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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

Pursuant to 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 February 2, 2007 through February 14, 2007. The last biweekly notice was published on February 13, 2007 (72 FR 6780).

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. Start Printed Page 8801

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

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

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

Date of amendment request: August 2, 2006.

Description of amendment request: The proposed amendment will modify the statistical summation error term “Z” and one of the allowable values for certain steam generator water level trip setpoints used in the Reactor Trip System and Engineered Safety Feature Actuation System 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:

Start Printed Page 8802

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

Response: No.

The proposed change to revise the statistical summation error term “Z” and one of the allowable values for certain steam generator water level (SGWL) reactor protection and engineered safety feature actuation functions continues to follow the current setpoint methodology previously approved for HNP [Shearon Harris Nuclear Power Plant, Unit 1] while addressing newly identified level uncertainty considerations. The proposed change does not alter the installed plant configuration for the affected instrumentation or the associated equipment system interfaces. The proposed change continues to maintain the assumptions for the specified instrument loops used in the Final Safety Analysis Report (FSAR) for HNP, and the channel statistical allowances (CSA) or calculated total loop uncertainties remain bounded by the total allowance (TA) values presented in the HNP Technical Specifications (TS). The proposed change does not alter the accident analyses or the causes for any accident described in the FSAR that credit the SGWL setpoint actuations. The proposed amendment will not modify, degrade, prevent actions or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the FSAR.

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 change to revise the statistical summation error term “Z” and one of the allowable values for certain SGWL reactor protection and engineered safety feature actuation functions addresses newly identified level uncertainty considerations. The proposed change does not implement any physical changes to the systems, structures, or components for the affected instrumentation loops or to the associated equipment system interfaces. No new or different accident initiators or sequences are created by the proposed change. The proposed change continues to maintain the safety analysis limits used in the safety analyses that credit the specified actuation 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 change to revise the statistical summation error term “Z” and one of the allowable values for certain SGWL reactor protection and engineered safety feature actuation functions addresses newly identified level uncertainty considerations and does not involve a reduction in the margin of safety for plant operation. Consistent with the requirements of the HNP FSAR, the proposed change has been evaluated to ensure that the assumptions for the specified instrument loops used in the FSAR continue to be maintained and that the CSA or calculated total loop uncertainties remain bounded by the TA values presented in the HNP TS. The proposed change continues to follow the current setpoint methodology previously approved for HNP, and the revised uncertainty analysis results in acceptable calculational margin.

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 Acting Branch Chief: Margaret H. Chernoff.

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: December 20, 2006.

Description of amendment request: The amendment will revise Technical Specification (TS) 6.12 “High Radiation Area.” Specifically, the proposed amendment would align the requirements with the revised 10 CFR 20 as described in Regulatory Guide 8.38, Revision 1, “Control of Access to High and Very High Radiation Areas in Nuclear Power Plants.”

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 changes are administrative and affect personnel access control requirements for high radiation areas. The changes do not affect the operation, physical configuration, or function of plant equipment or systems. The changes do not impact the initiators or assumptions of analyzed events; nor do they impact the mitigation of accidents or transient events. Therefore, these changes do not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The changes are administrative and affect personnel access control requirements for high radiation areas. The changes do not alter plant configuration, require installation of new equipment, alter assumptions about previously analyzed accidents, or impact the operation or function of plant equipment or systems. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The changes are administrative and affect personnel access control requirements for high radiation areas. The changes do not impact any safety assumptions; nor do the changes have the potential to reduce any margin of safety as described in the HNP [Shearon Harris Nuclear Power Plant, Unit 1] TS Bases. The proposed changes maintain an equivalent level of protection for radiation workers and, thereby, provide reasonable assurance that individuals will not exceed regulatory dose limits. The proposed changes are consistent with: (1) The guidance of Regulatory Guide (RG) 8.38, “Control of Access to High and Very High Radiation Areas in Nuclear Power Plants,” Section C, Regulatory Position 2.4, Alternative Methods for Access Control, with the exception that “should” has been changed to “shall”; and (2) other nuclear plants' existing TSs such as those at Brunswick Steam Electric Plant Units 1 & 2. Based on this evaluation, 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Branch Chief: Margaret H. Chernoff.

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

Date of amendment request: January 4, 2007.

Description of amendment request: The proposed amendments would remove gaseous radioactivity monitoring from the Technical Specifications (TSs) as an acceptable option for reactor coolant leakage detection.

Basis for proposed no significant hazards consideration determination: Start Printed Page 8803As 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:

Pursuant to 10 CFR 50.91, Duke has made the determination that this amendment request does not involve a significant hazards consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. This ensures 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.

The removal of the gaseous containment atmosphere radioactivity monitor from [the] TS as an acceptable alternative to the particulate containment atmosphere radioactivity monitor will not reduce the number of operable leak detection channels which the Technical Specification LCO [limiting condition for operation] currently provides. The gaseous monitor which is being removed from [the] Technical Specifications is the least sensitive and has the highest response time of the three available leakage monitors currently in the Technical Specification. The remaining particulate radioactivity monitor will provide greater leak detection capability by comparison. Therefore, removal of the gaseous radioactivity monitor from the Technical Specification LCO cannot increase the probability or consequence of an accident.

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

RCS [reactor coolant system] leakage detection instrumentation functions to provide control room operators with information which is indicative of a degraded RCS pressure boundary. Removal of RIA 49 from [the] TS will, in effect, remove the “weakest link” in the leakage detection system requirements of the LCO. It is important to note that RIA 49 will remain available. The change only removes it from the LCO, not from the plant. So, the result will be an enhanced capability for detecting RCS leakage in a timely manner. This enhancement, although small, could enable the operator to identify a precursor to a LOCA [loss-of-coolant accident] and take actions to safely shutdown the plant for repairs prior to actually experiencing a significant transient (LOCA). While the leakage detection system cannot prevent all LOCAs, these are accidents which have been evaluated in the UFSAR [updated final safety analysis report]. In no case would this enhancement be capable of creating a new or different kind of accident than previously evaluated.

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

The proposed change does not reduce the number of instrument channels required by the LCO for the leakage detection system. The LCO will still ensure that both a normal sump level instrument and a containment atmosphere radioactivity instrument are operable as before. It only removes one available option for satisfying the requirement for a containment atmosphere radioactivity monitor. The remaining containment atmosphere radioactivity monitor has greater sensitivity and faster response time than the monitor that is being removed from the Technical Specification. No other plant equipment is affected by the proposed change. Thus, there is no adverse impact on the capability to detect an RCS leak. 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: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202.

NRC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: May 31, 2005, as supplemented by letters dated February 8, 2006, and January 5, 2007.

Description of amendment request: The proposed 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 proposed change incorporates clarifying requirements in surveillance testing of diesel generators and new actions for an inoperable battery charger. The proposed 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.”

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 emergency diesel generators (DGs) and their associated emergency loads are accident-mitigating features. As such, testing of the DGs themselves is not associated with any potential accident initiating mechanism. Each DG is dedicated to a specific vital bus and these buses and DGs are independent of each other. There is no common mode failure provided by the testing changes proposed in this license amendment request (LAR) that would cause multiple bus failures. Therefore, there will be no significant impact on any accident probabilities by the approval of the requested amendment.

SR [surveillance requirement] changes that are consistent with Industry/Technical Specification Task Force (TSTF) Standard Technical Specification (STS) change TSTF-283, Revision 3 have been approved by the NRC and the online tests allowed by the TSTF are only to be performed for the purpose of establishing operability. Performance of these SRs during normally restricted modes will require an assessment to assure plant safety is maintained or enhanced.

The proposed changes restructure the TS for the direct current (DC) electrical power system, consistent with TSTF-360, Revision 1. The proposed changes add actions to specifically address battery and battery charger inoperability. The DC electrical power system, including associated battery chargers, is not an initiator of any accident sequence analyzed in the Final Safety Analysis Report (FSAR). Operation in accordance with the proposed TS ensures that the DC electrical power system is capable of performing its function as described in the FSAR. Therefore, the mitigating functions supported by the DC electrical power system will continue to provide the protection assumed by the analysis.

The relocation of preventive maintenance surveillances, and certain operating limits and actions, to a newly-created licensee-controlled Battery Monitoring and Maintenance Program will not challenge the ability of the DC electrical power system to perform its design function. Appropriate monitoring and maintenance, consistent with industry standards, will continue to be performed. In addition, the DC electrical power system is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure the control of maintenance activities associated with the DC electrical power system.

The integrity of fission product barriers, plant configuration, and operating procedures as described in the FSAR will not be affected by the proposed changes. Therefore, the consequences of previously analyzed accidents will not increase by implementing these changes.

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

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

Response: No. Start Printed Page 8804

The proposed changes involve restructuring the TS for the DC electrical power system. The DC electrical power system, including associated battery chargers, is not an initiator to any accident sequence analyzed in the FSAR. Rather, the DC electrical power system is used to supply equipment used to mitigate an accident.

The proposed change would create no new accidents since no changes are being made to the plant that would introduce any new accident causal mechanisms. Diesel Generators will be operated in the same configuration currently allowed by other DG SRs that allow testing in plant Modes 1 and 2 and 3. This license amendment request does not impact any plant systems that are accident initiators or adversely impact any accident mitigating systems.

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

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

Response: No.

The proposed change does not involve a significant reduction in the margin of safety. The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed changes to the testing requirements for the plant DGs do not affect the operability requirements for the DGs, as verification of such operability will continue to be performed as required. Continued verification of operability supports the capability of the DGs to perform their required function of providing emergency power to plant equipment that supports or constitutes the fission product barriers. Consequently, the performance of these fission product barriers will not be impacted by implementation of this proposed amendment.

In addition, the margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient AC and DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new battery maintenance and monitoring program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical sources will continue to provide adequate power to safety related loads in accordance with analysis assumptions.

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

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

Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.

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

Date of amendment request: September 25, 2006.

Description of amendment request: The proposed amendment would revise the Palisades Nuclear Plant (PNP) licensing bases to adopt the alternative source term (AST) as described in Title 10 of the Code of Federal Regulations (CFR) Section 50.67 following the guidance provided in Regulatory Guide (RG) 1.183. This application includes an amendment to the Technical Specifications, Definition 1.1, Dose Equivalent I-131.

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

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

Response: No.

Alternative source term calculations have been performed for PNP that demonstrate the dose consequences remain below limits specified in NRC Regulatory Guide 1.183 and 10 CFR 50.67. The proposed change does not modify the design or operation of the plant. The use of an AST changes only the regulatory assumptions regarding the analytical treatment of the design basis accidents and has no direct effect on the probability of any accident.

The AST has been utilized in the analysis of the limiting design basis accidents listed above [Loss-of-Coolant Accident, Main Steam Line Break, Steam Generator Tube Rupture, Small Line Break Outside Containment, Control Rod Ejection, Fuel Handling Accident, and Spent Fuel Cask Drop]. The results of the analyses, which include the proposed changes to the Technical Specifications, demonstrate that the dose consequences of these limiting events are all within the regulatory limits.

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

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

Response: No.

The proposed change does not affect any plant structures, systems, or components. The proposed operation of plant systems and equipment affected by this change does not create the possibility of a new or different kind of accident previously evaluated. The proposed modifications and post-modification testing are intended to enhance the capability of the plant to comply with the revised post accident dose results presented in this submittal. Since the alternative source term is a revised methodology used to estimate resulting accident doses, it is not an accident initiator.

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

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

Response: No.

The proposed implementation of the alternative source term methodology is consistent with NRC Regulatory Guide 1.183. Conservative methodologies, per the guidance of RG 1.183, have been used in performing the accident analyses. The radiological consequences of these accidents are all within the regulatory acceptance criteria associated with use of the alternative source term methodology.

The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries and in the control room are within the corresponding regulatory limits of RG 1.183 and 10 CFR 50.67. The margin of safety for the radiological consequences of these accidents is considered to be that provided by meeting the applicable regulatory limits, which are set at or below the 10 CFR 50.67 limits. An acceptable margin of safety is inherent in these limits.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the 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.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: December 14, 2006.

Description of amendment request: The proposed amendments would revise the reference to “trash racks and screens” in Technical Specification (TS) 3.5.2, “ECCS [Emergency Core Cooling System]—Operating”, Surveillance Requirement (SR) 3.5.2.8 and revise the Start Printed Page 8805required Refueling Water Storage Tank (RWST) level in TS 3.5.4, “Refueling Water Storage Tank (RWST).” This License Amendment Request (LAR) fulfills the commitment made in the supplement to Nuclear Management Company Response to Generic Letter 2004-02, “Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors,” to submit an LAR to revise SR 3.5.2.8 by December 31, 2006.

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

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

Response: No.

This license amendment request proposes to revise the Technical Specifications by changing the containment sump inlet debris interceptor description in Surveillance Requirement 3.5.2.8 and increasing the Refueling Water Storage Tank level in Surveillance Requirement 3.5.4.1 to 265,000 gallons which corresponds to approximately 90% indicated instrumentation level. These changes support resolution of containment sump blockage issues raised in Nuclear Regulatory Commission Bulletin 2003-01, “Potential Impact Of Debris Blockage On Emergency Sump Recirculation At Pressurized-Water Reactors” and Generic Letter 2004-02, “Potential Impact Of Debris Blockage On Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors.”

The containment sump inlet debris interceptor is a plant design feature which mitigates accidents and does not initiate accidents. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The new sump strainers for use as debris interceptors have been evaluated to withstand the applicable post accident loads without trash racks and thus the description change in Surveillance Requirement 3.5.2.8 does not involve a significant increase in the consequences of an accident previously evaluated.

The Refueling Water Storage Tank is required for accident mitigation and is not an accident initiator, thus requiring additional water volume in the tank does not involve a significant increase in the probability of an accident previously evaluated. Since the proposed change increases the water volume in the Refueling Water Storage Tank available for accident mitigation, this change may decrease the consequences of an accident.

Thus, the changes proposed in this license amendment request do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

This license amendment request proposes to revise the Technical Specifications by changing the containment sump inlet debris interceptor description in Surveillance Requirement 3.5.2.8 and increasing the Refueling Water Storage Tank level in Surveillance Requirement 3.5.4.1 to 265,000 gallons which corresponds to approximately 90% indicated instrumentation level. These changes support resolution of containment sump blockage issues raised in Nuclear Regulatory Commission Bulletin 2003-01, “Potential Impact Of Debris Blockage On Emergency Sump Recirculation At Pressurized-Water Reactors” and Generic Letter 2004-02, “Potential Impact Of Debris Blockage On Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors.”

The proposed Technical Specification containment sump suction inlet debris interceptor description revision does not create the possibility of a new or different kind of accident. There are no new failure modes or mechanisms created by the new strainers and there are no new accident precursors generated due to this change. The new strainers do not change the way in which the plant is operated.

The proposed Technical Specification Refueling Water Storage Tank level increase does not involve a change in system operation or the use of the Refueling Water Storage Tank. It does increase the quantity of water in the Refueling Water Storage Tank available for accident mitigation. There are no new failure modes or mechanisms created by the availability or use of an additional water volume in the Refueling Water Storage Tank as proposed by this Technical Specification change. There are no new accident precursors generated with the storage of additional water in the Refueling Water Storage Tank.

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

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

Response: No.

This license amendment request proposes to revise the Technical Specifications by changing the containment sump inlet debris interceptor description in Surveillance Requirement 3.5.2.8 and increasing the Refueling Water Storage Tank level in Surveillance Requirement 3.5.4.1 to 265,000 gallons which corresponds to approximately 90% indicated instrumentation level. These changes support resolution of containment sump blockage issues raised in Nuclear Regulatory Commission Bulletin 2003-01, “Potential Impact Of Debris Blockage On Emergency Sump Recirculation At Pressurized-Water Reactors” and Generic Letter 2004-02, “Potential Impact Of Debris Blockage On Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors.”

The proposed Technical Specification containment sump debris interceptor description revision does not involve a significant reduction in a margin of safety. The new sump strainers for use as debris interceptors have been evaluated to withstand the applicable post accident loads without trash racks and thus do not involve a significant reduction in a margin of safety. The new strainers provide additional debris interceptor flow area to the sump and thus may improve plant margins of safety.

The proposed change will increase the required water volume to be stored in the Refueling Water Storage Tank which means additional water will be available to mitigate accidents. This change does not involve a decrease in the margin of safety, but may involve an increase in the margin of safety.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve 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: P. Milano.

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: February 21, 2006.

Brief description of amendments: The amendments revise the Technical Specification (TS) 1.1, “Definitions,” and TS 3.4.16, “RCS [Reactor Coolant System] Specific Activity,” by removing the current TS 3.4.16 limits on RCS gross-specific activity with 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 would be revised to allow the use of alternate, NRC-approved thyroid dose conversion factors.

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

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

Response: No.

The proposed changes to add new thyroid dose conversion factor reference[s] to the definition of DOSE EQUIVALENT I-131, eliminate the definition of Ē—AVERAGE DISINTEGRATION ENERGY, add a new Start Printed Page 8806definition of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a limit on noble gas specific activity in the form of a Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, and make corresponding changes to TS 3.4.16 to reflect all of the above are not accident initiators and have no impact on the probability of occurrence for any design basis accidents.

The proposed changes will have no impact on the consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS iodine concentration excursion to the value currently associated with full power operation, which is more restrictive on plant operation than the existing allowable RCS iodine specific activity at lower power levels.

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

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

Response: No.

The proposed changes do not alter any physical part of the plant nor do they affect any plant operating parameters besides the allowable specific activity in the RCS. The changes which impact the allowable specific activity in the RCS are consistent with the assumptions assumed in the current radiological consequence analyses.

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

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

Response: No.

The acceptance criteria related to the proposed changes involve the allowable Control Room and offsite radiological consequences following a design basis accident. The proposed changes will have no impact on the radiological consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS iodine specific activity excursion to the value currently associated with full power operation, which is more restrictive on plant operation than the existing allowable RCS iodine specific activity at lower power levels.

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

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

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

NRC Branch Chief: David Terao.

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

Date of amendment request: January 31, 2007.

Description of amendment request: The proposed change revises the Technical Specification (TS) surveillance requirements (SR) for addressing a missed surveillance, and is consistent with the Nuclear Regulatory Commission (NRC) approved Revision 6 of Technical Specification Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-358, “Missed Surveillance Requirements.”

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

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

The proposed change to incorporate the requirements of improved STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and SR 4.0.3, respectively, does not affect the design or operation of the plant. The proposed change involves revising the existing Surry custom TS to be consistent with NUREG-1431, Revision 3, to facilitate the incorporation of TSTF-358 into the TS. The proposed change involves no technical changes to the existing TS as it merely clarifies how SRs are met. As such, these changes are administrative in nature and do not affect initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change to incorporate the requirements of improved STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and SR 4.0.3, respectively, does not involve a physical alteration to the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change revises the existing Surry TS to be consistent with NUREG-1431, Revision 3, to clarify how SRs are met and facilitates the incorporation of TSTF-358 for addressing missed surveillances. As such, the proposed change will not impose any new or different requirements or eliminate any existing requirements. 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?

The proposed change to incorporate the requirements of improved STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and SR 4.0.3, respectively, does not affect plant operation or safety analysis assumptions in any way. The change provides additional clarification on how a surveillance is met and facilitates the incorporation of TSTF-358 for addressing missed surveillances. The change is administrative in nature and does not affect the operation of safety-related systems, structures, or components. 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: Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Branch Chief: Evangelos C. Marinos.

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.

Carolina Power & Light, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: January 19, 2007.

Brief description of amendment request: The proposed amendment would modify Technical Specification Start Printed Page 8807(TS) 5.5.9 to add steam generator (SG) alternate repair criteria and TS 5.6.8 to add additional SG reporting requirements.

Date of publication of individual notice in Federal Register: January 30, 2007 (72 FR 4300).

Expiration date of individual notice: April 2, 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: February 7, 2006, as supplemented by letters dated August 14 and November 16, 2006.

Brief description of amendment: The amendment revised the Millstone Power Station, Unit No. 2 Technical Specifications to permit an increase in the allowed outage time from 72 hours to 7 days for the inoperability of the steam supply to the turbine-driven auxiliary feedwater (AFW) pump or the inoperability of the turbine-driven AFW pump under certain operating mode restrictions.

Date of issuance: January 31, 2007.

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

Amendment No.: 297.

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

Date of initial notice in Federal Register: April 11, 2006 (71 FR 18372). The supplements dated August 14, and November 16, 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.

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.

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

Date of application for amendment: April 17, 2006.

Brief description of amendment: This amendment changed the method for calculating fuel pool decay heat load from the original licensing basis methodology of ORIGEN to ORIGEN-ARP.

Date of issuance: February 8, 2007.

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

Amendment No.: 200.

Facility Operating License No. NPF-21: The amendment revised the Final Safety Analysis Report.

Date of initial notice in Federal Register : May 23, 2006 (71 FR 29674). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 8, 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: April 18, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.1.2 by changing the test frequency of the drywell-to-suppression chamber bypass leakage test from 24 months to 120 months. The amendment also added new TS SRs 3.6.1.1.3 and 3.6.1.1.4, to test the suppression chamber-to-drywell vacuum breakers on a 24-month frequency.

Date of issuance: February 9, 2007.

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

Amendment No.: 201.

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

Date of initial notice in Federal Register: May 23, 2006 (71 FR 29674). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 9, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 14, 2006, as supplemented by letter dated November 7, 2006.

Brief description of amendment: The amendment approved the removal of Surveillance Requirement 4.8.1.1.2.f from the Waterford Steam Electric Station, Unit 3, Technical Specifications. Entergy Operations, Inc. has committed to relocate this surveillance requirement, which is associated with vendor recommended inspections of the emergency diesel generators, to the Technical Requirements Manual.

Date of issuance: February 6, 2007.

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

Amendment No.: 211.

Facility Operating License No. NPF-38: The amendment revised the Operating License and the Technical Specifications.

Date of initial notice in Federal Register: August 15, 2006 (71 FR Start Printed Page 880846931). The November 7, 2006, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 6, 2007.

No significant hazards consideration comments received: No.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: March 1, 2006, as supplemented by letter dated August 17, 2006.

Brief description of amendment: The amendment modifies Special Operations Limiting Condition for Operation (LCO) 3.10.1, “System Leakage and Hydrostatic Testing Operation,” to allow more efficient testing during a refueling outage. Specifically, the LCO 3.10.1 allowance for operation with the average reactor coolant temperature greater than 212 °F (while considering operational conditions to be in Mode 4), is extended to include operations where temperature exceeds 212 °F: (1) As a consequence of maintaining adequate reactor pressure for a system leakage or hydrostatic test; or (2) as a consequence of maintaining adequate reactor pressure for control rod scram time testing initiated in conjunction with a system leakage or hydrostatic test. This change is based on the NRC-approved Technical Specification Task Force (TSTF) standard TS change TSTF-484, Revision 0.

Date of issuance: February 5, 2007.

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

Amendment No.: 264

Facility Operating License No. DPR-49: The amendment revises the TSs.

Date of initial notice in Federal Register : (71 FR 70560) December 5, 2006. The supplement 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, as published in the Federal Register on December 5, 2006 (71 FR 70560).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 5, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 10, 2006.

Brief description of amendment: The amendment revises Technical Specification Surveillance Requirement 4.1.1.3 to extend the containment airlock surveillance frequency from once per year to once every five years.

Date of issuance: February 7, 2007.

Effective date: February 7, 2007.

Amendment No.: 61.

Possession Only License No. DPR-73: The amendment revises the Technical Specifications.

Date of initial notice in Federal Register : December 5, 2006 (71 FR 70560). The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated February 7, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 14, 2006, as supplemented on December 28, 2006.

Brief description of amendment: The amendment revised Table 3.3.5.1-1, “Emergency Core Cooling System Instrumentation,” of the MNGP Technical Specifications, to permit a one-time extension of the quarterly surveillance interval (i.e., from 92 days to 140 days), for three low pressure coolant injection loop select logic functions.

Date of issuance: January 18, 2007.

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

Amendment No: 149.

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

The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

Date of initial notice in Federal Register : December 19, 2006 (71 FR 75995). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 18, 2007.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendment: November 9, 2006 (TS-458).

Brief description of amendment: The amendment deleted the Technical Specification (TS) Surveillance Requirement to verify the position of a low pressure coolant injection crosstie valve.

Date of issuance: February 6, 2007.

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

Amendment No.: 268.

Renewed Facility Operating License No. DPR-33: Amendment revised the TSs.

Date of initial notice in Federal Register : November 20, 2006 (71 FR 671600). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated: February 6, 2007.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 15th 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. E7-3199 Filed 2-26-07; 8:45 am]

BILLING CODE 7590-01-P