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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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Start Preamble Start Printed Page 33779

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 May 24, 2007, to June 6, 2007. The last biweekly notice was published on June 5, 2007 (72 FR 31097).

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 Start Printed Page 33780fact. 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.

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

Date of amendment request: May 10, 2007.

Description of amendment request: In 2004, the Nuclear Regulatory Commission (NRC) imposed a license condition that requires the submission of a coupon surveillance program for the Unit 1 Spent Fuel Pool (SFP) racks. The coupon surveillance program is necessary to support an approved license amendment which established acceptable boron concentrations in the Unit 1 SFP.

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

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

The proposed surveillance program supports evaluation of degradation of the neutron absorbing material in the Unit 1 Spent Fuel Pool (SFP). The function of the neutron absorbing material is to provide one means of maintaining criticality safety of the nuclear fuel stored in the SFP.

The postulated accidents for the SFP are basically five types; (1) dropped fuel assembly on top of the storage rack, (2) a misloading accident, (3) an abnormal location of a fuel assembly, (4) loss-of-normal cooling to the SFP, and (5) dilution of boron in the SFP water.

The proposed change in the coupon surveillance program for the Unit 1 SFP racks does not affect any of these previously evaluated accidents. The coupon trees have been evaluated as required by our plant modifications program and have been determined to have no effect on accidents in the SFP.

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

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

The proposed surveillance program supports evaluation of degradation of the neutron absorbing material in the Unit 1 SFP. The function of the neutron absorbing material is to provide one means of maintaining criticality safety of the nuclear fuel stored in the SFP.

The coupon trees have been evaluated as required by our plant modifications program and do not create the possibility of a new or different kind of accident in the SFP. The surveillance coupons have existed in the SFP since the Unit 1 SFP racks were installed. The form and function of the surveillance coupon trees is not changed because of the need to change the coupon surveillance program. The interaction of the coupons with the spent fuel racks and the SFP is not changed due to the proposed surveillance program change.

The proposed change will not result in any other change in the plant configuration or equipment design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed coupon surveillance program supports evaluation of degradation of the neutron absorbing material in the Unit 1 SFP. The function of the neutron absorbing material is to provide one means of maintaining criticality safety of the nuclear fuel stored in the SFP. Evaluation of the coupons as part of an ongoing surveillance program provides assurance that the fuel will remain subcritical under all postulated conditions.

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 proposed to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Branch Chief: Mark G. Kowal. Start Printed Page 33781

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

Date of amendments request: May 2, 2007.

Description of amendments request: The proposed amendment would modify Technical Specification (TS) requirements for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9. The changes are consistent with the Nuclear Regulatory Commission approved Technical Specification Task Force (TSTF)-427, Revision 2. The availability of this TS improvement was published in the Federal Register on October 3, 2006 (71 FR 58444) as part of the consolidated line item improvement process.

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

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

The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in Regulatory Guide 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation (71 FR 58449). Therefore, this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Operations, Inc., Docket No. 50-003, Indian Point, Unit 1, Buchanan, New York

Date of application for amendment: February 22, 2007.

Description of amendment request: The proposed amendment would enable the licensee to make changes to the Final Safety Analysis Report (FSAR) to reflect use of the non-single-failure-proof Fuel Handling Building (FHB) 75 ton crane for dry spent fuel cask handling operations.

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:

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

Response: No.

The proposed amendment introduces no new mode of plant operations and does not affect Structures, Systems, and Components (SSCs) associated with power production, accident mitigation, or safe plant shutdown. The SSCs affected by this proposed amendment are the Indian Point, Unit 1 (IP-1) FHB 75-ton crane, the FHB concrete structure, the spent fuel storage canister, the spent fuel transfer cask, and the spent fuel inside the storage canister. A hypothetical drop of a 30 ton spent fuel shipping cask has been previously evaluated by the NRC and found to be acceptable based on the physical arrangement of plant equipment and the fact that the load path is entirely over concrete floors founded on bedrock or engineered fill over bedrock. The increased mass of the HI-TRAC transfer cask containing a fuel-loaded Multi-Purpose Canister (MPC)consequently results in no change to the basis for the original cask handling approval.

With this amendment, fewer HI-TRAC casks will be required to be loaded, lifted, and handled, a planned total of five, than the previous cask handling effort which involved loading and handling 120 casks. The HI-TRAC cask is within the design capability of the IP-1 FHB 75 ton crane, therefore the probability of an accident is not increased.

The new analyses of hypothetical drops of a loaded transfer cask confirm that there is no release of radioactive material from the storage canister and no unacceptable damage to the fuel, MPC, or transfer cask.

The hypothetical drop of a spent fuel canister lid into an open, fuel-filled canister in the cask loading pool during fuel loading has been evaluated. [Additionally, the drop of a single spent fuel assembly into an open fuel-filled canister in the cask loading pool, due to the potential damage of spent fuel assemblies in the canister, has been evaluated.] The radiological consequences of these events are less than 2% of regulatory requirements and are bounded by the licensing basis of IP-1.

Since the hypothetical drops result in lesser g loads on the fuel than the design criterion, there is no rearrangement of the fuel or deformation of the fuel basket in the canister such that a critical geometry is created.

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

Response: No.

The proposed amendment introduces no new mode of plant operations and does not affect SSCs associated with power production, accident mitigation, fuel pool cooling, or SAFSTOR configuration. The SSCs affected by this proposed amendment are the non-single-failure proof 75 ton crane, structural portions of the FHB, the spent fuel canister, the spent fuel transfer cask, and the spent fuel inside the canister.

The design function of the IP-1 FHB 75 ton crane is not changed. The HI-STORM System Start Printed Page 33782load drops create the possibility of a new initiator of an accident previously evaluated (failure of fuel cladding) caused by the postulated non-mechanistic single failure of a component in the FHB 75 ton crane.

The current licensing basis includes evaluations of the consequences of a spent fuel cask drop into the cask load pool. The new initiators include the drop of a fuel transfer cask and a drop of a spent fuel canister lid into the open, fuel filled canister in the cask loading pool and a drop of individual assemblies into the MPC. These new initiators create hypothetical accidents that are comparable in consequences to and bounded by those previously evaluated. For the drop of a spent fuel transfer cask, the consequences of cask impact on facility SSCs are bounded by the current licensing scenario of a shipping cask drop. That is, there is no significant damage to the FHB structure or on any SSCs used for safe storage of spent fuel, and there is no release of radioactive material. These new analyses of the drop of a loaded transfer cask confirm that there is no release of radioactive material from the storage container and no unacceptable damage to the fuel, MPC, or transfer cask.

For the drop of the spent fuel canister lid, with the maximum number of assemblies in the canister at 32, or the drop of a single spent fuel assembly into a fuel-filled canister, doses are calculated to be less than 2% of regulatory limits. Further the previously analyzed 100 percent cladding failure of 160 assemblies bounds the event. There is no rearrangement of the fuel in the canister such that a critical geometry is created as a result of an MPC lid drop.

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

Response: No.

The proposed amendment introduces no new mode of plant operations and does not affect SSCs associated with spent fuel storage, spent fuel pool cooling, or the integrity of SSCs in the SAFSTOR mode. The SSCs affected by this proposed amendment are the non-single-failure-proof FHB 75 ton crane, structural portions of the FHB, the spent fuel storage canister, the spent fuel transfer cask, and the spent fuel inside the canister. This amendment does not affect the fuel stored in the spent fuel pool or any SSC associated with safe storage of the fuel. The design function of the 75 ton crane is not changed. The proposed changes to plant procedures needed to implement dry cask storage do not exceed or alter a design basis or safety limit associated with accident mitigation, SAFSTOR, or fuel clad integrity.

This proposed amendment results in a net benefit based upon the larger capacity cask being used to move and store the fuel (32 assemblies per canister versus two assemblies). All the fuel can be removed from the spent fuel pool with far fewer cask lifts, welding evolutions, and storage placement. Because the maximum weight of the cask loaded with spent fuel is the same as the original design and tested capacity of the crane, design safety margins for use of the 75 ton crane remain unchanged.

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, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Acting Branch Chief: John Buckley.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of amendment request: March 15, 2007.

Description of amendment request: The proposed amendment would change Technical Specification (TS) Section 1.4 and Section 5. Changes to TS 1.4 would incorporate Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specification Changes TSTF-284, “Add ‘Met vs. Perform’ to Specification 1.4, Frequency,” Revision 3, TSTF-485-A, “Correction Example 1.4-1,” Revision 0, and make administrative changes. Changes to TS Section 5 would incorporate NRC-approved TSTF-258, “Changes to Section 5.0, Administrative Controls,” Revision 4, NRC-approved TSTF-273, “[Safety Functions Determination Program] SFDP Clarifications,” Revision 2, as amended by Westinghouse Owners Group (WOG) editorial change WOG-ED-23, and make administrative changes.

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 does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative or provide clarification only.

The proposed changes do not have any impact on the integrity of any plant system, structure, or component that initiates an analyzed event. The proposed changes will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Thus, the probability of any accident previously evaluated is not significantly increased.

The proposed changes do not affect the ability to mitigate previously evaluated accidents, and do not affect radiological assumptions used in the evaluations. The proposed changes do not change or alter the design criteria for the systems or components used to mitigate the consequences of any design basis accident. The proposed amendment does not involve operation of the required structures, systems, or components (SSCs) in a manner or configuration different from those previously recognized or evaluated. Thus, the radiological consequences of any accident previously evaluated are not increased.

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

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

Response: No.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not involve a physical alteration of any SSC or a change in the way any SSC is operated. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested.

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

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

Response: No.

The amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not affect any margin of safety. The proposed amendment does not involve any physical changes to the plant or manner in which the plant is operated.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

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

Attorney for licensee: William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: L. Raghavan.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of amendment request: April 18, 2007. Start Printed Page 33783

Description of amendment request: The proposed amendment would change Technical Specification (TS) Surveillance Requirement (SR) 3.5.2.9, to support resolution of containment sump issues raised in Nuclear Regulatory Commission (NRC) Generic Letter (GL) 2004-02, “Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors.” The proposed change to TS SR 3.5.2.9 would make the surveillance consistent with the plant design following planned modifications to the containment sump.

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 does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes to TS SR 3.5.2.9 do not have any impact on the integrity of any plant system, structure, or component (SSC) that initiates an analyzed event. The proposed changes do not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Thus, the probability of any accident previously evaluated is not significantly increased.

The proposed changes do not affect the ability to mitigate previously evaluated accidents, and do not affect radiological assumptions used in the evaluations. The proposed changes to TS SR 3.5.2.9 do not change or alter the design criteria for the systems or components used to mitigate the consequences of any design basis accident. The proposed amendment does not involve operation of the required structures, systems, or components in a manner or configuration different from those previously recognized or evaluated. The proposed changes to TS SR 3.5.2.9 provide assurance that the sump flowpath is unrestricted and stays in proper operating condition. Thus, the radiological consequences of any accident previously evaluated are not increased.

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

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

Response: No.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment to modify TS SR [3.]5.2.9 does not involve a physical alteration of any SSC or a change in the way any SSC is operated. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested.

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

3. Does the proposed amendment 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 changes do not adversely affect any plant safety limits, set points, or design parameters. The proposed changes do not adversely affect the fuel, fuel cladding, primary coolant system (PCS), or containment integrity. The proposed TS SR 3.5.2.9 changes ensure that the containment sump is unrestricted and stays in proper operating condition. The proposed changes would make the surveillance consistent with the plant design following planned modifications to the containment sump.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

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

Attorney for licensee: William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: April 22, 2007.

Description of amendment request: The proposed amendments would delete the Unit 2 license condition that requires reporting violations of other requirements conditions and delete Technical Specifications (TS) 6.6 for both units that require the NRC be notified of reportable events pursuant to 10 CFR 50.73. This request also includes an administrative TS change for both Units by changing references of the “Topical Quality Assurance Report” to the “Quality Assurance Topical Report.” The NRC staff issued a notice of opportunity to comment in the Federal Register on August 29, 2005 (70 FR 51098), on possible amendments to eliminate the license condition involving reporting of violations of other requirements (typically in License Condition 2.C) in the operating license, 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 model for referencing in license amendment applications in the Federal Register on November 4, 2005 (70 FR 67202).

The licensee affirmed the applicability of the NSHC determination in its application dated April 22, 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:

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

Response: No.

The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the 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 deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Thomas H. Boyce.

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

Date of amendment request: May 11, 2007.

Description of amendment request: The proposed amendment would modify Surveillance Requirement (SR) 3.3.1.18, pertaining to the reactor trip on turbine trip function, in the Technical Start Printed Page 33784Specifications (TS). The existing SR requires that the SR be met before reaching the P-7 interlock (approximately at 10 percent reactor power). The licensee proposed to change the SR such that the SR will be met before reaching the P-8 interlock (approximately at 31 percent reactor power). This proposed change would ensure consistency between the SR and the mode of applicability for the reactor trip on turbine trip function.

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

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

Response: No.

The proposed change revises a Technical Specification (TS) [s]urveillance [r]equirement (SR) [f]requency associated with the reactor trip on turbine trip function to be consistent with the mode of applicability for the function. The change to the frequency from prior to exceeding the P-7 interlock to prior to exceeding the P-8 interlock does not create any new credible single failure. The P-7 and P-8 interlocks are not accident initiators. The reactor trip on turbine trip function is an anticipatory trip, and the safety analysis does not credit this trip for protecting the reactor core. The consequences of accidents previously evaluated are unaffected by this change because no change to any accident mitigation scenario has resulted and there are no additional challenges to fission product barrier integrity.

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.

No changes are being made to the plant that would introduce any new accident causal mechanisms. The proposed change to the interlock at which the surveillance is performed in support of a reactor trip on turbine trip does not adversely affect previously identified accident initiators and does not create any new accident initiators. The change does not affect how the associated trip function operates. No new single failures or accident scenarios are created by the proposed change and the proposed change does not result in any event previously deemed incredible being made credible.

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.

No safety analyses were changed or modified as a result of the proposed change in the surveillance frequency. All margins associated with the current safety analyses acceptance criteria are unaffected. The current safety analyses remain bounding. The safety systems credited in the safety analyses will continue to be available to perform their mitigation functions. The proposed change does not affect the availability or operability of safety-related systems and components.

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

The Nuclear Regulatory Commission (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: James M. Petro, Jr., Esquire, One Cook Place, Bridgman, MI 49106.

NRC Acting Branch Chief: Travis L. Tate.

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

Date of amendment request: May 16, 2007.

Description of amendment request: 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 STS change traveler TS Task Force (TSTF)-448, Revision 3, “Control Room Envelope Habitability.”

The proposed amendment would revise the TS 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 May 16, 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 Start Printed Page 33785functioning 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: James R. Curtiss, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: May 21, 2007.

Description of amendment request: The proposed amendment would add Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4).

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252) for model safety evaluation and November 24, 2004 (69 FR 68420) for NSHC. The licensee affirmed the applicability of the model NSHC determination in its application dated May 21, 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:

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed. The postulated seismic event requiring snubbers is a low-probability occurrence and the overall TS system safety function would still be available for the vast majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff 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.

RC Branch Chief: Evangelos C. Marinos.

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

Date of amendment request: March 14, 2007, as supplemented by letters dated April 18 and May 9, 2007.

Description of amendment request: The amendment would revise Technical Specifications (TSs) 3.3.2, “Engineered Safety Features Actuation System Instrumentation”; 3.7.2, “Main Steam Isolation Valves (MSIVs)”; and 3.7.3, “Main Feedwater Isolation Valves (MFIVs).” The proposed TS changes address the following changes to the plant and/or plant TSs: (1) The modification of the main steam and feedwater isolation system (MSFIS), which provides the signal to actuate the MSIVs and MFIVs, and changes to TS 3.3.2; (2) the replacement of the MSIVs and MFIVs, and associated actuators; (3) the addition of the main feedwater regulating valves (MFRVs), and associated MFRV bypass valves, to TS 3.7.3; (4) the relocation of the MSIV and MFIV isolation times from TSs 3.7.2 and 3.7.3 to the TS Bases; and (5) the changes to page numbers in the TS Table of Contents.

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

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

Response: No. Start Printed Page 33786

Evaluations and/or reanalysis assessing the impact of the replacement MSFIS, MSIVs and MFIVs and actuators, and the increased closure time on non-LOCA [non-loss-of-coolant accident] transients; SBLOCA [small-break LOCA] transients; main steam line break mass and energy releases inside and outside containment; containment pressure and temperature response to a postulated main steam line break; environmental qualification of equipment; and the steam generator tube rupture transients and associated radiological consequences, were performed. The increase in closure times and the changes to the MSFIS, MSIVs, and MFIVs either do not provide an adverse impact or do not result in accident acceptance criteria being challenged.

The modifications to the MSFIS controls will not affect any design basis accidents since the logic which currently exists will continue to be performed. The replacement controls are functionally the same as the current system since the same logic functions are performed, the same inputs received, and the same outputs produced.

The replacement of the MSFIS controls, replacement of the MSIV and MFIVs, and replacement of the electro-hydraulic actuators with system-medium actuators [with the longer closure time] will not result in a significant increase in the probability or consequence of an accident previously evaluated. [The replacement equipment for the MSFIS, MSIVs, and MFIVs does not reduce the reliability of the existing equipment being replaced.]

The relocation of the specific isolation times from the TSs to the TS Bases does not impact the design safety function of the valves to close. The TS requirements continue to provide the same level of assurance as before that the MSIVs and MFIVs are capable of performing their intended safety function. The addition of the MFRVs and MFRV bypass valves and extending the Completion Time for one or more MFIVs inoperable, is not an accident initiator and does not change the probability that an accident will occur. The increase in time that the MFIV is unavailable is small and the probability of an event occurring during this time period which would require isolation of the flow path is low. The redundancy provided by the MFRVs and MFRV bypass valves, which have the same actuation signals, provides adequate assurance that automatic feedwater isolation will occur.

Based on all of the above, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously analyzed.

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

Response: No.

The increase in MSIV and MFIV closure time as a result of the replacement of the MSFIS controls, MSIVs and MFIVs and associated actuators, will not prevent the Main Steam System, Main Feedwater System, or Auxiliary Feedwater System from performing their safety functions. The increased closure time will not affect the normal method of plant operation. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced with the proposed modifications and increased closure times. Although the modification does alter the design of the MSFIS and MSIV and MFIV actuators, it does not prevent the systems, subsystems, and components from performing their safety functions. [The replacement equipment for the MSFIS, MSIVs, and MFIVs are not initiators of accidents.]

The relocation of the specific isolation times from the TSs to the TS Bases and the addition of the MFRVs and MFRV bypass valves and extending the Completion Time for one or more MFIVs inoperable does not affect the assumptions of any accident analysis or the OPERABILITY of plant equipment.

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

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

Response: No.

The replacement of the MSFIS controls, replacement of the MSIVs and MFIVs and associated actuators and resulting increased closure time, does not affect the manner in which safety limits or limiting safety system settings are determined, nor will there be any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no significant impact on the overpower limit, departure from nucleate boiling ratio limits, heat flux hot channel factor, nuclear enthalpy rise hot channel factor, LOCA peak cladding temperature, peak local density, or any other margin of safety. The radiological dose consequence acceptance criteria listed in the Standard Review Plan will continue to be met.

Therefore, the proposed change[s] [do] not involve a significant reduction in the margin of safety.

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

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Thomas G. Hiltz.

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.

Florida Power and Light Company, Docket No. 50-250, Turkey Point Plant Unit 3, Miami-Dade County, Florida

Date of application for amendments: May 17, 2007.

Description of amendments request: The proposed amendment would allow the use of an alternate method of determining rod position for a control rod with inoperable rod position indication.

Date of publication of individual notice in the Federal Register: May 24, 2007 (72 FR 29186).

Expiration date of individual notice: June 25, 2007 (Public comments) and July 23, 2007 (Hearing requests).

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 Start Printed Page 33787made 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.

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

Date of application for amendment: June 30, 2006.

Brief description of amendment: The amendment revises the note preceding Technical Specification Surveillance Requirement 3.4.6.1 to be consistent with the wording in NUREG-1434, “Standard Technical Specifications for General Electric Plants, BWR/6,” Revision 3. Specifically, the note will be revised to read, “Not required to be performed in MODE 3.”

Date of issuance: May 24, 2007.

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

Amendment No.: 176.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: August 15, 2006 (71 FR 46930) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 13, 2006, as supplemented by letter dated March 6, 2007.

Brief description of amendments: The amendment revised the Millstone Power Station, Unit No. 2 (MPS2) Technical Specifications to modify the MPS2 licensing basis in the area of radiological dose analysis for design-basis accidents using the alternative source term permitted by Title 10 of the Code of Federal Regulations 50.67, “Accident source term”. Additionally, the amendment revises the MPS2 Technical Specifications consistent with the amended licensing-basis.

Date of issuance: May 31, 2007.

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

Amendment No: 298.

Facility Operating License Nos. DPR-65: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 29, 2006 (71 FR 51226). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 31, 2007.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut

Date of application for amendments: May 31, 2006, as supplemented by letters dated February 14 and April 26, 2005.

Brief description of amendments: The amendments revised the Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications (TSs) related to steam generator (SG) tube integrity. Specifically, the amendment revises the SG tube surveillance program consistent with the Nuclear Regulatory Commission-approved TS Task Force (TSTF) Standard TS Change Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4. TSTF-449 is part of the consolidated line item improvement process.

Date of issuance: May 31, 2007

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

Amendment Nos: 299 and 238

Facility Operating License Nos. DPR-65 and NPF-49: Amendments revised the TSs.

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75992).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 31, 2007. No significant hazards consideration comments received: No.

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

Date of application for amendment: February 8, 2007.

Brief description of amendment: The amendment modified Grand Gulf Nuclear Station, Unit 1 (GGNS) technical specifications (TSs) requirements for MODE change limitations in Limiting Condition of Operation (LCO) 3.0.4 and Surveillance Requirement (SR) 3.0.4. The TS changes are consistent with Revision 9 of NRC-approved Industry TS Task Force (TSTF) Standard TS Change Traveler, TSTF-359, “Increase Flexibility in MODE Restraints.” In addition, the amendment also changed TS Section 1.4, “Frequency,” Example 1.4-1, “Surveillance Requirements,” to accurately reflect the changes made by TSTF-359, which is consistent with NRC-approved TSTF-485, Revision 0, “Correct Example 1.4-1.”

Date of issuance: May 30, 2007.

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

Amendment No.: 175.

Facility Operating License No. NPF-29: The amendment revises the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : March 27, 2007 (72 FR 14304).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 25, 2006, as supplemented January 22, and April 16, 2007.

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

Date of Issuance: May 29, 2007.

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

Amendment No.: 147.

Renewed Facility Operating License No. NPF-16: Amendment revised the TSs. Start Printed Page 33788

Date of initial notice in Federal Register : July 18, 2006 (71 FR 40747). The January 22, and April 16, 2007, supplements did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 29, 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: December 13, 2006.

Brief description of amendment: The amendment deletes Technical Specification 6.8.1.3, which provided the requirement for submittal of the annual occupational radiation exposure report.

Date of issuance: May 25, 2007.

Effective date: May 25, 2007.

Amendment No.: 62.

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

Date of initial notice in Federal Register : February 13, 2007 (72 FR 6780)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated May 25, 2007.

No significant hazards consideration comments received: No.

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 application for amendments: November 21, 2005, supplemented by letters dated June 16, August 31, September 29, and October 30, 2006, March 15, and May 10, 2007.

Brief description of amendments: The amendments extend the Required Action Completion Times (CT) specified in technical specification (TS) 3.8.1, “AC Sources—Operating,” to restore an inoperable emergency diesel generator (EDG) to operable status from the current 7 days to 14 days. Specifically, the proposed changes would revise the current 7-day CT specified in TS 3.8.1 Required Action B.4 to allow 14 days to restore an inoperable EDG to operable status.

Date of issuance: May 30, 2007.

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

Amendment Nos.: 178 and 168.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 3, 2006 (71 FR 151).

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

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

No significant hazards consideration comments received: No.

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

Date of amendment request: December 20, 2006.

Brief description of amendment: The amendment deleted the Technical Specification requirements associated with the hydrogen purge system. The change is consistent with revisions of 10 CFR 50.44, “Combustible gas control for nuclear power reactors,” that became effective on October 16, 2003. This operating license improvement was made available by the U.S. Nuclear Regulatory Commission on September 25, 2003 (68 FR 55416) as part of the consolidated line item improvement process (CLIIP).

Date of issuance: June 6, 2007.

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

Amendment No.: 250.

Renewed Facility Operating License No. DPR-40: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register : January 30, 2007 (72 FR 4309)

The Commission's related evaluation of the amendment is contained in a safety evaluation dated June 6, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 7, 2006.

Brief description of amendments: The amendments revise TS 3.7.1, “Main Steam Safety Valves,” operability requirements and Linear Power Level High Trip setpoints.

Date of issuance: June 5, 2007.

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

Amendment Nos.: Unit 2-212; Unit 3-204.

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

Date of initial notice in Federal Register : December 19, 2006 (71 FR 75999).

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

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-328, Sequoyah Nuclear Plant, Unit 2, Hamilton County, Tennessee.

Date of application for amendments: February 15, 2006, as supplemented August 7, 2006, August 30, 2006, November 30, 2006, and April 2, 2007.

Brief description of amendments: The amendment revises the existing steam generator (SG) tube surveillance program through technical specification (TS) changes modeled after TS Task Force (TSTF) traveler TSTF-449, Revision 4, “Steam Generator Tube Integrity,” and the model safety evaluation prepared by the NRC and published in the Federal Register on March 2, 2005 (70 FR 10298). The amendment includes changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the SG tube surveillance program, changes to the SG reporting requirements, and associated changes to the TS Bases.

The amendment also deletes condition 2.C(8)(b) of Facility Operating License No. DPR-79.

This license condition references previous commitments for SG inspection that are bounded by the above TS changes.

Date of issuance: May 22, 2007.

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

Amendment No.: 305.

Facility Operating License No. DPR-79: Amendment revised the license and technical specifications.

Date of initial notice in Federal Register: March 28, 2006 (71 FR 15488).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 22, 2007.

No significant hazards consideration comments received: No. Start Printed Page 33789

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, 2006, as supplemented on February 23, June 21, and July 28, 2006.

Brief Description of amendments: These amendments revised the Technical Specifications to incorporate the changes to the operation of the containment, as discussed in Generic Letter 2004-02, “Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized-Water Reactor,” dated September 13, 2004.

Date of issuance: October 12, 2006.

Effective date: Unit 1 (fall 2007 refueling outage) and Unit 2 (fall 2006 refueling outage).

Amendment Nos.: 250 and 249.

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

Date of initial notice in Federal Register : March 14, 2006 (71 FR 13182).

The February 23, June 21, and July 28, 2006, supplements contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 12, 2006.

No significant hazards consideration comments received: No.

Start Signature

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

For The Nuclear Regulatory Commission.

Timothy J. McGinty,

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

End Signature End Preamble

[FR Doc. E7-11567 Filed 6-18-07; 8:45 am]

BILLING CODE 7590-01-P