident from any accident previously evaluated.

The proposed change to the SLC pump test discharge pressure is consistent with the functional requirements of the ATWS rule (10 CFR 50.62). This proposed change does not involve the installation of any new or different type of equipment, does not introduce any new modes of plant operation, and does not change any methods governing normal plant operation.

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?

Response: No.

The proposed change eliminates the APRM flow-biased simulated thermal power setdown requirement and substitutes power and flow dependent adjustments to the MCPR and LHGR thermal limits. Replacement of the APRM setdown requirement with power and flow dependent adjustments to the MCPR and LHGR thermal limits will continue to ensure that margins to the fuel cladding Safety Limit are preserved during operation at other than rated conditions. Thermal limits will be determined using NRC approved analytical methods. The power and flow dependent adjustments will ensure that the MCPR safety limit will not be violated as a result of any AOO, and that the fuel thermal and mechanical design bases will be maintained.

The proposed change also expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased simulated thermal power Allowable Value and the replacement of the current flow-biased RBM with a new power dependent RBM. The APRM flow-biased simulated thermal power Allowable Value and associated scram trip setpoint will continue to initiate a scram to protect the MCPR safety limit. The RBM will continue to prevent rod withdrawal when the power dependent RBM rod block setpoint is reached. The MCPR and LHGR thermal limits will be developed to ensure that fuel thermal mechanical design bases remain within the licensing limits during a control rod withdrawal error event and to ensure that the MCPR safety limit will not be violated as a result of a control rod withdrawal error event. Operation in the expanded operating domain will not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. AOOs and postulated accidents within the expanded operating domain will continue to be evaluated using NRC approved methods. The 10 CFR 50.46 acceptance criteria for the performance of the ECCS [emergency core cooling system] following postulated LOCAs [loss-of-coolant accidents] will continue to be met.

The proposed change to the SLC pump discharge test pressure does not alter the results of any accident analyses. The proposed change is consistent with the functional requirements of the ATWS rule (10 CFR 50.62). The ability of the SLCS to respond to and mitigate an ATWS event is not affected.

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

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

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006.

NRC Branch Chief: Mark G. Kowal.

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

Date of amendment requests: April 17, 2007.

Description of amendment requests: A change is proposed to the standard technical specifications (STS) (NUREGs 1430 through 1434) and plant-specific technical specifications (TS), to strengthen TS requirements regarding control room envelope (CRE) habitability by changing the action and surveillance requirements associated with the limiting condition for operation operability requirements for the CRE emergency ventilation system, and by adding a new TS administrative controls program on CRE habitability. Accompanying the proposed TS change are appropriate conforming technical changes to the TS Bases. The proposed revision to the Bases also includes editorial and administrative changes to reflect applicable changes to the corresponding STS Bases, which were made to improve clarity, conform with the latest information and references, correct factual errors, and achieve more consistency among the STS NUREGs. The proposed revision to the TS and associated Bases is consistent with STS as revised by TS Task Force (TSTF)-448, Revision 3, “Control Room Envelope Habilitability.”

The proposed amendment would revise the TS Improvement To Modify Requirements Regarding CRE Habitability using the Consolidated Line Item Improvement Process, based on the NRC-approved to TSTF-448, Revision 3. The NRC staff issued a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments adopting TSTF-448, 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 January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the following NSHC determination in its application dated April 17, 2007.

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

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

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. 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: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: February 28, 2007.

Brief description of amendments: The proposed amendment request would revise the language of Technical Specification (TS) 3.7.1.2, “Auxiliary Feedwater System,” Action b from “MODE 3 may be entered with an inoperable turbine-driven auxiliary feedwater pump for the purposes of performing Surveillance Requirement 4.7.1.2.1a.2” to “MODE 3 may be entered with an inoperable turbine-driven auxiliary feedwater pump.”

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 deletion of the existing words in TS 3.7.1.2 Action b is an administrative change that will clarify the Licensing Basis for the turbine-driven auxiliary feedwater pump. Since this change does not change the Licensing Basis for TS 3.7.1.2, this change cannot affect the probability or consequence of any accident previously evaluated.

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

Response: No.

The proposed deletion of the existing words in TS 3.7.1.2 Action b is an administrative change that will clarify the Licensing Basis for the turbine-driven auxiliary feedwater pump. Since this change does not change the Licensing Basis for TS 3.7.1.2, this change cannot affect the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed deletion of the existing words in TS 3.7.1.2 Action b is an administrative change that will clarify the Licensing Basis for the turbine-driven auxiliary feedwater pump. Since this change does not change the Licensing Basis for TS 3.7.1.2, this change cannot 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: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: April 5, 2007.

Description of amendment request: The proposed amendments would revise technical specifications (TSs) to change the surveillance frequency for the turbine trip functions of the reactor trip system instrumentation. The current frequency is prior to each reactor startup and the proposed change will revise this to be prior to exceeding the Permissive P-9 interlock whenever the unit has been in hot standby. The proposed change is consistent with NRC-approved Technical Specification Task Force Traveler TSTF-311, as incorporated into the latest revision of Standard TSs (NUREG-1431, Revision 3).

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

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

Response: No.

The proposed changes revise the surveillance frequency for reactor trip functions from a turbine trip event. These changes do not alter these functions physically or how they are maintained. Delaying the performance of the surveillance up to the P-9 interlock will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation. The incorporation of a surveillance frequency that is consistent with the applicability for the function eliminates potential misapplication of the TS requirements. The frequency changes support turbine trip operability during plant startup and are consistent with their ability to perform the reactor trip functions. Since these changes will not affect the ability of these trips to perform the initiation of reactor trips when appropriate, the off-site dose consequences for an accident will not be impacted. Equally, the potential to cause an accident is not affected because no plant system or component has been altered by the proposed changes. 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 changes only affect the surveillance frequency requirement for the turbine trip functions. This does not affect any physical features of the plant or the manner in which these functions are utilized. The proposed surveillance frequency will require the functions to be verified operable before the turbine trip functions are applicable and able to perform their trip functions. Delaying the performance of the surveillance up to the P-9 interlock will continue to ensure operability of the function before the plant is in a condition that would benefit from the associated actuation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes do not alter any plant setpoints or functions that are assumed to actuate in the event of postulated accidents. In fact, the proposed changes do not alter any plant feature and only alter the requirements for when the function must be verified to be operable through surveillance testing. The proposed changes ensure the functionality of the turbine trips when assumed in the analysis for accident mitigation. 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Thomas H. Boyce.

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

Date of amendment request: December 19, 2006.

Brief description of amendments: The proposed amendment request would revise the requirements in Technical Specification (TS) 5.5.8, “Inservice Testing Program,” to update references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves, and address the applicability of Surveillance Requirement 3.0.2 to other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program.

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 changes revise TS 5.5.8, “Inservice 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 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. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

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

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 changes 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, these proposed changes do not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No.

The proposed changes revise TS 5.5.8, “Inservice Testing Program,” for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves. The proposed changes incorporate 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, these proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Thomas G. Hiltz.

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 51.22(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 [phone], [phone] or by e-mail to [e-mail].

Consumers Energy Company, Entergy Nuclear Palisades, LLC, and Entergy Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point Facility, Charlevoix County, Michigan

Date of application for amendment: October 31, 2006.

Brief description of amendment: The license amendment reflects the changes in ownership and operating authority for the Big Rock Facility and its Independent Spent Fuel Storage Installation.

Date of issuance: April 11, 2007.

Effective date: As of the date of issuance.

Amendment No.: 127.

Facility Operating License No. DPR-06: The license amendment reflects the changes in ownership and operating authority for the Big Rock Facility and its Independent Spent Fuel Storage Installation.

Date of initial notice in Federal Register: January 30, 2007 (72 FR 4302-4303). The Commission's related evaluation of the amendment is contained in a safety evaluation report dated April 6, 2007, which is accessible to members of the public through ADAMS (Accession Number ML070920385).

No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin

Date of application for amendment: January 10, 2007, as supplemented by letters dated April 5 and 27, 2007.

Brief description of amendment: The amendment modifies the emergency diesel generators short-time load testing requirements.

Date of issuance: May 1, 2007.

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

Amendment No.: 191.

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

Date of initial notice in Federal Register: February 5, 2007 ( 72 FR 5303). The supplemental letters provided clarifying information that did not expand the scope of the original application or change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 1, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 11, 2006.

Brief description of amendments: The amendments revised an organizational description in the Technical Specification Section 5.2.1, “Onsite and Offsite Organizations.” The change revises the title of Executive Vice President to Group Vice President to reflect title changes made by the licensee following the indirect transfer of the facility operating licenses. The indirect transfer was reviewed and approved by the NRC. This change is solely administrative in nature.

Date of issuance: April 13, 2007.

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

Amendment Nos.: 239, 221.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: March 13, 2007 (72 FR 11387). The Commission's related evaluation, final no significant hazards consideration finding, and State consultation of the amendments is contained in a Safety Evaluation dated April 13, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 8, 2007, as supplemented March 27, April 13, and May 3, 2007.

Brief description of amendments: The amendments revise the McGuire Nuclear Station, Units 1 and 2, Technical Specification 3.5.2.8, and the associated Bases and authorize changes to the Updated Final Safety Analysis Report (USFAR) concerning modifications to the emergency core cooling system sump.

Date of issuance: May 4, 2007.

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

Amendment Nos.: 240, 222.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications and authorize changes to the UFSAR.

Date of initial notice in Federal Register: March 19, 2007 (72 FR 12835).

The supplements dated March 27, April 13, and May 3, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation, final no significant hazards finding, and state consultation of the amendments are contained in a Safety Evaluation dated May 4, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 19, 2006, as supplemented by letter dated February 28, 2007.

Brief description of amendment: The amendment revised River Bend Station (RBS), Unit 1, Technical Specifications (TS) Surveillance Requirement (SR) 3.6.1.3.5 to replace the currently specified frequency for leak testing containment purge supply and exhaust isolation valves with resilient seal materials with a requirement to test these valves in accordance with the RBS's Primary Containment Leakage Rate Testing Program. RBS's Primary Containment Leakage Rate Testing Program is implemented in accordance with the Title 10 of the Code of Federal Regulations, Part 50, Appendix J, Option B, and Regulatory Guide (RG) 1.163, “Performance-Based Containment Leak Test Program,” dated September 1995. RG 1.163 allows a nominal test interval of 30 months for containment purge and vent valves.

Date of issuance: May 3, 2007.

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

Amendment No.: 152.

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

Date of initial notice in Federal Register: October 24, 2006 (71 FR 62310). The supplement dated February 28, 2007, 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 May 3, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 31, 2005, as supplemented by letters dated February 8, 2006, and January 5, February 13, February 22, and March 22, 2007.

Brief description of amendment: The amendment modifies Technical Specification (TS) Sections 3.8.1, “AC [Alternating Current] Sources—Operating,” 3.8.4, “DC [Direct Current] Sources—Operating,” 3.8.5, “DC Sources—Shutdown,” 3.8.6, “Battery Cell Parameters,” and 5.5, “Programs and Manuals.” The change incorporates clarifying requirements in surveillance testing of diesel generators and new actions for an inoperable battery charger. The change includes a revision to the Administrative Program to be consistent with Institute of Electrical and Electronics Engineers Standard 450-2002, and changes consistent with TS Task Force (TSTF) Traveler TSTF-360, Revision 1, “DC Electrical Rewrite,” and TSTF-283, Revision 3, “Modify Section 3.8 Mode Restriction Notes.”

Date of issuance: May 1, 2007.

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

Amendment No.: 204.

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

Date of initial notice in Federal Register: February 27, 2007 (72 FR 8803). The supplemental letters dated February 8, 2006, and January 5, February 13, February 22, and March 22, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 1, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 18, 2005, as supplemented by letter dated February 23, 2007.

Description of amendment request: The proposed amendment revised applicability requirements related to single control rod withdrawal allowances in shutdown modes. The amendment also corrected a typographical error and administratively relocated the existing TS 3/4.10.D, “Multiple Control Rod Removal,” to TS 3/4.14.E to be consistent with the intent and presentation of special operations.

Date of issuance: April 25, 2007.

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

Amendment No.: 228.

Facility Operating License No. DPR-35: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: January 3, 2006 (71 FR 148). The February 23, 2007, supplemental letter provided additional information that clarified the application, but 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 April 25, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 16, 2007, as supplemented by letter dated April 10, 2007.

Brief description of amendment: The amendment revises the values of the safety limit minimum critical power ratio (SLMCPR) in the Quad Cities Nuclear Power Station (Quad Cities), Unit 1, Technical Specification (TS) Section 2.1.1, “Reactor Core SLs [Safety Limits].” Specifically, the proposed change would require that for Unit 1, the minimum critical power ratio shall be greater than or equal to 1.11 for two recirculation loop operation, or greater than or equal to 1.13 for single recirculation loop operation. This change is needed to support the next cycle of operation for Quad Cities, Unit 1.

Date of issuance: May 2, 2007.

Effective date: As of the date of issuance and shall be implemented prior to startup from Q1R19 Refueling Outage.

Amendment No.: 234.

Renewed Facility Operating License No. DPR-29: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register: March 13, 2007 (71 FR 11388). The supplements contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 2, 2007.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: February 14, 2006, as supplemented by letters dated October 17, 2006, and February 8, 2007.

Brief description of amendment: The amendment revised Perry Nuclear Power Plant, Unit No. 1, Technical Specifications (TSs) to change the frequency of the Mode 5 Intermediate Range Monitoring Instrumentation CHANNEL FUNCTIONAL TEST contained in TS 3.3.1.1 from 7 days to 31 days.

Date of issuance: April 27, 2007.

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

Amendment No.: 141.

Facility Operating License No. NPF-58: This amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: March 28, 2006 (71 FR 15484) The October 17, 2006 and February 8, 2007 supplements, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 27, 2007.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: November 21, 2005, as supplemented by letter dated February 22, 2007.

Brief description of amendment: This amendment revised the acceptance criteria of technical specification (TS) surveillance requirements associated with TS 3.8.1, to modify the emergency diesel generator start tests to provide minimum voltage and frequency limits and clarified other limits as steady state parameters.

Date of issuance: April 30, 2007.

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

Amendment No.: 142.

Facility Operating License No. NPF-58: This amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: January 17, 2006 (71 FR 2591) The February 22, 2007, supplement contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

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

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of application for amendments: April 27, 2006, as supplemented December 5, 2006 and March 1, 2007.

Brief description of amendments: These amendments revised 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: April 27, 2007.

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

Amendment Nos: 233 and 228.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40748). The supplements dated December 5, 2006, and March 1, 2007, provided additional information clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 27, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 6, 2006.

Brief description of amendments: The amendments revise information in the Final Safety Analysis Report (FSAR) regarding the reactor pressure vessel Charpy upper shelf energy (USE) requirements of Title 10 of the Code of Federal Regulations Part 50, Appendix G, Section IV.A.1.c. The change updates the analysis for satisfying the RPV Charpy USE requirements through the end of the current operating licenses.

Date of issuance: May 10, 2007.

Effective date: As of the date of issuance and shall be incorporated into the FSAR during the next update of the FSAR, as required by 10 CFR 50.71(c).

Amendment Nos.: 227 and 232.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revise the Final Safety Analysis Report and the Licenses.

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40750).

The Commission's related evaluation of the amendments is contained in a safety evaluation dated May 10, 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: May 31, 2006.

Brief description of amendments: The amendments correct administrative errors in the SSES 1 and 2 Technical Specifications (TSs) by adding a logical “AND” connector in Condition B of TS 3.8.1 for SSES 1, “AC Sources—Operating,” and correct the routing of Interstate Route 80 on Figure 4.1-2 of TSs 4.1.2, “Low Population Zone,” for SSES 1 and 2.

Date of issuance: April 26, 2007.

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

Amendment Nos.: 243 and 221.

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

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75996).

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

No significant hazards consideration comments received: No.

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 application for amendments: November 15, 2006 January 11, 2007, as supplemented by letters dated January 11, and April 24, 2007.

Description of amendment request: The amendments revised the Fire Protection License Condition numbers (13), (14), and (7) for Units 1, 2, and 3, respectively, to accommodate operation.

Date of issuance: April 25, 2007.

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

Amendment Nos.: 271, 300, and 259.

Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the Operating Licenses.

Date of initial notice in Federal Register: December 19, 2006 (71 FR 76000). The supplements dated January 11, and April 24, 2007, 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 amendments is contained in a Safety Evaluation dated April 25, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 22, 2006, supplemented by letter dated September 12, 2006.

Brief description of amendments: The amendments revised the Technical Specification (TS) 3.8.1 entitled, “AC Sources—Operating.” Specifically, the proposed change would revise the completion time for TS 3.8.1, Condition F, Required Action F.1 from 12 hours to 24 hours.

Currently, TS 3.8.1, Condition F requires that an inoperable safety injection (SI) sequencer must be restored to operable status within 12 hours. If this completion time is not met, Condition G becomes applicable and the plant must be shutdown to at least Mode 3 within the following 6 hours. The proposed change to the completion time for TS 3.8.1, Condition F, Required Action F.1 provides more time to complete necessary repairs and required post-work testing to restore an inoperable SI sequencer to operable status prior to commencing a plant shutdown to Mode 3.

Date of issuance: April 27, 2007.

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

Amendment Nos.: NPF-87—138, NPF-89—138.

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

Date of initial notice in Federal Register: March 28, 2007 (72 FR 14623).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 27, 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: June 7, 2006.

Brief description of amendment: The amendment deleted Required Action D.1.2 in Technical Specification (TS) 3.7.10, “Control Room Emergency Ventilation System (CREVS),” and Required Action C.1.2 in TS 3.7.11, “Control Room Air Conditioning System (CRACS).” For TS 3.7.13, “Emergency Exhaust System (EES),” the amendment also deletes the phrase “in MODE 1, 2, 3, or 4” from Condition A (one EES train inoperable) and revised Condition D to state the following: “Required Action and associated Completion Time of Condition A not met during movement of irradiated fuel assemblies in the fuel building.”

Date of issuance: May 9, 2007.

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

Amendment No.: 184.

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

Date of initial notice in Federal Register: August 1, 2006 (71 FR 43536)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 9, 2007.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: January 31, 2007.

Brief Description of amendments: These amendments revised the Technical Specification surveillance requirements for addressing a missed surveillance, and is consistent with the Nuclear Regulatory Commission approved Revision 6 of Technical Specification Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-358, “Missed Surveillance Requirements.”

Date of issuance: May 3, 2007.

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

Amendment Nos.: 253, 252.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments changed the licenses and the technical specifications.

Date of initial notice in Federal Register: February 27, 2007 (72 FR 8806).

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 3, 2007.

Dated at Rockville, Maryland, this 11th day of May, 2007.

For the Nuclear Regulatory Commission.

Timothy McGinty,

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