Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 33779-33789 [E7-11567]

Download as PDF Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations rwilkins on PROD1PC63 with NOTICES 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 60day 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 PO 00000 Frm 00045 Fmt 4703 Sfmt 4703 33779 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 E:\FR\FM\19JNN1.SGM 19JNN1 rwilkins on PROD1PC63 with NOTICES 33780 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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- VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 PO 00000 Frm 00046 Fmt 4703 Sfmt 4703 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. E:\FR\FM\19JNN1.SGM 19JNN1 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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 rwilkins on PROD1PC63 with NOTICES 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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-failureproof 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: PO 00000 Frm 00047 Fmt 4703 Sfmt 4703 33781 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 E:\FR\FM\19JNN1.SGM 19JNN1 rwilkins on PROD1PC63 with NOTICES 33782 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices load 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. VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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. PO 00000 Frm 00048 Fmt 4703 Sfmt 4703 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. E:\FR\FM\19JNN1.SGM 19JNN1 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices rwilkins on PROD1PC63 with NOTICES 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 PO 00000 Frm 00049 Fmt 4703 Sfmt 4703 33783 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 E:\FR\FM\19JNN1.SGM 19JNN1 33784 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices rwilkins on PROD1PC63 with NOTICES Specifications (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? VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 PO 00000 Frm 00050 Fmt 4703 Sfmt 4703 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 E:\FR\FM\19JNN1.SGM 19JNN1 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety. rwilkins on PROD1PC63 with NOTICES 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 PO 00000 Frm 00051 Fmt 4703 Sfmt 4703 33785 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. E:\FR\FM\19JNN1.SGM 19JNN1 rwilkins on PROD1PC63 with NOTICES 33786 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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-ofcoolant accident] transients; SBLOCA [smallbreak 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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 PO 00000 Frm 00052 Fmt 4703 Sfmt 4703 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 E:\FR\FM\19JNN1.SGM 19JNN1 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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. rwilkins on PROD1PC63 with NOTICES 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 VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 Specifications to modify the MPS2 licensing basis in the area of radiological dose analysis for designbasis 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. PO 00000 Frm 00053 Fmt 4703 Sfmt 4703 33787 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 NRCapproved 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. E:\FR\FM\19JNN1.SGM 19JNN1 33788 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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. rwilkins on PROD1PC63 with NOTICES 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. VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 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. 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. 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. Tennessee Valley Authority, Docket Nos. 50–328, Sequoyah Nuclear Plant, Unit 2, Hamilton County, Tennessee. 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. PO 00000 Frm 00054 Fmt 4703 Sfmt 4703 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. E:\FR\FM\19JNN1.SGM 19JNN1 Federal Register / Vol. 72, No. 117 / Tuesday, June 19, 2007 / Notices 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. 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. [FR Doc. E7–11567 Filed 6–18–07; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [Docket No. 50–400 License No. NPF–63] rwilkins on PROD1PC63 with NOTICES Carolina Power & Light Company; Notice of Issuance of Director’s Decision Under 10 CFR 2.206 Notice is hereby given that the Director of the Office of Nuclear Reactor Regulation has issued a director’s decision with regard to a petition dated September 20, 2006, filed by Mr. John D. Runkle, attorney for North Carolina Waste Awareness and Reduction Network and numerous other VerDate Aug<31>2005 18:32 Jun 18, 2007 Jkt 211001 organizations, hereinafter referred to as the ‘‘Petitioners.’’ The petition was supplemented by documents dated September 21, October 30, November 29, 2006, and February 8, 2007. The petition concerns longstanding fire protection issues at the Shearon Harris Nuclear Power Plant (SHNPP or the Licensee). The Petitioners requested that the Nuclear Regulatory Commission (NRC) staff take enforcement action in the form of an order that would revoke SHNPP’s operating license or impose maximum fines for each violation for each day the plant has been in violation of fire protection regulations. As the basis for this request, the Petitioners discussed several fire safety issues at SHNPP that they believe could affect the safe operation of the plant and safe shutdown of the plant in emergency situations. The Petitioners’ concerns focused on noncompliances, the risk associated with the noncompliances, reliance on compensatory measures, the NRC’s policy on the use of enforcement discretion regarding certain fire protection issues, and intentional acts of sabotage or terrorism. On November 13, 2006, the NRC conducted a public meeting at NRC headquarters regarding fire protection issues at SHNPP. The meeting gave the Petitioners and the SHNPP Licensee an opportunity to provide additional information to the NRC’s Petition Review Board and to clarify issues raised in the petition. The NRC staff sent a copy of the proposed Director’s Decision to the Petitioners and to the SHNPP Licensee for comment by letters dated April 2, 2007. The Petitioners and the Licensee submitted comments by letters dated May 1, 2007, and these comments are addressed in the final Director’s Decision. The Director of the Office of Nuclear Reactor Regulation has determined that the requests to revoke SHNPP’s Operating License or impose maximum fines for each violation for each day the plant has been in violation of fire protection regulations are denied. The reasons for this decision are explained in the Director’s Decision pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 2.206 (DD–07–03), the complete text of which is available in ADAMS for inspection at the Commission’s Public Document Room, located at One White Flint North, Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and from the ADAMS Public Library component on the NRC’s Web site, http://www.nrc.gov/readingrm.html (the Public Electronic Reading PO 00000 Frm 00055 Fmt 4703 Sfmt 4703 33789 Room) using Accession Number ML071490145. In summary, the Director’s Decision denies the Petitioners’ requests due to the determination by the NRC staff that the plant may continue operation and the Licensee’s efforts to transition to the risk-informed, performance-based standards in 10 CFR 50.48(c). In addition, the Licensee is actively identifying and completing corrective actions, including plant modifications and reanalysis efforts associated with meeting the new standards in 10 CFR 50.48(c), and has in place compensatory measures to account for existing noncompliances. The Licensee continues to have available several levels of defense-in-depth in fire protection. The Licensee has been granted enforcement discretion under the NRC’s ‘‘Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48(c)).’’ The NRC has followed and continues to follow existing regulatory processes, policies and programs to verify that the Licensee is properly implementing its fire protection program at SHNPP in accordance with NRC rules and regulations. A copy of the director’s decision will be filed with the Secretary of the Commission for the Commission’s review in accordance with 10 CFR 2.206 of the Commission’s regulations. As provided for by this regulation, the director’s decision will constitute the final action of the Commission 25 days after the date of the decision, unless the Commission, on its own motion, institutes a review of the director’s decision in that time. Dated at Rockville, Maryland, this 13 day of June, 2007. For the Nuclear Regulatory Commission. James T. Wiggins, Acting Director, Office of Nuclear Reactor Regulation. [FR Doc. E7–11814 Filed 6–18–07; 8:45 am] BILLING CODE 7590–01–P OVERSEAS PRIVATE INVESTMENT CORPORATION Submission of OMB Review; Comments Request Overseas Private Investment Corporation (OPIC). ACTION: Request for comments. AGENCY: SUMMARY: Under the provisions of the Paperwork Reduction Act (44 U.S.C. Chapter 35), agencies are required to publish a Notice in the Federal Register notifying the public that the Agency has prepared an information collection E:\FR\FM\19JNN1.SGM 19JNN1

Agencies

[Federal Register Volume 72, Number 117 (Tuesday, June 19, 2007)]
[Notices]
[Pages 33779-33789]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-11567]



[[Page 33779]]

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NUCLEAR REGULATORY COMMISSION


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

I. Background

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

[[Page 33780]]

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.

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.

[[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

[[Page 33782]]

load 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.

[[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

[[Page 33784]]

Specifications (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

[[Page 33785]]

functioning during accident conditions as assumed in the licensing 
basis analyses of design basis accident radiological consequences to 
CRE occupants. No new or different accidents result from performing 
the new surveillance or following the new program. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a 
significant change in the methods governing normal plant operation. 
The proposed change does not alter any safety analysis assumptions 
and is consistent with current plant operating practice. Therefore, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

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

[[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 sp