Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 4304-4312 [E7-1259]

Download as PDF 4304 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices (Tentative). d. System Energy Resources, Inc. (Early Site Permit for Grand Gulf ESP Site); response to NEPA/ terrorism issue (Tentative). Tuesday, January 30, 2007 10 a.m. Discussion of Security Issues (Closed-Ex. 3). 1:30 p.m. Discussion of Security Issues (Closed-Ex. 1). Thursday, February 1, 2007 9:25 a.m. Affirmation Session (Public Meeting) (Tentative) a. USEC, Inc. (American Centrifuge Plant) (Tentative). 9:30 a.m. Discussion of Management Issues (Closed-Ex. 2). 1:30 p.m. Briefing on Strategic Workforce Planning and Human Capital Initiatives (Public Meeting) (Contact: Mary Ellen Beach, 301 415– 6803). This meeting will be webcast live at the Web address—www.nrc.gov Week of February 5, 2007—Tentative There are no meetings scheduled for the Week of February 5, 2007. Week of February 12, 2007—Tentative Thursday, February 15, 2007 9:30 a.m. Briefing on Office of Chief Financial Officer (OCFO) Programs, Performance, and Plans (Public Meeting) (Contact: Edward New, 301– 415–5646). This meeting will be webcast live at the Web address—www.nrc.gov. Week of February 19, 2007—Tentative There are no meetings scheduled for the Week of February 19, 2007. Week of February 26, 2007—Tentative Wednesday, February 28, 2007 9:30 a.m. Periodic Briefing on New Reactor Issues (Public Meeting) (Contact: Donna Williams, 301–415– 1322). This meeting will be webcast live at the Web address—www.nrc.gov. Week of March 5, 2007—Tentative ycherry on PROD1PC64 with NOTICES Monday, March 5, 2007 1 p.m. Meeting with Department of Energy on New Reactor Issues (Public Meeting). This meeting will be webcast live at the Web address—www.nrc.gov. Tuesday, March 6, 2007 1 p.m. Discussion of Management Issues (Closed-Ex. 2) (Tentative). Wednesday, March 7, 2007 9:30 a.m. Briefing on Office of Nuclear Security and Incident Response VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 (NSIR) Programs, Performance, and Plans (Public Meeting). This meeting will be webcast live at the Web address—www.nrc.gov. 1 p.m. Discussion of Security Issues (Closed-Ex. 1 and 3). Thursday, March 8, 2007 10 a.m. Briefing on Nuclear Materials Safety and Safeguards (NMSS) Programs, Performance, and Plans (Public Meeting). This meeting will be webcast live at the Web address—www.nrc.gov. 1 p.m. Briefing on Office of Nuclear Reactor Regulation (NRR) Programs, Performance, and Plans (Public Meeting). This meeting will be webcast live at the Web address—www.nrc.gov. * * * * * *The schedule for Commission meetings is subject to change on short notice. To verify the status of meetings call (recording)—(301) 415–1292. Contact person for more information: Michelle Schroll, (301) 415–1662. * * * * * Additional Information Affirmation of ‘‘Pacific Gas & Electric Co. (Diablo Canyon ISFSI), Docket No. 72–26–ISFSI, response to the Supreme Court’s potential denial of certiorari’’ tentatively scheduled on Monday, January 29, 2007, at 10:50 a.m. has been postponed and will be rescheduled. ‘‘Discussion of Security Issues (Closed-Ex. 1 & 3)’’ previously scheduled on Wednesday, January 31, 2007, at 9:30 a.m. has been postponed and will be rescheduled. * * * * * The NRC Commission Meeting Schedule can be found on the Internet at: www.nrc.gov/what-we-do/policymaking/schedule.html. * * * * * The NRC provides reasonable accommodation to individuals with disabilities where appropriate. If you need a reasonable accommodation to participate in these public meetings, or need this meeting notice or the transcript or other information from the public meetings in another format (e.g. braille, large print), please notify the NRC’s Disability Program Coordinator, Deborah Chan, at 301–415–7041, TDD: 301–415–2100, or by e-mail at DLC@nrc.gov. Determinations on requests for reasonable accommodation will be made on a case-by-case basis. * * * * * This notice is distributed by mail to several hundred subscribers; if you no longer wish to receive it, or would like PO 00000 Frm 00074 Fmt 4703 Sfmt 4703 to be added to the distribution, please contact the Office of the Secretary, Washington, D.C. 20555 (301–415– 1969). In addition, distribution of this meeting notice over the Internet system is available. If you are interested in receiving this Commission meeting schedule electronically, please send an electronic message to dkw@nrc.gov. Dated: January 25, 2007. R. Michelle Schroll, Office of the Secretary. [FR Doc. 07–415 Filed 1–26–07; 1:50 pm] BILLING CODE 7590–01–P 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 January 5, 2007 to January 18, 2007. The last biweekly notice was published on January 16, 2007 (72 FR 1779). 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) E:\FR\FM\30JAN1.SGM 30JAN1 ycherry on PROD1PC64 with NOTICES Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices 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 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 VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 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 PO 00000 Frm 00075 Fmt 4703 Sfmt 4703 4305 fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/ requestor to relief. A petitioner/ requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and E:\FR\FM\30JAN1.SGM 30JAN1 4306 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices ycherry on PROD1PC64 with NOTICES 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 email 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. AmerGen Energy Company, LLC, Docket No. 50–461, Clinton Power Station, Unit 1, DeWitt County, Illinois Date of amendment request: November 16, 2006. Description of amendment request: The proposed amendment would revise Technical Specification (TS) Section 3.6.5.1, ‘‘Drywell,’’ Surveillance Requirement (SR) 3.6.5.1.3 to delay the performance of the next drywell bypass leakage rate test (DBLRT) from the current requirement of ‘‘November 23, 2008’’ to ‘‘prior to startup from the C1R12 refueling outage’’ which is currently scheduled for January 2010. This request would also revise TS 5.5.13, ‘‘Primary Containment Leakage Rate Testing Program,’’ to delay the VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 performance of the next primary containment Type A integrated leak rate test (ILTR) from the current requirement of ‘‘no later than November 23, 2008’’ to ‘‘prior to startup from the C1R12 refueling outage.’’ Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change will revise TS 3.6.5.1, ‘‘Drywell,’’ SR 3.6.5.1 .3 to defer the performance of the next DBLRT to prior to startup from the C1R12 refueling outage. This request will also revise CPS TS 5.5.13, ‘‘Primary Containment Leakage Rate Testing Program,’’ to reflect a one-time deferral of the primary containment Type A test to prior to startup from the C1R12 refueling outage. The current Type A test and DBLRT interval of 15 years, based on past performance, would be extended on a onetime basis to 16.25 years (i.e., approximately 15 years plus 15 months) from the last Type A test and DBLRT. The drywell houses the reactor pressure vessel, the reactor coolant recirculation loops, and branch connections of the Reactor Coolant System (RCS), which have isolation valves at the primary containment boundary. The function of the drywell is to maintain a pressure boundary that channels steam resulting from a Loss of Coolant Accident (LOCA) to the suppression pool, where it is condensed. Air forced from the drywell is released into the primary containment through the suppression pool. The suppression pool is a concentric open container of water with a stainless steel liner that is located at the bottom of the primary containment. The suppression pool is designed to absorb the decay heat and sensible heat released during a reactor blowdown from safety/relief valve (SRV) discharges or from a LOCA. The function of the Mark III containment is to isolate and contain fission products released from the RCS following a design basis LOCA and to confine the postulated release of radioactive material to within limits. The test interval associated with the drywell bypass leakage and Type A testing is not a precursor of any accident previously evaluated. Therefore, extending these test intervals on a one-time basis from 15 years to 16.25 years does not result in an increase in the probability of occurrence of an accident. The successful performance history of the drywell bypass leakage and Type A testing provides assurance that the CPS drywell and primary containment will not exceed allowable leakage rate values specified in the TS and will continue to perform its design function following an accident. The risk assessment of the proposed changes has concluded that there is an insignificant increase in total population dose rate and an insignificant increase in the conditional containment failure probability. PO 00000 Frm 00076 Fmt 4703 Sfmt 4703 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes for a one-time extension of the DBLRT and Type A test will not affect the control parameters governing unit operations or the response of plant equipment to transient and accident conditions. The proposed changes do not introduce any new equipment or modes of system operation. No installed equipment will be operated in a new or different manner. As such, no new failure mechanisms are introduced. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. CPS is a General Electric BWR/6 plant with a Mark III containment system. The Mark III containment design is a single-barrier pressure containment and a multi-barrier fission containment system consisting of the drywell and primary containment. The drywell houses the reactor pressure vessel, the reactor coolant recirculation loops, and branch connections of the RCS, which have isolation valves at the primary containment boundary. The function of the drywell is to maintain a pressure boundary that channels steam from a LOCA to the suppression pool, where it is condensed. The suppression pool is an annular pool of demineralized water between the drywell and the outer primary containment boundary. This pool covers the horizontal vent openings in the drywell to maintain a water seal between the drywell interior and the remainder of the containment volume. The primary containment consists of a steel-lined, reinforced concrete vessel, which surrounds the RCS and provides an essentially leaktight barrier against an uncontrolled release of radioactive material to the environment. Additionally, the containment structure provides shielding from the fission products that may be present in the primary containment atmosphere following accident conditions. The primary containment is penetrated by access, piping and electrical penetrations. The integrity of the drywell is periodically verified by performance of the DBLRT. This test ensures that the measured drywell bypass leakage is bounded by the safety analysis assumptions. The drywell integrity is further verified by a number of additional tests, including drywell airlock door seal leakage tests, overall drywell airlock leakage tests and periodic visual inspections of exposed accessible interior and exterior drywell surfaces. Additional confidence that significant degradation in the drywell leaktightness has not developed is provided by the periodic qualitative assessment of drywell performance. The integrity of the primary containment penetrations and isolation valves is verified E:\FR\FM\30JAN1.SGM 30JAN1 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices through Type B and Type C local leak rate tests (LLRTs) and the overall leak-tight integrity of the primary containment is verified by a Type A integrated leak rate test (ILRT) as required by 10 CFR 50, Appendix J. These tests are performed to verify the essentially leak-tight characteristics of the primary containment at the design basis accident pressure. The proposed changes for a one-time extension of the drywell bypass leakage and Type A tests do not affect the method for drywell or containment testing or the test acceptance criteria. AmerGen has conducted a risk assessment to determine the impact of a change to the CPS Type A ILRT and DBLRT schedule from the originally licensed baseline frequency of three tests in 10 years to one test in 15 years plus 15 months (i.e., approximately 16.25 years) for the risk measures of Large Early Release Frequency (i.e., LERF), Population Dose, and Conditional Containment Failure Probability (i.e., CCFP). This assessment indicated that the proposed CPS interval extension has a small change in risk to the public and is an acceptable plant change from a risk perspective. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, AmerGen concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, ‘‘Issuance of amendment,’’ paragraph (c), and, accordingly, a finding of ‘‘no significant hazards consideration’’ is justified. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. NRC Branch Chief: Michael L. Marshall, Jr. ycherry on PROD1PC64 with NOTICES Entergy Nuclear Operations, Inc., Docket No. 50–293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of amendment request: November 2, 2006. Description of amendment request: The proposed amendment would modify requirements for inoperable snubbers consistent with the Technical Specification Task Force 372, Revision 4. 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: VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 Criterion 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 inoperable snubber if risk is assessed and managed. The postulated seismic event requiring snubbers is a lowprobability 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 [limiting condition for operation] 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. Criterion 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. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in [a] 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 [Regulatory Guide] 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. PO 00000 Frm 00077 Fmt 4703 Sfmt 4703 4307 The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360–5599 NRC Branch Chief: Richard Laufer. Exelon Generation Company, LLC, Docket Nos. 50–254 and 50–265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: November 7, 2006. Description of amendment request: The proposed change revises Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1 to increase the allowable as-found main steam safety valve (MSSV) lift setpoint tolerance from ± 1 percent to ± 3 percent. In addition, the proposed change revises SR 3.1.7.10 to increase the enrichment of sodium pentaborate used in the Standby Liquid Control (SLC) system from ≥ 30.0 atom percent boron-10 to ≥ 45.0 atom percent boron-10. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change increases the allowable as-found MSSV lift setpoint tolerance, determined by test after the valves have been removed from service, from ± 1% to ± 3%. The proposed change does not alter the TS requirements for the number of MSSVs required to be operable, the nominal lift setpoints, the allowable as-left lift setpoint tolerance, the MSSV testing frequency, or the manner in which the valves are operated. Consistent with current TS requirements, the proposed change continues to require that the MSSVs be adjusted to within ± 1% of their nominal lift setpoints following testing. Since the proposed change does not alter the manner in which the valves are operated, there is no significant impact on reactor operation. The proposed change does not involve a physical change to the valves, nor does it change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating E:\FR\FM\30JAN1.SGM 30JAN1 ycherry on PROD1PC64 with NOTICES 4308 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices conditions and no changes to existing structures, systems, or components, with the exception of the SLC system enrichment change. The proposed change to increase the enrichment of sodium pentaborate used in the SLC system will ensure that the requirements of 10 CFR 50.62, ‘‘Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants,’’ continue to be met. The SLC system is not an initiator to an accident; rather, the SLC system is used to mitigate an ATWS event. Therefore, these changes will not increase the probability of an accident previously evaluated. Generic considerations related to the change in setpoint tolerance were addressed in NEDC–3175310, ‘‘BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report,’’ and were reviewed and approved by the NRC in a safety evaluation dated March 8, 1993. General Electric Company (GE) completed plant-specific analyses to assess the impact of the setpoint tolerance increase on Dresden Nuclear Power Station Units 2 and 3 and QCNPS [Quad Cities Nuclear Power Station] Units 1 and 2. The impact of the MSSV setpoint tolerance increase, as addressed in this analysis, included vessel overpressure, Updated Final Safety Analysis Report (UFSAR) Chapter 15 events, ATWS, Loss of Coolant Accident (LOCA), containment response and loads, high pressure systems performance, Appendix R fire protection, vessel thermal cycle, operating mode and equipment out of service review, and extended power uprate evaluation review. The proposed change to 3% setpoint tolerance is supported by Westinghouse SVEA–96 Optimal fuel analysis of events that credit the MSSVs. The plant specific evaluations, required by the NRC’s safety evaluation and performed to support this proposed change, show that there is no change to the design core thermal limits and adequate margin to the reactor vessel pressure limits using a ±3% lift setpoint tolerance. These analyses also show that operation of Emergency Core Cooling Systems is not affected, and the containment response following a LOCA is acceptable. The plant systems associated with these proposed changes are capable of meeting applicable design basis requirements and retain the capability to mitigate the consequences of accidents described in the UFSAR. The accident analyses that credit the initiation of SLC as a dose mitigation feature are not impacted by the proposed change because the chemical properties of the SLC boron solution are not affected. Therefore, these changes do not involve an increase in the consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change increases the allowable as-found lift setpoint tolerance for VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 the QCNPS MSSVs, and increases the required enrichment of sodium pentaborate used in the SLC system. The proposed change to increase the enrichment of sodium pentaborate used in the SLC system will ensure that the requirements of 10 CFR 50.62 continue to be met. The proposed change to increase the MSSV tolerance was developed in accordance with the provisions contained in the NRC safety evaluation for NEDC–31753P. MSSVs installed in the plant following testing or refurbishment will continue to meet the current tolerance acceptance criteria of ± 1% of the nominal setpoint. The proposed change does not affect the manner in which the overpressure protection system is operated; therefore, there are no new failure mechanisms for the overpressure protection system. The proposed change to allow an increase in the MSSV setpoint tolerance does not alter the nominal MSSV lift setpoints or the number of MSSVs currently required to be operable by QCNPS TS. The proposed change does not involve physical changes to the valves, nor does it change the safety function of the valves. There is no alteration to the parameters within which the plant is normally operated. As a result, no new failure modes are being introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. Establishment of the ± 3% MSSV setpoint tolerance limit does not adversely impact the operation of any safety-related component or equipment. Evaluations performed in accordance with the NRC safety evaluation for NEDC–31753P have concluded that all design limits will continue to be met. The proposed change to increase the enrichment of sodium pentaborate used in the SLC system will ensure that the requirements of 10 CFR 50.62 continue to be met. Therefore, the proposed change does not involve a significant reduction in the margin of safety. Based upon the above, EGC [Exelon Generation Company] concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and, accordingly, a finding of no significant hazards consideration is justified. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three PO 00000 Frm 00078 Fmt 4703 Sfmt 4703 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. NRC Branch Chief: Michael L. Marshall, Jr. Omaha Public Power District, Docket No. 50–285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: December 20, 2006. Description of amendment request: The proposed amendment would remove annotations referencing Technical Data Book (TDB)–VIII, ‘‘Equipment Operability Guidance,’’ and annotations referencing Technical Specification Interpretations (TSIs) from the NRC Authority File. These documents are used by Omaha Public Power District (OPPD) personnel for additional guidance in applying certain Limiting Conditions for Operation requirements to specific equipment and/ or situations. OPPD has annotated references to these documents in the Technical Specification (TS) copies used at Fort Calhoun Station (FCS); however, the annotations are ‘‘pointers’’ to additional guidance and are not officially a part of the FCS TS. The proposed amendment also corrects an administrative discrepancy in TS 2.10.4(1)(c). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The correction of administrative discrepancies in the Fort Calhoun Station (FCS) Technical Specifications (TS) is not an initiator of any previously evaluated accident. The proposed changes will not prevent safety systems from performing their accident mitigation function as assumed in the safety analysis. Therefore, this change does not involve an increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes only affect the Technical Specifications and do not involve a physical change to the plant. Modifications E:\FR\FM\30JAN1.SGM 30JAN1 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices will not be made to existing components nor will any new or different types of equipment be installed. This change will not alter assumptions made in safety analysis and licensing bases. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The correction of administrative discrepancies in the Technical Specifications has no impact on any safety analysis assumptions and thus this TS change does not involve a reduction in a margin of safety. ycherry on PROD1PC64 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 request involves no significant hazards consideration. Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006– 3817. NRC Branch Chief: David Terao. Omaha Public Power District, Docket No. 50–285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: December 20, 2006. Description of amendment request: The proposed amendment would delete the Technical Specification (TS) requirements related to the hydrogen purge system in TS 2.6(3) and TS Table 3–5, Item 17. The proposed TS changes support implementation of the revisions to 10 CFR 50.44, ‘‘Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,’’ that became effective on September 16, 2003. The changes are consistent with Revision 1 of NRC-approved Industry/ Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF– 447, ‘‘Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors.’’ The NRC staff issued a notice of opportunity to comment in the Federal Register dated August 2, 2002 (67 FR 50374), on possible amendments for the elimination of requirements for hydrogen recombiners, and hydrogen and oxygen monitors from the TSs, 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 September 25, VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 2003 (68 FR 55416). The licensee affirmed the applicability of the NSHC in its application dated December 20, 2006. 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 revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity. With the elimination of the design-basis LOCA hydrogen release, hydrogen [and oxygen] monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen [and oxygen] monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. [Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment.] The regulatory requirements for the hydrogen [and oxygen] monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, [classification of the oxygen monitors as Category 2] and removal of the hydrogen [and oxygen] monitors from TS will not prevent an accident management strategy through the PO 00000 Frm 00079 Fmt 4703 Sfmt 4703 4309 use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs). Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen [and oxygen] monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated. The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen [and oxygen] monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen [and oxygen] monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen [and oxygen] monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety. The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen [and oxygen] monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a designbasis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Therefore, this change does not involve a significant reduction in the margin of safety. [The intent of the requirements established as a result of the TMI, Unit 2 accident can be E:\FR\FM\30JAN1.SGM 30JAN1 4310 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices adequately met without reliance on safetyrelated oxygen monitors.] Removal of hydrogen [and oxygen] monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability. Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006– 3817. NRC Branch Chief: David Terao. ycherry on PROD1PC64 with NOTICES Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action, see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 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. 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: December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006. Brief description of amendments: The amendments revised the San Onofre Nuclear Generating Station, Units 2 and 3, accident source term used in the design-basis radiological consequence analyses. The amendments were in accordance with the requirements of 10 CFR 50.67, which addresses the use of an alternative source term (AST) at operating reactors, and relevant guidance of Regulatory Guide (RG) 1.183. The amendments represent fullscope implementation of the AST described in RG 1.183. Date of issuance: December 29, 2006. Effective date: As of the date of issuance and shall be implemented within 180 days of issuance. Amendment Nos.: Unit 2—210; Unit 3—202. Facility Operating License Nos. NPF– 10 and NPF–15: The amendments revised the Updated Final Safety Analysis Report. Date of initial notice in Federal Register: February 1, 2005 (70 FR 5248). The supplemental letters dated October 27, 2005, March 10, and October 6, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated December 29, 2006. No significant hazards consideration comments received: No. PO 00000 Frm 00080 Fmt 4703 Sfmt 4703 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: March 10, 2006, as supplemented by submittal dated May 16, 2006. Brief description of amendments: The amendments conform the Facility Operating Licenses NPF–10 and NPF–15 for the San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3) to reflect their transfer from the City of Anaheim (Anaheim) to Southern California Edison (SCE). The license transfers, which were approved by the Order dated September 27, 2006, permitted the transfer of the 3.16percent undivided ownership interest in the facilities held by Anaheim to SCE, excluding Anaheim’s interest in its spent fuel and in the SONGS 2 and 3 independent spent fuel storage installation. SCE retains exclusive responsibility and control over the operation of SONGS 2 and 3. Date of issuance: December 29, 2006. Effective date: At the time the transfer is completed. Amendment Nos.: Unit 2—209; Unit 3—201. Facility Operating License Nos. NPF– 10 and NPF–15: The amendments revised the Facility Operating Licenses. Date of initial notice in Federal Register: June 8, 2006 (71 FR 33321) The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 2006. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50–259 Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama Date of application for amendment: January 6, 2006 (TS–443), as supplemented by letter dated October 2, 2006. Brief description of amendment: Activation of thermal-hydraulic stability monitoring instrumentation. The Oscillation Power Range Monitor System is designed to provide the licensee’s solution regarding reactor stability. Date of issuance: December 29, 2006. Effective date: Date of issuance, to be implemented within 60 days. Amendment No.: 266. Renewed Facility Operating License No. DPR–33: Amendment revised the TSs. Date of initial notice in Federal Register: April 5, 2006 (71 FR 23962). The October 2, 2006, supplement, contained clarifying information and E:\FR\FM\30JAN1.SGM 30JAN1 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices ycherry on PROD1PC64 with NOTICES did not change the NRC staff’s initial proposed finding of no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated December 29, 2006. No significant hazards consideration comments received: No. Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances) 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 for the amendment 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. Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing. For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee’s facility of the licensee’s application and of the Commission’s proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments. In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant’s licensed power level, the Commission may not have had an VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible. Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved. The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, 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 System’s (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, PO 00000 Frm 00081 Fmt 4703 Sfmt 4703 4311 (301) 415–4737 or by e-mail to pdr@nrc.gov. The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737, or by email to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, 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 identify the specific contentions which the petitioner/ E:\FR\FM\30JAN1.SGM 30JAN1 4312 Federal Register / Vol. 72, No. 19 / Tuesday, January 30, 2007 / Notices ycherry on PROD1PC64 with NOTICES 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 intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.1 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 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. Each contention shall be given a separate numeric or alpha designation within one of the following groups: 1. Technical—primarily concerns/ issues relating to technical and/or health and safety matters discussed or referenced in the applications. 2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications. 3. Miscellaneous—does not fall into one of the categories outlined above. As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/ requestors shall jointly designate a representative who shall have the authority to act for the petitioners/ requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the 1 To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant’s counsel and discuss the need for a protective order. VerDate Aug<31>2005 15:36 Jan 29, 2007 Jkt 211001 authority to act for the petitioners/ requestors with respect to that contention. 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect. 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 email 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 or 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). STP Nuclear Operating Company, Docket No. 50–498, South Texas Project, Unit 1, Matagorda County, Texas Date of amendment request: December 20, 2006, as supplemented by letter dated December 28, 2006. Description of amendment request: The amendment, for a one-time change, PO 00000 Frm 00082 Fmt 4703 Sfmt 4703 revised Technical Specification (TS) 3.3.2 for the loss of power (LOP) instrumentation (Functional Unit 8, ‘‘loss of power’’) in TS Table 3.3–3, ‘‘Engineered Safety Features Actuation System Instrumentation.’’ A note is added to TS Table 3.3–3, Action 20, which is the TS-required action for inoperable LOP instrumentation, to allow a one-time provision for corrective maintenance on an inoperable Unit 1 LOP instrumentation channel when the number of operable channels are more than one less than the total number of channels. This provision for corrective maintenance expires 30 days after the amendment is approved. Date of issuance: January 11, 2007. Effective date: Effective as of its date of issuance and shall be implemented by January 15, 2007. Amendment No.: 176. Facility Operating License No. NPF– 76: The amendment revised the Technical Specifications and Facility Operating License. Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission’s related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated January 11, 2007. Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004. NRC Branch Chief: David Terao. Dated at Rockville, Maryland, this 22nd day of January 2007. For the Nuclear Regulatory Commission. John W. Lubinski, Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. E7–1259 Filed 1–29–07; 8:45 am] BILLING CODE 7590–01–P SECURITIES AND EXCHANGE COMMISSION [Release Number IC–27677; File No. 812– 13321] Integrity Life Insurance Company, et al. January 24, 2007. Securities and Exchange Commission (the ‘‘Commission’’). ACTION: Notice of application for an order of approval pursuant to Section 26(c) of the Investment Company Act of 1940, as amended (the ‘‘Act’’). AGENCY: Integrity Life Insurance Company (‘‘Integrity’’), Separate APPLICANTS: E:\FR\FM\30JAN1.SGM 30JAN1

Agencies

[Federal Register Volume 72, Number 19 (Tuesday, January 30, 2007)]
[Notices]
[Pages 4304-4312]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-1259]


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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 January 5, 2007 to January 18, 2007. The 
last biweekly notice was published on January 16, 2007 (72 FR 1779).

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)

[[Page 4305]]

involve a significant reduction in a margin of safety. The basis for 
this proposed determination for each amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and

[[Page 4306]]

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.

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

    Date of amendment request: November 16, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.6.5.1, ``Drywell,'' 
Surveillance Requirement (SR) 3.6.5.1.3 to delay the performance of the 
next drywell bypass leakage rate test (DBLRT) from the current 
requirement of ``November 23, 2008'' to ``prior to startup from the 
C1R12 refueling outage'' which is currently scheduled for January 2010. 
This request would also revise TS 5.5.13, ``Primary Containment Leakage 
Rate Testing Program,'' to delay the performance of the next primary 
containment Type A integrated leak rate test (ILTR) from the current 
requirement of ``no later than November 23, 2008'' to ``prior to 
startup from the C1R12 refueling outage.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise TS 3.6.5.1, ``Drywell,'' SR 
3.6.5.1 .3 to defer the performance of the next DBLRT to prior to 
startup from the C1R12 refueling outage. This request will also 
revise CPS TS 5.5.13, ``Primary Containment Leakage Rate Testing 
Program,'' to reflect a one-time deferral of the primary containment 
Type A test to prior to startup from the C1R12 refueling outage. The 
current Type A test and DBLRT interval of 15 years, based on past 
performance, would be extended on a onetime basis to 16.25 years 
(i.e., approximately 15 years plus 15 months) from the last Type A 
test and DBLRT.
    The drywell houses the reactor pressure vessel, the reactor 
coolant recirculation loops, and branch connections of the Reactor 
Coolant System (RCS), which have isolation valves at the primary 
containment boundary. The function of the drywell is to maintain a 
pressure boundary that channels steam resulting from a Loss of 
Coolant Accident (LOCA) to the suppression pool, where it is 
condensed. Air forced from the drywell is released into the primary 
containment through the suppression pool. The suppression pool is a 
concentric open container of water with a stainless steel liner that 
is located at the bottom of the primary containment. The suppression 
pool is designed to absorb the decay heat and sensible heat released 
during a reactor blowdown from safety/relief valve (SRV) discharges 
or from a LOCA.
    The function of the Mark III containment is to isolate and 
contain fission products released from the RCS following a design 
basis LOCA and to confine the postulated release of radioactive 
material to within limits. The test interval associated with the 
drywell bypass leakage and Type A testing is not a precursor of any 
accident previously evaluated. Therefore, extending these test 
intervals on a one-time basis from 15 years to 16.25 years does not 
result in an increase in the probability of occurrence of an 
accident. The successful performance history of the drywell bypass 
leakage and Type A testing provides assurance that the CPS drywell 
and primary containment will not exceed allowable leakage rate 
values specified in the TS and will continue to perform its design 
function following an accident. The risk assessment of the proposed 
changes has concluded that there is an insignificant increase in 
total population dose rate and an insignificant increase in the 
conditional containment failure probability.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes for a one-time extension of the DBLRT and 
Type A test will not affect the control parameters governing unit 
operations or the response of plant equipment to transient and 
accident conditions. The proposed changes do not introduce any new 
equipment or modes of system operation. No installed equipment will 
be operated in a new or different manner. As such, no new failure 
mechanisms are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    CPS is a General Electric BWR/6 plant with a Mark III 
containment system. The Mark III containment design is a single-
barrier pressure containment and a multi-barrier fission containment 
system consisting of the drywell and primary containment. The 
drywell houses the reactor pressure vessel, the reactor coolant 
recirculation loops, and branch connections of the RCS, which have 
isolation valves at the primary containment boundary. The function 
of the drywell is to maintain a pressure boundary that channels 
steam from a LOCA to the suppression pool, where it is condensed. 
The suppression pool is an annular pool of demineralized water 
between the drywell and the outer primary containment boundary. This 
pool covers the horizontal vent openings in the drywell to maintain 
a water seal between the drywell interior and the remainder of the 
containment volume. The primary containment consists of a steel-
lined, reinforced concrete vessel, which surrounds the RCS and 
provides an essentially leak-tight barrier against an uncontrolled 
release of radioactive material to the environment. Additionally, 
the containment structure provides shielding from the fission 
products that may be present in the primary containment atmosphere 
following accident conditions. The primary containment is penetrated 
by access, piping and electrical penetrations.
    The integrity of the drywell is periodically verified by 
performance of the DBLRT. This test ensures that the measured 
drywell bypass leakage is bounded by the safety analysis 
assumptions. The drywell integrity is further verified by a number 
of additional tests, including drywell airlock door seal leakage 
tests, overall drywell airlock leakage tests and periodic visual 
inspections of exposed accessible interior and exterior drywell 
surfaces. Additional confidence that significant degradation in the 
drywell leaktightness has not developed is provided by the periodic 
qualitative assessment of drywell performance.
    The integrity of the primary containment penetrations and 
isolation valves is verified

[[Page 4307]]

through Type B and Type C local leak rate tests (LLRTs) and the 
overall leak-tight integrity of the primary containment is verified 
by a Type A integrated leak rate test (ILRT) as required by 10 CFR 
50, Appendix J. These tests are performed to verify the essentially 
leak-tight characteristics of the primary containment at the design 
basis accident pressure. The proposed changes for a one-time 
extension of the drywell bypass leakage and Type A tests do not 
affect the method for drywell or containment testing or the test 
acceptance criteria.
    AmerGen has conducted a risk assessment to determine the impact 
of a change to the CPS Type A ILRT and DBLRT schedule from the 
originally licensed baseline frequency of three tests in 10 years to 
one test in 15 years plus 15 months (i.e., approximately 16.25 
years) for the risk measures of Large Early Release Frequency (i.e., 
LERF), Population Dose, and Conditional Containment Failure 
Probability (i.e., CCFP). This assessment indicated that the 
proposed CPS interval extension has a small change in risk to the 
public and is an acceptable plant change from a risk perspective.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, AmerGen concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92, ``Issuance of amendment,'' 
paragraph (c), and, accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: November 2, 2006.
    Description of amendment request: The proposed amendment would 
modify requirements for inoperable snubbers consistent with the 
Technical Specification Task Force 372, Revision 4.
    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:

    Criterion 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 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 
[limiting condition for operation] 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.
    Criterion 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.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] 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 [Regulatory Guide] 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 has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599
    NRC Branch Chief: Richard Laufer.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: November 7, 2006.
    Description of amendment request: The proposed change revises 
Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1 to 
increase the allowable as-found main steam safety valve (MSSV) lift 
setpoint tolerance from  1 percent to  3 
percent. In addition, the proposed change revises SR 3.1.7.10 to 
increase the enrichment of sodium pentaborate used in the Standby 
Liquid Control (SLC) system from >= 30.0 atom percent boron-10 to >= 
45.0 atom percent boron-10.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change increases the allowable as-found MSSV lift 
setpoint tolerance, determined by test after the valves have been 
removed from service, from  1% to  3%. The 
proposed change does not alter the TS requirements for the number of 
MSSVs required to be operable, the nominal lift setpoints, the 
allowable as-left lift setpoint tolerance, the MSSV testing 
frequency, or the manner in which the valves are operated.
    Consistent with current TS requirements, the proposed change 
continues to require that the MSSVs be adjusted to within  1% of their nominal lift setpoints following testing. Since 
the proposed change does not alter the manner in which the valves 
are operated, there is no significant impact on reactor operation.
    The proposed change does not involve a physical change to the 
valves, nor does it change the safety function of the valves. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating

[[Page 4308]]

conditions and no changes to existing structures, systems, or 
components, with the exception of the SLC system enrichment change. 
The proposed change to increase the enrichment of sodium pentaborate 
used in the SLC system will ensure that the requirements of 10 CFR 
50.62, ``Requirements for reduction of risk from anticipated 
transients without scram (ATWS) events for light-water-cooled 
nuclear power plants,'' continue to be met. The SLC system is not an 
initiator to an accident; rather, the SLC system is used to mitigate 
an ATWS event. Therefore, these changes will not increase the 
probability of an accident previously evaluated.
    Generic considerations related to the change in setpoint 
tolerance were addressed in NEDC-3175310, ``BWROG In-Service 
Pressure Relief Technical Specification Revision Licensing Topical 
Report,'' and were reviewed and approved by the NRC in a safety 
evaluation dated March 8, 1993. General Electric Company (GE) 
completed plant-specific analyses to assess the impact of the 
setpoint tolerance increase on Dresden Nuclear Power Station Units 2 
and 3 and QCNPS [Quad Cities Nuclear Power Station] Units 1 and 2. 
The impact of the MSSV setpoint tolerance increase, as addressed in 
this analysis, included vessel overpressure, Updated Final Safety 
Analysis Report (UFSAR) Chapter 15 events, ATWS, Loss of Coolant 
Accident (LOCA), containment response and loads, high pressure 
systems performance, Appendix R fire protection, vessel thermal 
cycle, operating mode and equipment out of service review, and 
extended power uprate evaluation review. The proposed change to 3% 
setpoint tolerance is supported by Westinghouse SVEA-96 Optimal fuel 
analysis of events that credit the MSSVs.
    The plant specific evaluations, required by the NRC's safety 
evaluation and performed to support this proposed change, show that 
there is no change to the design core thermal limits and adequate 
margin to the reactor vessel pressure limits using a 3% 
lift setpoint tolerance. These analyses also show that operation of 
Emergency Core Cooling Systems is not affected, and the containment 
response following a LOCA is acceptable. The plant systems 
associated with these proposed changes are capable of meeting 
applicable design basis requirements and retain the capability to 
mitigate the consequences of accidents described in the UFSAR. The 
accident analyses that credit the initiation of SLC as a dose 
mitigation feature are not impacted by the proposed change because 
the chemical properties of the SLC boron solution are not affected. 
Therefore, these changes do not involve an increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change increases the allowable as-found lift 
setpoint tolerance for the QCNPS MSSVs, and increases the required 
enrichment of sodium pentaborate used in the SLC system. The 
proposed change to increase the enrichment of sodium pentaborate 
used in the SLC system will ensure that the requirements of 10 CFR 
50.62 continue to be met.
    The proposed change to increase the MSSV tolerance was developed 
in accordance with the provisions contained in the NRC safety 
evaluation for NEDC-31753P. MSSVs installed in the plant following 
testing or refurbishment will continue to meet the current tolerance 
acceptance criteria of  1% of the nominal setpoint. The 
proposed change does not affect the manner in which the overpressure 
protection system is operated; therefore, there are no new failure 
mechanisms for the overpressure protection system. The proposed 
change to allow an increase in the MSSV setpoint tolerance does not 
alter the nominal MSSV lift setpoints or the number of MSSVs 
currently required to be operable by QCNPS TS. The proposed change 
does not involve physical changes to the valves, nor does it change 
the safety function of the valves. There is no alteration to the 
parameters within which the plant is normally operated. As a result, 
no new failure modes are being introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not modify the safety limits or setpoints 
at which protective actions are initiated, and does not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety.
    Establishment of the  3% MSSV setpoint tolerance 
limit does not adversely impact the operation of any safety-related 
component or equipment. Evaluations performed in accordance with the 
NRC safety evaluation for NEDC-31753P have concluded that all design 
limits will continue to be met.
    The proposed change to increase the enrichment of sodium 
pentaborate used in the SLC system will ensure that the requirements 
of 10 CFR 50.62 continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based upon the above, EGC [Exelon Generation Company] concludes 
that the proposed amendment presents no significant hazards 
consideration under the standards set forth in 10 CFR 50.92 (c), 
and, accordingly, a finding of no significant hazards consideration 
is justified.

    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: December 20, 2006.
    Description of amendment request: The proposed amendment would 
remove annotations referencing Technical Data Book (TDB)-VIII, 
``Equipment Operability Guidance,'' and annotations referencing 
Technical Specification Interpretations (TSIs) from the NRC Authority 
File. These documents are used by Omaha Public Power District (OPPD) 
personnel for additional guidance in applying certain Limiting 
Conditions for Operation requirements to specific equipment and/or 
situations. OPPD has annotated references to these documents in the 
Technical Specification (TS) copies used at Fort Calhoun Station (FCS); 
however, the annotations are ``pointers'' to additional guidance and 
are not officially a part of the FCS TS. The proposed amendment also 
corrects an administrative discrepancy in TS 2.10.4(1)(c).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The correction of administrative discrepancies in the Fort 
Calhoun Station (FCS) Technical Specifications (TS) is not an 
initiator of any previously evaluated accident. The proposed changes 
will not prevent safety systems from performing their accident 
mitigation function as assumed in the safety analysis.
    Therefore, this change does not involve an increase in the 
probability or consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes only affect the Technical Specifications 
and do not involve a physical change to the plant. Modifications

[[Page 4309]]

will not be made to existing components nor will any new or 
different types of equipment be installed. This change will not 
alter assumptions made in safety analysis and licensing bases.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The correction of administrative discrepancies in the Technical 
Specifications has no impact on any safety analysis assumptions and 
thus this TS change does not involve a reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: December 20, 2006.
    Description of amendment request: The proposed amendment would 
delete the Technical Specification (TS) requirements related to the 
hydrogen purge system in TS 2.6(3) and TS Table 3-5, Item 17. The 
proposed TS changes support implementation of the revisions to 10 CFR 
50.44, ``Standards for Combustible Gas Control System in Light-Water-
Cooled Power Reactors,'' that became effective on September 16, 2003. 
The changes are consistent with Revision 1 of NRC-approved Industry/
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.''
    The NRC staff issued a notice of opportunity to comment in the 
Federal Register dated August 2, 2002 (67 FR 50374), on possible 
amendments for the elimination of requirements for hydrogen 
recombiners, and hydrogen and oxygen monitors from the TSs, 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 September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the NSHC in its application dated 
December 20, 2006.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen [and oxygen] monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG 1.97 Category 1, is intended for key variables that 
most directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen [and oxygen] monitors no 
longer meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44 the Commission found that Category 
3, as defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents. [Also, as part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.]
    The regulatory requirements for the hydrogen [and oxygen] 
monitors can be relaxed without degrading the plant emergency 
response. The emergency response, in this sense, refers to the 
methodologies used in ascertaining the condition of the reactor 
core, mitigating the consequences of an accident, assessing and 
projecting offsite releases of radioactivity, and establishing 
protective action recommendations to be communicated to offsite 
authorities. Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2] and removal of 
the hydrogen [and oxygen] monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs, the 
emergency plan (EP), the emergency operating procedures (EOP), and 
site survey monitoring that support modification of emergency plan 
protective action recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen [and oxygen] monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen [and oxygen] monitor equipment was intended to mitigate 
a design-basis hydrogen release. The hydrogen recombiner and 
hydrogen [and oxygen] monitor equipment are not considered accident 
precursors, nor does their existence or elimination have any adverse 
impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. [The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be

[[Page 4310]]

adequately met without reliance on safety-related oxygen monitors.] 
Removal of hydrogen [and oxygen] monitoring from TS will not result 
in a significant reduction in their functionality, reliability, and 
availability.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action, see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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: December 27, 2004, as 
supplemented by letters dated October 27, 2005, March 10, and October 
6, 2006.
    Brief description of amendments: The amendments revised the San 
Onofre Nuclear Generating Station, Units 2 and 3, accident source term 
used in the design-basis radiological consequence analyses. The 
amendments were in accordance with the requirements of 10 CFR 50.67, 
which addresses the use of an alternative source term (AST) at 
operating reactors, and relevant guidance of Regulatory Guide (RG) 
1.183. The amendments represent full-scope implementation of the AST 
described in RG 1.183.
    Date of issuance: December 29, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: Unit 2--210; Unit 3--202.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5248). The supplemental letters dated October 27, 2005, March 10, and 
October 6, 2006, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 10, 2006, as supplemented 
by submittal dated May 16, 2006.
    Brief description of amendments: The amendments conform the 
Facility Operating Licenses NPF-10 and NPF-15 for the San Onofre 
Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3) to reflect 
their transfer from the City of Anaheim (Anaheim) to Southern 
California Edison (SCE). The license transfers, which were approved by 
the Order dated September 27, 2006, permitted the transfer of the 3.16-
percent undivided ownership interest in the facilities held by Anaheim 
to SCE, excluding Anaheim's interest in its spent fuel and in the SONGS 
2 and 3 independent spent fuel storage installation. SCE retains 
exclusive responsibility and control over the operation of SONGS 2 and 
3.
    Date of issuance: December 29, 2006.
    Effective date: At the time the transfer is completed.
    Amendment Nos.: Unit 2--209; Unit 3--201.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: June 8, 2006 (71 FR 
33321) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 27, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 6, 2006 (TS-443), as 
supplemented by letter dated October 2, 2006.
    Brief description of amendment: Activation of thermal-hydraulic 
stability monitoring instrumentation. The Oscillation Power Range 
Monitor System is designed to provide the licensee's solution regarding 
reactor stability.
    Date of issuance: December 29, 2006.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 266.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the TSs.
    Date of initial notice in Federal Register: April 5, 2006 (71 FR 
23962). The October 2, 2006, supplement, contained clarifying 
information and

[[Page 4311]]

did not change the NRC staff's initial proposed finding of no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 29, 2006.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, 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 System's (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.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, 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 identify the specific contentions which the petitioner/

[[Page 4312]]

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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ 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 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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    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 or 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).

STP Nuclear Operating Company, Docket No. 50-498, South Texas Project, 
Unit 1, Matagorda County, Texas

    Date of amendment request: December 20, 2006, as supplemented by 
letter dated December 28, 2006.
    Description of amendment request: The amendment, for a one-time 
change, revised Technical Specification (TS) 3.3.2 for the loss of 
power (LOP) instrumentation (Functional Unit 8, ``loss of power'') in 
TS Table 3.3-3, ``Engineered Safety Features Actuation System 
Instrumentation.'' A note is added to TS Table 3.3-3, Action 20, which 
is the TS-required action for inoperable LOP instrumentation, to allow 
a one-time provision for corrective maintenance on an inoperable Unit 1 
LOP instrumentation channel when the number of operable channels are 
more than one less than the total number of channels. This provision 
for corrective maintenance expires 30 days after the amendment is 
approved.
    Date of issuance: January 11, 2007.
    Effective date: Effective as of its date of issuance and shall be 
implemented by January 15, 2007.
    Amendment No.: 176.
    Facility Operating License No. NPF-76: The amendment revised the 
Technical Specifications and Facility Operating License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated January 11, 
200