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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 54085-54094 [05-17888] :: Nuclear Regulatory Commission :: Agencies And Commissions :: Regulation Tracker :: Justia
Justia Regulation Tracker Agencies And Commissions Nuclear Regulatory Commission Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 54085-54094 [05-17888]
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 54085-54094 [05-17888]
Download as PDF Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices Week of October 17, 2005—Tentative NATIONAL TRANSPORTATION SAFETY BOARD Meeting Notice 9:30 a.m., Tuesday, September 20, 2005. PLACE: NTSB Board Room, 429 L’Enfant Plaza, S.W., Washington, DC 20594. STATUS: The one item is Open to the Public. MATTER TO BE CONSIDERED: 5299R—Most Wanted Safety Recommendations Program—2005 Update on State Issues. NEWS MEDIA CONTACT: Telephone: (202) 314–6100. Individuals requesting specific accommodations should contact Ms. Carolyn Dargan at (202) 314–6305 by Friday, September 16, 2005. The public may view the meeting via a live or archived webcast by accessing a link under ‘‘News & Events’’ on the NTSB home page at http:// www.ntsb.gov. FOR MORE INFORMATION CONTACT: Vicky D’Onofrio, (202) 314–6410. TIME AND DATE: Dated: September 9, 2005. Vicky D’Onofrio, Federal Register Liaison Officer. [FR Doc. 05–18294 Filed 9–9–05; 3:57 am] BILLING CODE 7533–01–M NUCLEAR REGULATORY COMMISSION Sunshine Act; Meetings Weeks of September 12, 19, 26, October 3, 10, 17, 2005. PLACE: Commissioners’ Conference Room, 11555 Rockville Pike, Rockville, Maryland. STATUS: Public and Closed. MATTERS TO BE CONSIDERED: DATE: Week of September 12, 2005 There are no meetings scheduled for the Week of September 12, 2005. Week of September 19, 2005—Tentative There are no meetings scheduled for the Week of September 19, 2005. Week of September 26, 2005—Tentative There are no meetings scheduled for the Week of September 26, 2005. Week of October 3, 2005—Tentative There are no meetings scheduled for the Week of October 3, 2005. Week of October 10, 2005—Tentative There are no meetings scheduled for the Week of October 10, 2005. VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 Tuesday, October 18, 2005 9:30 a.m.—Briefing on Decommissioning Activities and Status (Public Meeting). This meeting will be webcast live at the Web address http://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. * * * * * The NRC Commission Meeting Schedule can be found on the Internet at: http://www.nrc.gov/what-we-do/ policy-making/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, August Spector, at 301–415–7080, TDD: 301–415–2100, or by e-mail at aks@nrc.gov. Determinations on request 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 to be added to the distribution, please contact the Office of the Secretary, Washington, DC 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: September 8, 2005. R. Michelle Schroll, Office of the Secretary. [FR Doc. 05–18191 Filed 9–9–05; 10:33 am] BILLING CODE 7590–01–M 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 PO 00000 Frm 00081 Fmt 4703 Sfmt 4703 54085 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 August 19, 2005, to August 31, 2005. The last biweekly notice was published on August 30, 2005 (70 FR 51378). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards E:\FR\FM\13SEN1.SGM 13SEN1 54086 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices 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, Rules and Directives 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 VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 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 PO 00000 Frm 00082 Fmt 4703 Sfmt 4703 intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415–1101, verification number is (301) 415–1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415–3725 or by 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). E:\FR\FM\13SEN1.SGM 13SEN1 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397– 4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. Carolina Power & Light Company, Docket Nos. 50–325 and 50–324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendments request: August 11, 2005. Description of amendments request: The proposed change would revise Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.9 with respect to the allowed leakage rate through each Main Steam Isolation Valve (MSIV). Specifically, the limit is revised from an allowable leakage rate of less than or equal to 11.5 standard cubic feet per hour (scfh) through each MSIV to less than or equal to 100 scfh through each main steam line (MSL) with the combined leakage of the four MSLs being less than or equal to 150 scfh. Also, changes to TS 3.3.7.1, ‘‘Control Room Emergency Ventilation (CREV) System Instrumentation,’’ are also included to incorporate new automatic initiation functions for the CREV system to support the MSIV leakage rate change proposal. 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 not involve a significant increase in the probability or consequences of an accident previously evaluated. Response: No. The proposed change revises SR 3.6.1.3.9 with respect to the allowed leakage rate through each MSIV. Specifically, the limit is revised from an allowable leakage rate of less than or equal to 11.5 scfh per MSIV to less than or equal to 100 scfh for any one MSL with the combined leakage of the four MSLs being less than or equal to 150 scfh. Also, to support the MSIV leakage rate change, additional automatic initiation functions for the CREV system will be implemented. The associated changes to TS 3.3.7.1, ‘‘Control VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 Room Emergency Ventilation (CREV) System Instrumentation,’’ are also made. The proposed change to the MSIV leakage limit does not involve physical change to any plant structure, system, or component. As a result, no new failure modes of the MSIVs has been introduced. The CREV system initiation logic is being modified; however, this system performs a mitigating function and has no impact on any initiating event frequency. Therefore, the proposed changes cannot increase in the probability a previously evaluated accident. A plant-specific radiological analysis has been performed to assess the effects of the proposed increase in MSIV leakage acceptance criteria in terms of offsite doses and control room doses. The analysis shows the dose contribution from the proposed increase in leakage acceptance criteria is acceptable compared to dose limits prescribed in 10 CFR 50.67(b)(2)(i) for the exclusion area, 10 CFR 50.67(b)(2)(ii) for the low population zone, and 10 CFR 50.67(b)(2)(iii) for control room personnel. The CREV system initiation logic modification will result in automatic initiation of the CREV system based on signals from the secondary containment isolation logic as an input to each division of the CREV control logic. This change is made to ensure that doses to control room personnel remain within the requirements of 10 CFR 50.67(b)(2)(iii) in the event of a lossof-coolant-accident [LOCA]. 2. Does not create the possibility of a new or different type of accident from any accident previously evaluated. Response: No. The proposed change to the MSIV leakage limit will not adversely impact MSIV functionality and will not create a failure of the MSIVs of a different kind than previously considered. The CREV system initiation logic is being modified to initiate automatically using signals from the secondary containment isolation logic. This provides redundant/diverse protection for control room operators in the event of a LOCA. The required logic modifications will be performed such that faults originating in the CREV logic cannot affect either the secondary containment isolation logic or the functions which initiate secondary containment isolation. 3. Does not involve a significant reduction in the margin of safety. Response: No. The allowable leak rate specified for the MSIVs is used to quantify a maximum amount of leakage assumed to bypass containment. The results of the re-analysis supporting these changes were evaluated against the dose limits contained in 10 CFR 50.67(b)(2)(i) for the exclusion area, 10 CFR 50.67(b)(2)(ii) for the low population zone, and 10 CFR 50.67(b)(2)(iii) for control room personnel. Sufficient margin relative to the regulatory limits is maintained even when conservative assumptions and methods are utilized. The CREV system initiation logic is being modified to initiate automatically using signals from the secondary containment isolation logic. This provides redundant/ diverse protection for control room operators in the event of a LOCA. PO 00000 Frm 00083 Fmt 4703 Sfmt 4703 54087 The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David T. Conley, Associate General Counsel II— Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Section Chief: Michael L. Marshall, Jr. Carolina Power & Light Company, Docket Nos. 50–325 and 50–324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendments request: August 11, 2005. Description of amendments request: The proposed change would revise Technical Specification (TS) 5.5.12, ‘‘Primary Containment Leakage Rate Testing Program,’’ by removing an exception that allows for compensation of flow meter instrument inaccuracies in accordance with ANSI [American National Standards Institute]/ANS [American Nuclear Society]–56.8–1987 rather than meeting the instrument accuracy requirements in ANSI/ANS– 56.8–1994. The exception is no longer necessary due to the availability of test instruments capable of satisfying the instrument accuracy requirements of ANSI/ANS–56.8–1994. 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: This request has been evaluated against the standards in 10 CFR 50.92, and has been determined to not involve a significant hazards consideration. In support of this conclusion, the following analysis is provided: 1. Does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed removal, from Technical Specification 5.5.12, of an exception that allows for compensation of instrumentation inaccuracies in accordance with ANSI/ANS– 56.8–1987, rather than ANSI/ANS–56.8– 1994, does not involve physical changes to any plant structure, system, or component. Furthermore, removal of the exception allowing for the accounting for containment leakage rate test instrumentation accuracy using ANSI/ANS–56.8–1987 has no impact on the initiating frequency for any previously evaluated accident. Therefore, the proposed change cannot increase the probability of a previously evaluated accident. E:\FR\FM\13SEN1.SGM 13SEN1 54088 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices The consequences of a previously evaluated accident are dependent on the initial conditions assumed for the analysis, the behavior of the fuel during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the evaluated event, and the setpoints at which these actions are initiated. Use of leakage rate test instruments that meet the accuracy provisions of ANSI/ ANS–56.8–1994 complies with NEI [Nuclear Energy Institute] 94–01, Revision 0, ‘‘Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J,’’ and Regulatory Guide 1.163, ‘‘Performance-Based Containment Leak—Test Program,’’ September 1995, and ensures that measured containment leakage rates are maintained within specified limits. 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? Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors to that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change regarding containment leakage test instrument accuracy does not involve installation of any new or different equipment. No installed equipment is being operated in a different manner than currently evaluated. No new initiating events or transients will result from the use of more accurate containment leakage test instruments. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does not involve a significant reduction in the margin of safety. The proposed removal, from Technical Specification 5.5.12, of an exception that allows for compensation of instrumentation inaccuracies in accordance with ANSI/ANS– 56.8–1987 rather than ANSI/ANS–56.8–1994 does not alter the assumptions of the accident analyses or the Technical Specification Bases. The margin of safety is established through the design of the plant structures, systems, and components; through the parameters within which the plant is operated; through the establishment of setpoints for actuation of equipment relied upon to respond to an event; and through margins contained within the safety analyses. The use of industry standard ANSI/ANS– 56.8–1994, rather than ANSI/ANS–56.8– 1987, in accounting for the accuracy of containment leakage rate testing instrumentation will not adversely impact the performance of plant structures, systems, components, and setpoints relied upon to respond to mitigate an accident or transient. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David T. Conley, Associate General Counsel II— Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Section Chief: Michael L. Marshall, Jr. PPL Susquehanna, LLC, Docket Nos. 50– 387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania Date of amendment request: June 27, 2005. Description of amendment request: The proposed amendments would change the SSES 1 and 2 technical specifications for reactor protection system and control rod block instrumentation, oscillation power range monitor (OPRM) instrumentation, recirculation loops operating, shutdown margin test—refueling, and the core operating limits report. The proposed changes involve the modification of the existing power range neutron monitor system (PRNM) by installation of the General Electric Nuclear Measurement Analysis and Control PRNM system. The modification of the PRNM system would replace analog technology with a more reliable digital upgrade. 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 probability (frequency of occurrence) of DBAs [design-basis accidents] occurring is not affected by the PRNM system, as the PRNM system does not interact with equipment whose failure could cause an accident. Compliance with the regulatory criteria established for plant equipment will be maintained with the installation of the upgraded PRNM system. Scram setpoints in the PRNM system will be established so that all analytical limits are met. The unavailability of the new system will be equal to or less than the existing system and, as a result, the scram reliability will be equal to or better than the existing system. No new challenges to safety-related equipment will result from the PRNM system modification. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. PO 00000 Frm 00084 Fmt 4703 Sfmt 4703 The proposed change will replace the currently installed and NRC approved OPRM Option III long-term stability solution with an NRC approved Option III long-term stability solution digitally integrated into the PRNM equipment. The PRNM hardware incorporates the OPRM Option III detect and suppress solution reviewed and approved by the NRC in the References 1, 2, 3 and 4 Licensing Topical Reports, the same as the currently installed separate OPRM system. The OPRM meets the GDC [general design criterion] 10, ‘‘Reactor Design,’’ and 12, ‘‘Suppression of Reactor Power Oscillations,’’ requirements by automatically detecting and suppressing design basis thermal-hydraulic oscillations prior to exceeding the fuel MCPR [minimum critical power ratio] Safety Limit. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Based on the above, the operation of the new PRNM system and replacement of the currently installed OPRM Option III stability solution with the Option III OPRM function integrated into the PRNM equipment will not increase 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 components of the PRNM system will be supplied to equivalent or better design and qualification criteria than is currently required for the plant. Equipment that could be affected by PRNM system has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or system interaction mode was identified. Therefore, the upgraded PRNM system will not adversely affect plant equipment. The new PRNM system uses digital equipment that has ‘‘control’’ processing points and software controlled digital processing compared to the existing PRNM system that uses mostly analog and discrete component processing (excluding the existing OPRM). Specific failures of hardware and potential software common cause failures are different from the existing system. The effects of potential software common cause failure are mitigated by specific hardware design and system architecture. Failure(s) on the system has the same overall effect. No new or different kind of accident is introduced. Therefore, the PRNM system will not adversely effect plant equipment. The current OPRM Option III plant design is replaced with an OPRM function digitally integrated into the PRNM. The currently installed Power Range Monitor system is replaced with a PRNM system that performs all of the existing PRNM functions plus OPRM. Failure of neither the APRM [average power range monitor] nor OPRM functions in the replacement system can cause an accident of a kind not previously evaluated in the SAR [safety analysis report]. Based on the above, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated. E:\FR\FM\13SEN1.SGM 13SEN1 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The upgraded PRNM system will not involve a reduction in a margin of safety, as loads on plant equipment will not increase, and reactions to, or results of transients and hypothetical accidents, will not increase from those presently evaluated. No change has been made to the Analytical Limits or Technical Specification Allowable Values. The present system characteristics such as drift, calibration setpoint, and accuracy envelop the new system requirements. The upgraded PRNM system response time and operator information is either maintained or improved over the current Power Range Neutron Monitor system. The upgraded PRNM system has improved channel trip accuracy compared to the current system. The current safety analyses demonstrate that the existing OPRM Option III related Technical Specification requirements are adequate to detect and suppress an instability event. There is no impact on the MCPR Safety Limit identified for an instability event. The replacement OPRM system integrated into the new PRNM equipment implements the same functions per the same requirements as the currently installed system and has equivalent Technical Specification requirements. Therefore, the margin of safety associated with the MCPR Safety Limit is still maintained. Based on the above, the proposed change will not involve a significant reduction in [a] margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101–1179. NRC Section Chief: Richard J. Laufer. Southern California Edison Company, et al., Docket Nos. 50–361 and 50–362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California Date of amendment requests: July 15, 2005. Description of amendment requests: The amendments are for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, operating licenses, but they will involve Unit 1, which is not an operating nuclear plant and is in the process of being decommissioned. The amendments would revise License Condition 2.B.(6) for both SONGS, Units 2 and 3 by (1) deleting the sentence ‘‘Transhipment of Unit 1 fuel between Units 1 and [2 or 3] shall be in VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 accordance with SCE [Southern California Edison] letters to U.S. Nuclear Regulatory Commission dated * * * and in accordance with the Quality Assurance requirements of 10 CFR Part 71’’ and (2) adding the phrase ‘‘and by the decommissioning of San Onofre Nuclear Generating Station Unit 1’’ to the remaining sentence in the license condition. This change would recognize that Unit 1 is now in the stage of decommissioning and that in the future any radioactive waste water produced in the further decommissioning of Unit 1 would be released from the San Onofre site by transferring the waste water from Unit 1 to Units 2 and 3. The processing (if required) and discharging of this waste water would be using the Units 2 and 3 radioactive waste system and ocean outfall discharge line. 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 any accident previously evaluated? Response: No. The proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated because there is no increase in the total San Onofre Nuclear Generating Station (SONGS) Unit 1 radioactive wastewater created by this change. The yard drain sump and all interconnecting piping will be entirely within the SONGS owner-controlled area. The new design will have more above ground piping, which presents an increase in [pipe] break probability. However, the system design complies with guidelines provided in NRC [Nuclear Regulatory Commission] Regulatory Guide 1.26 for nuclear service and with American National Standards Institute (ANSI) B31.1. Failure of the above ground piping is bounded by the Postulated Radioactive Releases Due to Liquid Tank failures, as described in the Updated Final Safety Analysis Report (UFSAR) Safety Analyses. The proposed change will allow wastewater produced and currently being discharged at Unit 1[, using approved programs and procedures as allowed by the SONGS Unit 1 license,] to be discharged through the SONGS [Unit] 2 or 3 ocean outfall using the established systems, programs, and procedures [as allowed by the SONGS, Units 2 and 3 licenses]. [Unit 1 is not operating and is in the process of being decommissioned.] There will be no increase in the total radioactivity or quantity of wastewater released from the site as a result of the change. The existing SONGS[, Units] 2 and 3 radioactive effluent control program PO 00000 Frm 00085 Fmt 4703 Sfmt 4703 54089 as required by Technical Specification 5.5.2.3 will still be met. Therefore, the probability or consequences of any accident previously evaluated [for Units 2 and 3] is not [significantly] increased. (2) Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The transfer of the SONGS Unit 1 sump discharge to the SONGS [Unit] 2 or 3 outfall does not create a new or different kind of accident. Within SONGS[, Unit] 2 and 3, the new piping will be constructed and supported consistent with the mechanical design standards for radioactive service water piping. These standards ensure design adequacy for intended function and service. The pipe routing is away from any plant system credited for either Unit’s safe shutdown, so a pipe rupture cannot impact the safe operation of SONGS[, Units] 2 and 3. The yard areas are already analyzed for postulated radioactive pipe rupture from the SONGS[, Units] 2 and 3 radwaste discharge piping. The addition of the Unit 1 yard sump pipeline that traverses SONGS[, Units] 2 and 3 does not create a new or different kind of accident. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created [for Units 2 and 3]. (3) Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change will allow radioactive or potentially radioactive waste water produced and currently being discharged at Unit 1 using approved programs and procedures as allowed by the SONGS Unit 1 license, to be discharged through the SONGS[, Units] 2 and 3 ocean outfalls using the approved programs and procedures as allowed by the SONGS[, Units] 2 and 3 licenses. A pipe rupture at SONGS[, Units] 2 and 3 will not significantly reduce the margin of safety. Any water from a rupture in this pipe will be collected and diverted to the yard drains, where it will mix with the SONGS [Unit] 2 or 3 outfalls. The discharge of the waste water from Unit 1 through either Unit 2 or 3 outfall will be performed in accordance with existing programs and procedures. In addition, the radiation monitor and its interlocks will be used to control the release from the yard drain sump. The concentration at the outfall will be below the regulatory limits in 10 CFR 20 Appendix B. The requirements of the radioactive effluent control program as required by Technical Specification 5.5.2.3 will continue to be met. Therefore, a significant reduction in a margin of safety is not involved [for Units 2 and 3]. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Attorney for licensee: Douglas K. Porter, Esquire, Southern California E:\FR\FM\13SEN1.SGM 13SEN1 54090 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770 NRC Section Chief: Daniel S. Collins, Acting. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration. For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice. Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50–424 and 50– 425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: August 12, 2005. Brief description of amendment request: The proposed amendments would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Technical Specifications (TS) 5.5.9, ‘‘Steam Generator (SG) Tube Surveillance Program,’’ on a one-time basis, to incorporate changes in the SG inspection scope fro VEGP, Unit 2 during Refueling Outage 11 and the subsequent operating cycle. The proposed changes are applicable to Unit 2 only for inspections during Refueling Outage 11 and for the subsequent operating cycle. The proposed changes modify the inspection requirements for portions of SG tubes within the hot leg tubesheet region of the SGs. The license for VEGP, Unit 1 is affected only due to the fact that Units 1 and 2 use common TSs. Date of publication of individual notice in Federal Register: August 22, 2005 (70 FR 48985). Expiration date of individual notice: September 21, 2005. 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 VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 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. Carolina Power & Light Company, Docket No. 50–261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of application for amendment: February 14, 2005, as supplemented by letter dated July 13, 2005. Brief description of amendment: The amendment revises the surveillance requirements (SR) for the station batteries as specified in SR 3.8.4.5, battery service test, and SR 3.8.4.6, PO 00000 Frm 00086 Fmt 4703 Sfmt 4703 battery performance test in TS 3.8.4, DC Sources—Operating. Date of issuance: August 25, 2005. Effective date: August 25, 2005. Amendment No.: 206. Renewed Facility Operating License No. DPR–23: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: May 24, 2005 (70 FR 29787). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated August 25, 2005. No significant hazards consideration comments received: No. Duke Energy Corporation, Docket Nos. 50–369 and 50–370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of application for amendments: January 19, 2005. Brief description of amendments: The amendments revised the Technical Specifications 5.6.7.b, ‘‘Core Operating Limits Report (COLR),’’ to add the topical report DPC–NE–1005P–A, ‘‘Duke Power Nuclear Design Methodology Using CASMO–4/SIMULATE–3 MOX.’’ This report has been previously approved by the Nuclear Regulatory Commission. Date of issuance: August 23, 2005. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment Nos.: 230 and 212. Renewed Facility Operating License Nos. NPF–9 and NPF–17: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: March 1, 2005 (70 FR 9990). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated August 23, 2005. No significant hazards consideration comments received: No. Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50–458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana Date of amendment request: September 23, 2004, as supplemented by letter dated April 19, 2005. Brief description of amendment: The amendment revises the Technical Specifications to allow revision of reactor operational limits, as specified in the River Bend Station Core Operating Limits Report, to compensate for the inoperability of the End of Cycle Recirculation Pump Trip Instrumentation. Date of issuance: August 25, 2005. Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance. E:\FR\FM\13SEN1.SGM 13SEN1 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices Amendment No.: 146. Facility Operating License No. NPF– 47: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: May 10, 2005 (70 FR 24650). The supplement dated April 19, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated August 25, 2005. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50–293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of application for amendment: December 14, 2004. Brief description of amendment: The amendment changed Technical Specifications (TSs) to reflect surveillance frequency improvements. Specifically, the amendment removed the additional requirement to perform functional testing of the average power range monitor (APRM) and anticipated transient without scram recirculation pump trip alternate rod insertion instrumentation on each startup, when the nominally-required quarterly testing is current. Additionally, performance of the APRM High Flux heat balance calibration was modified to apply only after 12 hours at > 25% power. Additional editorial clarifications related to TS Tables 4.2.A through 4.2.G, Note 2 and associated Table references were also included. Date of issuance: August 29, 2005. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 217. Facility Operating License No. DPR– 35: The amendment revised the TSs. Date of initial notice in Federal Register: March 1, 2005 (70 FR 9991). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated August 29, 2005. No significant hazards consideration comments received: No. VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50–271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Date of application for amendment: October 5, 2004, as supplemented on April 22, 2005. Brief description of amendment: The amendment revised Technical Specification (TS) 6.7.C ‘‘Primary Containment Leak Rate Testing Program,’’ to allow a one-time extension to the 10-year interval for performing the next Type A containment integrated leak rate test (ILRT). Specifically, the change would allow the test to be performed within 15 years from the last ILRT which was performed in April 1995. Date of Issuance: August 31, 2005. Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment No.: 227. Facility Operating License No. DPR– 28: The amendment revised the TSs. Date of initial notice in Federal Register: December 21, 2004 (69 FR 76492). The supplement contained clarifying information only, and did not change the initial no significant hazards consideration determination or expand the scope of the initial Federal Register notice. The Commission’s related evaluation of this amendment is contained in a Safety Evaluation dated August 31, 2005. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50–237 and 50–249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Date of application for amendments: January 21, 2005. Brief description of amendments: The amendments modify the Isolation Condenser System heat removal capability surveillance requirement (SR) by adding a note to the technical specification section SR 3.5.3.4. This note allows a delay of 12 hours after adequate reactor power is achieved to perform the test. Date of issuance: August 25, 2005. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment Nos.: 215,207. Facility Operating License Nos. DPR– 19 and DPR–25: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: May 24, 2005. The Commission’s related evaluation of the amendments is contained in a PO 00000 Frm 00087 Fmt 4703 Sfmt 4703 54091 Safety Evaluation dated August 25, 2005. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50–352 and 50–353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of application for amendments: May 20, 2004, as supplemented by letters dated February 18 and July 13, 2005. Brief description of amendments: The amendments revised the Limerick Generating Station Units 1 and 2 Technical Specifications (TSs) 2.2.1, ‘‘Reactor Protection System Instrumentation Setpoints,’’ TS 3/4.3.1, ‘‘Reactor Protection System Instrumentation,’’ TS 3/4.3.6, ‘‘Control Rod Block Instrumentation,’’ TS 3/4.4.1, ‘‘Recirculation System,’’ and TS 6.9.1, ‘‘Routine Reports,’’ and the associated TS Bases. The amendments support activation of the trip outputs of the oscillation power range monitor portion of the power range neutron monitoring system. Date of issuance: August 26, 2005. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment Nos.: 177 and 139. Facility Operating License Nos. NPF– 39 and NPF–85. The amendments revised the TSs. Date of initial notice in Federal Register: October 26, 2004 (69 FR 62474). The supplements dated February 18 and July 13, 2005, 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 August 26, 2005. No significant hazards consideration comments received: No. South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50–395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina Date of application for amendment: August 24, 2004, as supplemented August 10, 2005. Brief description of amendment: This amendment revises Technical Specifications (TSs) related to the surveillance requirements for the emergency feedwater system. Date of issuance: August 16, 2005. E:\FR\FM\13SEN1.SGM 13SEN1 54092 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment No.: 173. Renewed Facility Operating License No. NPF–12: Amendment revises the TSs. Date of initial notice in Federal Register: October 12, 2004 (69 FR 60685). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2005. 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: June 30, 2004, as supplemented by letters dated December 2, 2004, May 27, 2005, and July 18, 2005. Brief description of amendments: The proposed changes revise Technical Specification 5.5.2.15, ‘‘Containment Leakage Rate Testing Program,’’ to include a one-time extension of the 10year period of the performance-based leakage rate testing program for Type A tests as prescribed by the Nuclear Energy Institute (NEI) 94–01, Revision 0, ‘‘Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J.’’ Date of issuance: August 24, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 198 and 189. Facility Operating License Nos. NPF– 10 and NPF–15: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: August 3, 2004 (69 FR 46589). The supplemental letters dated December 2, 2004, May 27, and July 18, 2005, provided information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated August 24, 2005. No significant hazards consideration comments received: No. VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 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 PO 00000 Frm 00088 Fmt 4703 Sfmt 4703 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 E:\FR\FM\13SEN1.SGM 13SEN1 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices 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/ 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 VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 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 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 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. PO 00000 Frm 00089 Fmt 4703 Sfmt 4703 54093 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). Duke Energy Corporation, Docket Nos. 50–269, 50–270, and 50–287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: August 21, 2005, as supplemented August 22, 2005. Description of amendment request: The amendments revise Technical Specification Limiting Condition for Operation 3.8.1, Condition C.2.1, to permit a one-time extension of 96 hours of the Completion Times for Keowee Hydro Unit 2. Date of issuance: August 23, 2005. Effective date: August 23, 2005. E:\FR\FM\13SEN1.SGM 13SEN1 54094 Federal Register / Vol. 70, No. 176 / Tuesday, September 13, 2005 / Notices Amendment Nos.: 347, 349, and 348. Facility Operating License Nos. DPR– 38, DPR–47, and DPR–55: Amendments revises the technical specifications. 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 August 23, 2005. Dated at Rockville, Maryland, this 1st day of September, 2005. For the Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation. [FR Doc. 05–17888 Filed 9–12–05; 8:45 am] BILLING CODE 7590–01–P OFFICE OF PERSONNEL MANAGEMENT Proposed Collection; Comment Request for Clearance of a Revised Information Collection: Declaration for Federal Employment; Optional Form 306, OMB No. 3206–0182 Office of Personnel Management. ACTION: Notice. AGENCY: SUMMARY: In accordance with the Paperwork Reduction Act of 1995 (Public Law 104–13, May 22, 1995), this notice announces that the Office of Personnel Management (OPM) intends to submit a request to the Office of Management and Budget (OMB) for review of a revised information collection. The Optional Form (OF) 306, Declaration for Federal Employment, is completed by applicants who are under consideration for Federal or Federal contract employment. The OF 306 requests that the applicant provide personal identifying data, including, for example, general background information, information concerning retirement pay received or requested and information on Selective Service registration and military service. The revision is to make needed administrative updates. It is estimated that 474,000 individuals will respond annually. Each form takes approximately 15 minutes to complete. The annual estimated burden is 118,500 hours. Comments are particularly invited on: • Whether this collection of information is necessary for the proper performance of functions of OPM and its Center for Federal Investigative VerDate Aug<18>2005 16:06 Sep 12, 2005 Jkt 205001 Services, which administers background investigations; • Whether our estimate of the public burden of this collection is accurate and based on valid assumptions and methodology; • Ways in which we can minimize the burden of the collection of information on those who are to respond, through use of the appropriate technological collection techniques or other forms of information technology; • Ways in which we can enhance the quality, utility and clarity of the information to be collected. For copies of this proposal, contact Mary Beth Smith-Toomey, OPM Forms Officer, at (202) 606–8358, FAX (202) 418–3251 or mbtoomey@opm.gov. Please include your mailing address with your request. DATES: Comments on this proposal should be received within 60 calendar days from the date of this publication. ADDRESSES: Send or deliver comments to: Kathy Dillaman, Deputy Associate Director, Center for Federal Investigative Services, U.S. Office of Personnel Management, 1900 E Street, NW., Room 5416, Washington, DC 20415. For information regarding administrative coordination contact: Mary-Kay Brewer—Program Analyst, Standards and Evaluation Group, Center for Federal Investigative Services, U.S. Office of Personnel Management, (202) 606–1042. U.S. Office of Personnel Management. Linda M. Springer, Director. [FR Doc. 05–18140 Filed 9–12–05; 8:45 am] Approximately 2,893 SF 3102 forms are completed annually. Each form takes approximately 15 minutes to complete. The annual estimated burden is 723 hours. For copies of this proposal, contact Mary Beth Smith-Toomey on (202) 606– 8358, FAX (202) 418–3251 or via e-mail to mbtoomey@opm.gov. Please include a mailing address with your request. DATES: Comments on this proposal should be received within 30 calendar days from the date of this publication. ADDRESSES: Send or deliver comments to— Pamela S. Israel, Chief, Operations Support Group, Retirement Services Programs, U.S. Office of Personnel Management, 1900 E Street, NW., Room 3349, Washington, DC 20415; and Brenda Aguilar, OPM Desk Officer, Office of Information and Regulatory Affairs, Office of Management and Budget, New Executive Office Building, NW., Room 10235, Washington, DC 20503. For Information Regarding Administrative Coordination—Contact: Cyrus S. Benson, Team Leader, Publications Team, RIS Support Services/Support Group, (202) 606– 0623. U.S. Office of Personnel Management. Linda M. Springer, Director. [FR Doc. 05–18141 Filed 9–12–05; 8:45 am] BILLING CODE 6325–38–P BILLING CODE 6325–38–P OFFICE OF PERSONNEL MANAGEMENT OFFICE OF PERSONNEL MANAGEMENT Submission for OMB Review; Comment Request for Reclearance of an Information Collection: RI 25–49 Sumission for OMB Review; Comment Request for Reclearance of a Revised Information Collection: SF 3102 Office of Personnel Management. ACTION: Notice. AGENCY: SUMMARY: In accordance with the Paperwork Reduction Act of 1995 (Pub. L. 104–13, May 22, 1995), this notice announces that the Office of Personnel Management (OPM) has submitted to the Office of Management and Budget a request for reclearance of a revised information collection. SF 3102, Designation of Beneficiary (FERS), is used by an employee or an annuitant covered under the Federal Employees Retirement System to designate a beneficiary to receive any lump sum due in the event of his/her death. PO 00000 Frm 00090 Fmt 4703 Sfmt 4703 Office of Personnel Management. ACTION: Notice. AGENCY: SUMMARY: In accordance with the Paperwork Reduction Act of 1995 (Pub. L. 104–13, May 22, 1995), this notice announces that the Office of Personnel Management (OPM) has submitted to the Office of Management and Budget (OMB) a request for reclearance of an information collection. RI 25–49, Verification of Full-Time School Attendance, is used to verify that adult student annuitants are entitled to payments. OPM must confirm that a full-time enrollment has been maintained. Approximately 10,000 RI 25–49 forms are completed annually. The form takes approximately 60 minutes to complete. E:\FR\FM\13SEN1.SGM 13SEN1
[Pages 54085-54094]
[FR Doc No: 05-17888]
proposed to be issued from August 19, 2005, to August 31, 2005. The
last biweekly notice was published on August 30, 2005 (70 FR 51378).
involves no significant hazards
[[Page 54086]]
consideration. In addition, the Commission may issue the amendment
prior to the expiration of the 30-day comment period should
[[Page 54087]]
Date of amendments request: August 11, 2005.
Description of amendments request: The proposed change would revise
Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.9
with respect to the allowed leakage rate through each Main Steam
Isolation Valve (MSIV). Specifically, the limit is revised from an
allowable leakage rate of less than or equal to 11.5 standard cubic
feet per hour (scfh) through each MSIV to less than or equal to 100
scfh through each main steam line (MSL) with the combined leakage of
the four MSLs being less than or equal to 150 scfh. Also, changes to TS
3.3.7.1, ``Control Room Emergency Ventilation (CREV) System
Instrumentation,'' are also included to incorporate new automatic
initiation functions for the CREV system to support the MSIV leakage
rate change proposal.
The proposed change revises SR 3.6.1.3.9 with respect to the
allowed leakage rate through each MSIV. Specifically, the limit is
revised from an allowable leakage rate of less than or equal to 11.5
scfh per MSIV to less than or equal to 100 scfh for any one MSL with
the combined leakage of the four MSLs being less than or equal to
150 scfh. Also, to support the MSIV leakage rate change, additional
automatic initiation functions for the CREV system will be
implemented. The associated changes to TS 3.3.7.1, ``Control Room
Emergency Ventilation (CREV) System Instrumentation,'' are also
The proposed change to the MSIV leakage limit does not involve
physical change to any plant structure, system, or component. As a
result, no new failure modes of the MSIVs has been introduced. The
CREV system initiation logic is being modified; however, this system
performs a mitigating function and has no impact on any initiating
event frequency. Therefore, the proposed changes cannot increase in
the probability a previously evaluated accident.
A plant-specific radiological analysis has been performed to
assess the effects of the proposed increase in MSIV leakage
acceptance criteria in terms of offsite doses and control room
doses. The analysis shows the dose contribution from the proposed
increase in leakage acceptance criteria is acceptable compared to
dose limits prescribed in 10 CFR 50.67(b)(2)(i) for the exclusion
50.67(b)(2)(iii) for control room personnel. The CREV system
initiation logic modification will result in automatic initiation of
the CREV system based on signals from the secondary containment
isolation logic as an input to each division of the CREV control
logic. This change is made to ensure that doses to control room
personnel remain within the requirements of 10 CFR 50.67(b)(2)(iii)
in the event of a loss-of-coolant-accident [LOCA].
2. Does not create the possibility of a new or different type of
The proposed change to the MSIV leakage limit will not adversely
impact MSIV functionality and will not create a failure of the MSIVs
of a different kind than previously considered. The CREV system
initiation logic is being modified to initiate automatically using
signals from the secondary containment isolation logic. This
provides redundant/diverse protection for control room operators in
the event of a LOCA. The required logic modifications will be
performed such that faults originating in the CREV logic cannot
affect either the secondary containment isolation logic or the
functions which initiate secondary containment isolation.
3. Does not involve a significant reduction in the margin of
The allowable leak rate specified for the MSIVs is used to
quantify a maximum amount of leakage assumed to bypass containment.
The results of the re-analysis supporting these changes were
evaluated against the dose limits contained in 10 CFR 50.67(b)(2)(i)
for the exclusion area, 10 CFR 50.67(b)(2)(ii) for the low
population zone, and 10 CFR 50.67(b)(2)(iii) for control room
personnel. Sufficient margin relative to the regulatory limits is
maintained even when conservative assumptions and methods are
utilized. The CREV system initiation logic is being modified to
initiate automatically using signals from the secondary containment
isolation logic. This provides redundant/diverse protection for
control room operators in the event of a LOCA.
Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate
Testing Program,'' by removing an exception that allows for
compensation of flow meter instrument inaccuracies in accordance with
ANSI [American National Standards Institute]/ANS [American Nuclear
Society]-56.8-1987 rather than meeting the instrument accuracy
requirements in ANSI/ANS-56.8-1994. The exception is no longer
necessary due to the availability of test instruments capable of
satisfying the instrument accuracy requirements of ANSI/ANS-56.8-1994.
This request has been evaluated against the standards in 10 CFR
50.92, and has been determined to not involve a significant hazards
consideration. In support of this conclusion, the following analysis is
The proposed removal, from Technical Specification 5.5.12, of an
exception that allows for compensation of instrumentation
inaccuracies in accordance with ANSI/ANS-56.8-1987, rather than
ANSI/ANS-56.8-1994, does not involve physical changes to any plant
structure, system, or component. Furthermore, removal of the
exception allowing for the accounting for containment leakage rate
test instrumentation accuracy using ANSI/ANS-56.8-1987 has no impact
on the initiating frequency for any previously evaluated accident.
Therefore, the proposed change cannot increase the probability of a
[[Page 54088]]
The consequences of a previously evaluated accident are
dependent on the initial conditions assumed for the analysis, the
behavior of the fuel during the analyzed accident, the availability
and successful functioning of the equipment assumed to operate in
response to the evaluated event, and the setpoints at which these
actions are initiated. Use of leakage rate test instruments that
meet the accuracy provisions of ANSI/ANS-56.8-1994 complies with NEI
[Nuclear Energy Institute] 94-01, Revision 0, ``Industry Guideline
for Implementing Performance-Based Option of 10 CFR 50 Appendix J,''
and Regulatory Guide 1.163, ``Performance-Based Containment Leak--
Test Program,'' September 1995, and ensures that measured
containment leakage rates are maintained within specified limits.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors to
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. The proposed change regarding
containment leakage test instrument accuracy does not involve
installation of any new or different equipment. No installed
equipment is being operated in a different manner than currently
evaluated. No new initiating events or transients will result from
the use of more accurate containment leakage test instruments.
inaccuracies in accordance with ANSI/ANS-56.8-1987 rather than ANSI/
ANS-56.8-1994 does not alter the assumptions of the accident
analyses or the Technical Specification Bases. The margin of safety
is established through the design of the plant structures, systems,
and components; through the parameters within which the plant is
operated; through the establishment of setpoints for actuation of
equipment relied upon to respond to an event; and through margins
contained within the safety analyses. The use of industry standard
ANSI/ANS-56.8-1994, rather than ANSI/ANS-56.8-1987, in accounting
for the accuracy of containment leakage rate testing instrumentation
will not adversely impact the performance of plant structures,
systems, components, and setpoints relied upon to respond to
mitigate an accident or transient. Therefore, the proposed change
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Date of amendment request: June 27, 2005.
change the SSES 1 and 2 technical specifications for reactor protection
system and control rod block instrumentation, oscillation power range
monitor (OPRM) instrumentation, recirculation loops operating, shutdown
margin test--refueling, and the core operating limits report. The
proposed changes involve the modification of the existing power range
neutron monitor system (PRNM) by installation of the General Electric
Nuclear Measurement Analysis and Control PRNM system. The modification
of the PRNM system would replace analog technology with a more reliable
The probability (frequency of occurrence) of DBAs [design-basis
accidents] occurring is not affected by the PRNM system, as the PRNM
system does not interact with equipment whose failure could cause an
accident. Compliance with the regulatory criteria established for
plant equipment will be maintained with the installation of the
upgraded PRNM system. Scram setpoints in the PRNM system will be
established so that all analytical limits are met.
The unavailability of the new system will be equal to or less
than the existing system and, as a result, the scram reliability
will be equal to or better than the existing system. No new
challenges to safety-related equipment will result from the PRNM
system modification. Therefore, the proposed change does not involve
a significant increase in the probability of an accident previously
The proposed change will replace the currently installed and NRC
approved OPRM Option III long-term stability solution with an NRC
approved Option III long-term stability solution digitally
integrated into the PRNM equipment. The PRNM hardware incorporates
the OPRM Option III detect and suppress solution reviewed and
approved by the NRC in the References 1, 2, 3 and 4 Licensing
Topical Reports, the same as the currently installed separate OPRM
system. The OPRM meets the GDC [general design criterion] 10,
``Reactor Design,'' and 12, ``Suppression of Reactor Power
Oscillations,'' requirements by automatically detecting and
suppressing design basis thermal-hydraulic oscillations prior to
exceeding the fuel MCPR [minimum critical power ratio] Safety Limit.
Based on the above, the operation of the new PRNM system and
replacement of the currently installed OPRM Option III stability
solution with the Option III OPRM function integrated into the PRNM
equipment will not increase the probability or consequences of an
The components of the PRNM system will be supplied to equivalent
or better design and qualification criteria than is currently
required for the plant.
Equipment that could be affected by PRNM system has been
evaluated. No new operating mode, safety-related equipment lineup,
accident scenario, or system interaction mode was identified.
Therefore, the upgraded PRNM system will not adversely affect plant
The new PRNM system uses digital equipment that has ``control''
processing points and software controlled digital processing
compared to the existing PRNM system that uses mostly analog and
discrete component processing (excluding the existing OPRM).
Specific failures of hardware and potential software common cause
failures are different from the existing system. The effects of
potential software common cause failure are mitigated by specific
hardware design and system architecture. Failure(s) on the system
has the same overall effect. No new or different kind of accident is
Therefore, the PRNM system will not adversely effect plant
The current OPRM Option III plant design is replaced with an
OPRM function digitally integrated into the PRNM. The currently
installed Power Range Monitor system is replaced with a PRNM system
that performs all of the existing PRNM functions plus OPRM. Failure
of neither the APRM [average power range monitor] nor OPRM functions
in the replacement system can cause an accident of a kind not
previously evaluated in the SAR [safety analysis report].
Based on the above, the proposed change will not create the
[[Page 54089]]
The upgraded PRNM system will not involve a reduction in a
margin of safety, as loads on plant equipment will not increase, and
reactions to, or results of transients and hypothetical accidents,
will not increase from those presently evaluated.
No change has been made to the Analytical Limits or Technical
Specification Allowable Values. The present system characteristics
such as drift, calibration setpoint, and accuracy envelop the new
The upgraded PRNM system response time and operator information
is either maintained or improved over the current Power Range
Neutron Monitor system. The upgraded PRNM system has improved
channel trip accuracy compared to the current system.
The current safety analyses demonstrate that the existing OPRM
Option III related Technical Specification requirements are adequate
to detect and suppress an instability event. There is no impact on
the MCPR Safety Limit identified for an instability event. The
replacement OPRM system integrated into the new PRNM equipment
implements the same functions per the same requirements as the
currently installed system and has equivalent Technical
Specification requirements. Therefore, the margin of safety
associated with the MCPR Safety Limit is still maintained.
Based on the above, the proposed change will not involve a
significant reduction in [a] margin of safety.
Date of amendment requests: July 15, 2005.
Description of amendment requests: The amendments are for the San
Onofre Nuclear Generating Station (SONGS), Units 2 and 3, operating
licenses, but they will involve Unit 1, which is not an operating
nuclear plant and is in the process of being decommissioned. The
amendments would revise License Condition 2.B.(6) for both SONGS, Units
2 and 3 by (1) deleting the sentence ``Transhipment of Unit 1 fuel
between Units 1 and [2 or 3] shall be in accordance with SCE [Southern
California Edison] letters to U.S. Nuclear Regulatory Commission dated
* * * and in accordance with the Quality Assurance requirements of 10
CFR Part 71'' and (2) adding the phrase ``and by the decommissioning of
San Onofre Nuclear Generating Station Unit 1'' to the remaining
sentence in the license condition. This change would recognize that
Unit 1 is now in the stage of decommissioning and that in the future
any radioactive waste water produced in the further decommissioning of
Unit 1 would be released from the San Onofre site by transferring the
waste water from Unit 1 to Units 2 and 3. The processing (if required)
and discharging of this waste water would be using the Units 2 and 3
radioactive waste system and ocean outfall discharge line.
Response: No. The proposed change does not involve a significant
previously evaluated because there is no increase in the total San
Onofre Nuclear Generating Station (SONGS) Unit 1 radioactive
wastewater created by this change.
The yard drain sump and all interconnecting piping will be
entirely within the SONGS owner-controlled area. The new design will
have more above ground piping, which presents an increase in [pipe]
break probability. However, the system design complies with
guidelines provided in NRC [Nuclear Regulatory Commission]
Regulatory Guide 1.26 for nuclear service and with American National
Standards Institute (ANSI) B31.1. Failure of the above ground piping
is bounded by the Postulated Radioactive Releases Due to Liquid Tank
failures, as described in the Updated Final Safety Analysis Report
(UFSAR) Safety Analyses.
The proposed change will allow wastewater produced and currently
being discharged at Unit 1[, using approved programs and procedures
as allowed by the SONGS Unit 1 license,] to be discharged through
the SONGS [Unit] 2 or 3 ocean outfall using the established systems,
programs, and procedures [as allowed by the SONGS, Units 2 and 3
licenses]. [Unit 1 is not operating and is in the process of being
decommissioned.] There will be no increase in the total
radioactivity or quantity of wastewater released from the site as a
result of the change. The existing SONGS[, Units] 2 and 3
radioactive effluent control program as required by Technical
Specification 5.5.2.3 will still be met.
previously evaluated [for Units 2 and 3] is not [significantly]
(2) Does the proposed change create the possibility of a new or
Response: No. The transfer of the SONGS Unit 1 sump discharge to
the SONGS [Unit] 2 or 3 outfall does not create a new or different
kind of accident. Within SONGS[, Unit] 2 and 3, the new piping will
be constructed and supported consistent with the mechanical design
standards for radioactive service water piping. These standards
ensure design adequacy for intended function and service. The pipe
routing is away from any plant system credited for either Unit's
safe shutdown, so a pipe rupture cannot impact the safe operation of
SONGS[, Units] 2 and 3. The yard areas are already analyzed for
postulated radioactive pipe rupture from the SONGS[, Units] 2 and 3
radwaste discharge piping. The addition of the Unit 1 yard sump
pipeline that traverses SONGS[, Units] 2 and 3 does not create a new
or different kind of accident.
accident from any previously evaluated is not created [for Units 2
and 3].
(3) Does the proposed change involve a significant reduction in
Response: No. The proposed change will allow radioactive or
potentially radioactive waste water produced and currently being
discharged at Unit 1 using approved programs and procedures as
allowed by the SONGS Unit 1 license, to be discharged through the
SONGS[, Units] 2 and 3 ocean outfalls using the approved programs
and procedures as allowed by the SONGS[, Units] 2 and 3 licenses. A
pipe rupture at SONGS[, Units] 2 and 3 will not significantly reduce
the margin of safety. Any water from a rupture in this pipe will be
collected and diverted to the yard drains, where it will mix with
the SONGS [Unit] 2 or 3 outfalls.
The discharge of the waste water from Unit 1 through either Unit
2 or 3 outfall will be performed in accordance with existing
programs and procedures. In addition, the radiation monitor and its
interlocks will be used to control the release from the yard drain
sump. The concentration at the outfall will be below the regulatory
limits in 10 CFR 20 Appendix B. The requirements of the radioactive
effluent control program as required by Technical Specification
5.5.2.3 will continue to be met.
Therefore, a significant reduction in a margin of safety is not
involved [for Units 2 and 3].
[[Page 54090]]
Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
Date of amendment request: August 12, 2005.
Brief description of amendment request: The proposed amendments
would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and
2, Technical Specifications (TS) 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program,'' on a one-time basis, to incorporate changes in
the SG inspection scope fro VEGP, Unit 2 during Refueling Outage 11 and
the subsequent operating cycle. The proposed changes are applicable to
Unit 2 only for inspections during Refueling Outage 11 and for the
subsequent operating cycle. The proposed changes modify the inspection
requirements for portions of SG tubes within the hot leg tubesheet
region of the SGs. The license for VEGP, Unit 1 is affected only due to
the fact that Units 1 and 2 use common TSs.
August 22, 2005 (70 FR 48985).
Expiration date of individual notice: September 21, 2005.
Date of application for amendment: February 14, 2005, as
supplemented by letter dated July 13, 2005.
Brief description of amendment: The amendment revises the
surveillance requirements (SR) for the station batteries as specified
in SR 3.8.4.5, battery service test, and SR 3.8.4.6, battery
performance test in TS 3.8.4, DC Sources--Operating.
Date of issuance: August 25, 2005.
Renewed Facility Operating License No. DPR-23: Amendment revises
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29787). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 25, 2005.
Date of application for amendments: January 19, 2005.
Technical Specifications 5.6.7.b, ``Core Operating Limits Report
(COLR),'' to add the topical report DPC-NE-1005P-A, ``Duke Power
Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX.'' This report
has been previously approved by the Nuclear Regulatory Commission.
Date of issuance: August 23, 2005.
Amendment Nos.: 230 and 212.
in a Safety Evaluation dated August 23, 2005.
Date of amendment request: September 23, 2004, as supplemented by
letter dated April 19, 2005.
Specifications to allow revision of reactor operational limits, as
specified in the River Bend Station Core Operating Limits Report, to
compensate for the inoperability of the End of Cycle Recirculation Pump
Trip Instrumentation.
[[Page 54091]]
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24650). The supplement dated April 19, 2005, provided additional
in a Safety Evaluation dated August 25, 2005.
Date of application for amendment: December 14, 2004.
Brief description of amendment: The amendment changed Technical
Specifications (TSs) to reflect surveillance frequency improvements.
Specifically, the amendment removed the additional requirement to
perform functional testing of the average power range monitor (APRM)
and anticipated transient without scram recirculation pump trip
alternate rod insertion instrumentation on each startup, when the
nominally-required quarterly testing is current. Additionally,
performance of the APRM High Flux heat balance calibration was modified
to apply only after 12 hours at > 25% power. Additional editorial
clarifications related to TS Tables 4.2.A through 4.2.G, Note 2 and
associated Table references were also included.
Date of issuance: August 29, 2005.
9991).
in a Safety Evaluation dated August 29, 2005.
Date of application for amendment: October 5, 2004, as supplemented
on April 22, 2005.
Specification (TS) 6.7.C ``Primary Containment Leak Rate Testing
Program,'' to allow a one-time extension to the 10-year interval for
performing the next Type A containment integrated leak rate test
(ILRT). Specifically, the change would allow the test to be performed
within 15 years from the last ILRT which was performed in April 1995.
Date of Issuance: August 31, 2005.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76492). The supplement contained clarifying information only, and
determination or expand the scope of the initial Federal Register
in a Safety Evaluation dated August 31, 2005.
Date of application for amendments: January 21, 2005.
Brief description of amendments: The amendments modify the
Isolation Condenser System heat removal capability surveillance
requirement (SR) by adding a note to the technical specification
section SR 3.5.3.4. This note allows a delay of 12 hours after adequate
reactor power is achieved to perform the test.
Amendment Nos.: 215,207.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
Date of initial notice in Federal Register: May 24, 2005.
Date of application for amendments: May 20, 2004, as supplemented
by letters dated February 18 and July 13, 2005.
Limerick Generating Station Units 1 and 2 Technical Specifications
(TSs) 2.2.1, ``Reactor Protection System Instrumentation Setpoints,''
TS 3/4.3.1, ``Reactor Protection System Instrumentation,'' TS 3/4.3.6,
``Control Rod Block Instrumentation,'' TS 3/4.4.1, ``Recirculation
System,'' and TS 6.9.1, ``Routine Reports,'' and the associated TS
Bases. The amendments support activation of the trip outputs of the
oscillation power range monitor portion of the power range neutron
Date of issuance: August 26, 2005.
Amendment Nos.: 177 and 139.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62474). The supplements dated February 18 and July 13, 2005, provided
in a Safety Evaluation dated August 26, 2005.
Date of application for amendment: August 24, 2004, as supplemented
Brief description of amendment: This amendment revises Technical
Specifications (TSs) related to the surveillance requirements for the
emergency feedwater system.
Date of issuance: August 16, 2005.
[[Page 54092]]
Renewed Facility Operating License No. NPF-12: Amendment revises
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60685).
in a Safety Evaluation dated August 16, 2005.
Date of application for amendments: June 30, 2004, as supplemented
by letters dated December 2, 2004, May 27, 2005, and July 18, 2005.
Brief description of amendments: The proposed changes revise
Technical Specification 5.5.2.15, ``Containment Leakage Rate Testing
Program,'' to include a one-time extension of the 10-year period of the
performance-based leakage rate testing program for Type A tests as
prescribed by the Nuclear Energy Institute (NEI) 94-01, Revision 0,
``Industry Guideline for Implementing Performance-Based Option of 10
CFR part 50, Appendix J.''
Date of issuance: August 24, 2005.
Amendment Nos.: 198 and 189.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46589). The supplemental letters dated December 2, 2004, May 27, and
July 18, 2005, provided information that clarified the application, did
not change the NRC staff's original proposed no significant hazards
in a Safety Evaluation dated August 24, 2005.
of publication of this notice, the licensee may file a request for a
hearing with respect to
[[Page 54093]]
request for a hearing and a petition for leave to intervene. Requests
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
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
officer or the Atomic Safety and Licensing Board that the petition,
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: August 21, 2005, as supplemented August
Description of amendment request: The amendments revise Technical
Specification Limiting Condition for Operation 3.8.1, Condition C.2.1,
to permit a one-time extension of 96 hours of the Completion Times for
Keowee Hydro Unit 2.
[[Page 54094]]
Amendment Nos.: 347, 349, and 348.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revises the technical specifications.