Source: https://regulations.justia.com/regulations/fedreg/2005/10/11/05-20168.html
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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 59082-59096 [05-20168] :: 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, 59082-59096 [05-20168]
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 59082-59096 [05-20168]
Download as PDF 59082 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices available in the NRC Public Document Room (PDR), One White Flint North, Room O–1F21, 11555 Rockville Pike, Rockville, MD 20852–2738. ACNW meeting agenda, transcripts, and letter reports are available through the NRC Public Document Room at pdr@nrc.gov, by calling the PDR at 1–800–394–4209, or from the Publicly Available Records System (PARS) component of NRC’s document system (ADAMS) which is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/ adams.html or http://www.nrc.gov/ reading-rm/doc-collections/. A copy of the certified minutes of the meeting will be available at the same location up to three months following the meeting. Copies may be obtained upon payment of appropriate reproduction charges. (f) Video teleconferencing service is available for observing open sessions of some ACNW meetings. Those wishing to use this service for observing ACNW meetings should contact Mr. Theron Brown, ACNW Audio Visual Technician, (301–415–8066) between 7:30 a.m. and 3:45 p.m. Eastern Time at least 10 days before the meeting to ensure the availability of this service. Individuals or organizations requesting this service will be responsible for telephone line charges and for providing the equipment and facilities that they use to establish the video teleconferencing link. The availability of video teleconferencing services is not guaranteed. (g) The meeting room is handicapped accessible. ACNW Working Group Meetings From time to time the ACNW may sponsor an in-depth meeting on a specific technical issue to understand staff expectations and review work in progress. Such meetings are called Working Group meetings. These Working Group meetings will also be conducted in accordance with these procedures noted above for the ACNW meeting, as appropriate. When Working Group meetings are held at locations other than at NRC facilities, reproduction facilities may not be available at a reasonable cost. Accordingly, 50 additional copies of the materials to be used during the meeting should be provided for distribution at such meetings. Special Provisions When Proprietary Sessions Are To Be Held If it is necessary to hold closed sessions for the purpose of discussing matters involving proprietary information, persons with agreements permitting access to such information may attend those portions of the ACNW VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 meetings where this material is being discussed upon confirmation that such agreements are effective and related to the material being discussed. The DFO should be informed of such an agreement at least five working days prior to the meeting so that it can be confirmed, and a determination can be made regarding the applicability of the agreement to the material that will be discussed during the meeting. The minimum information provided should include information regarding the date of the agreement, the scope of material included in the agreement, the project or projects involved, and the names and titles of the persons signing the agreement. Additional information may be requested to identify the specific agreement involved. A copy of the executed agreement should be provided to the DFO prior to the beginning of the meeting for admittance to the closed session. Dated: October 5, 2005. Annette L. Vietti-Cook, Secretary of the Commission. [FR Doc. 05–20317 Filed 10–7–05; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from September 6, 2005, to September 29, 2005. The last biweekly notice was published on September 27, 2005 (70 FR 56499). PO 00000 Frm 00055 Fmt 4703 Sfmt 4703 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 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. E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 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 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 VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor’s/petitioner’s right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor’s/petitioner’s property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor’s/petitioner’s interest. The petition must also set forth the specific contentions which the petitioner/ requestor seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/ requestor to relief. A petitioner/ requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of PO 00000 Frm 00056 Fmt 4703 Sfmt 4703 59083 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). 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. E:\FR\FM\11OCN1.SGM 11OCN1 59084 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices Carolina Power & Light Company, et al., Docket No. 50–400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: April 6, 2005, as supplemented by letter dated August 8, 2005. Description of amendment request: The proposed amendment will modify Technical Specification (TS) 6.8.4.k, ‘‘Containment Leakage Rate Testing Program,’’ and TS Surveillance Requirement (SR) 4.6.1.6.1, ‘‘Containment Vessel Surfaces.’’ The proposed amendment would modify the TS to allow for a one-time extension of the containment Type A test interval from once in 10 years to once in 15 years. 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 change does not involve a significant hazards consideration for the following reasons: 1. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change to HNP [Harris Nuclear Plant] TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval from 10 years to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME [American Society of Mechanical Engineers] Section XI Code. The existing 10-year test interval is based on past test performance. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment vessel is designed to provide a leak-tight barrier against the uncontrolled release of radioactivity to the environment in the unlikely event of postulated accidents. As such, the containment vessel is not considered as the initiator of an accident. Therefore, the proposed TS change does not involve a significant increase in the probability of an accident previously evaluated. The proposed change involves only a onetime change to the interval between containment Type A tests. Type B and C leakage testing will continue to be performed at the intervals specified in 10 CFR Part 50, Appendix J, Option A, as required by the HNP TS. As documented in NUREG–1493, ‘‘Performance-Based Containment LeakageTest Program,’’ industry experience has shown that Type B and C containment leak rate tests have identified a very large percentage of containment leak paths, and that the percentage of containment leak paths that are detected only by Type A testing is very small. In fact, an analysis of 144 VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 integrated leak rate tests, including 23 failures, found that none of the failures involved a containment liner breach. NUREG–1493 also concluded, in part, that reducing the frequency of containment Type A testing to once per 20 years results in an imperceptible increase in risk. The HNP test history and risk-based evaluation of the proposed extension to the Type A test interval supports this conclusion. The design and construction requirements of the containment vessel, combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a high degree of assurance that the containment vessel will not degrade in a manner that is detectable only by Type A testing. Therefore, the proposed TS change does not involve a significant increase in the consequences of an accident previously evaluated. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME Section XI Code. The existing 10year test interval is based on past test performance. The proposed change to the Type A test interval does not result in any physical changes to HNP. In addition, the proposed test interval extension does not change the operation of HNP such that a failure mode involving the possibility of a new or different kind of accident from any accident previously evaluated is created. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed amendment does not involve a significant reduction in a margin of safety. The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval from 10 years to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME Section XI Code. The existing 10-year test interval is based on past test performance. The NUREG–1493 study of the effects of extending containment leak rate testing found that a 20 year extension for Type A testing resulted in an imperceptible increase in risk to the public. NUREG–1493 found that, generically, the design containment leak rate contributes a very small amount to the individual risk and that the decrease in Type A testing frequency would have a minimal affect on this risk since most potential leak paths are detected by Type B and C testing. The proposed change involves only a one-time extension of the interval for containment Type A testing; PO 00000 Frm 00057 Fmt 4703 Sfmt 4703 the overall containment leak rate specified by the HNP TS is being maintained. Type B and C testing will continue to be performed at the frequency required by the HNP TS. The regular containment inspections being performed in accordance with the ASME Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a high degree of assurance that the containment will not degrade in a manner that is only detectable by Type A testing. In addition, a plantspecific risk evaluation has demonstrated that the one-time extension of the Type A test interval from 10 years to 15 years results in a very small increase in risk for those accident sequences influenced by Type A testing. Therefore, this change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the 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, et al., Docket No. 50–400, Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: June 20, 2005. Description of amendment request: The amendment would revise Technical Specifications (TS) 3/4.4.7, ‘‘Reactor Coolant System Chemistry.’’ Specifically, the proposed amendment would revise the footnotes in Tables 3.4–2 and 4.4–3 of the TS to increase the temperature limit from 180 °F to 250 °F above which reactor coolant sampling and analysis for dissolved oxygen is required and dissolved oxygen limits apply. 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 amendment does not involve a significant hazards consideration for the following reasons: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Operation of HNP in accordance with the proposed amendment does not increase the E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices probability or consequences of accidents previously evaluated. The Final Safety Analysis Report (FSAR) documents the analyses of design basis accidents (DBA) at HNP. Any scenario or previously analyzed accident that results in offsite dose were evaluated as part of this analysis. The proposed amendment does not change or affect any accident previously evaluated in the FSAR. The proposed amendment does not modify any plant equipment. In addition, the proposed amendment does not result in a change to a structure, system, or component (SSC), or adversely affect its design function. The purpose of the temperature limit for RCS [Reactor Coolant System] oxygen control is to minimize corrosion at high temperatures on RCS components. Increasing the temperature at which oxygen levels are required to be maintained within specified limits from 180 °F to 250 °F is supported by industry and vendor data which indicates that the influence of dissolved oxygen at or below 250 °F is not significant with regard to stress corrosion cracking and general corrosion of RCS components. The proposed amendment is consistent with the Electric Power Research Institute’s (EPRI’s) guidelines for Pressurized Water Reactor (PWR) Primary Water Chemistry. This amendment places HNP in line with standard industry specifications for reactors of similar size and vintage. HNP’s proposed amendment to increase the temperature limit for applicability to 250 °F would decrease the time needed to achieve compliance with the dissolved oxygen limit and decrease the overall time to restart the plant from cold shutdown. Removing oxygen in a more expeditious fashion enhances RCS chemistry. Based on the above, RCS integrity is maintained by this amendment. Therefore, this amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Operation of HNP in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The FSAR documents the analyses of design basis accidents (DBA) at HNP. Any scenario or previously analyzed accident that results in offsite dose were evaluated as part of this analysis. The proposed amendment does not change or affect any accident previously evaluated in the FSAR, and no new or different scenarios are created by the proposed amendment to the TS. The proposed amendment does not modify any plant equipment. In addition, the proposed amendment does not result in a change to an SSC [structure, system, or component] or adversely affect its design function. The purpose of the temperature limit for RCS oxygen control is to minimize corrosion at high temperatures on RCS components. Increasing the temperature at which oxygen levels are required to be maintained within specified limits from 180 °F to 250 °F is supported by industry and vendor data VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 which indicates that the influence of dissolved oxygen at or below 250 °F is not significant with regard to stress corrosion cracking and general corrosion of RCS components. The proposed amendment is consistent with EPRI’s guidelines for PWR Primary Water Chemistry. This amendment places HNP in line with standard industry specifications for reactors of similar size and vintage. HNP’s proposed amendment to increase the temperature limit for applicability to 250 °F would decrease the time needed to achieve compliance with the dissolved oxygen limit and decrease the overall time to restart the plant from cold shutdown. Removing oxygen in a more expeditious fashion enhances RCS chemistry. Based on the above, RCS integrity is maintained by this amendment. Therefore, this amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. Operation of HNP in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. Existing TS operability and surveillance requirements are not reduced by the proposed amendment. The proposed amendment does not modify any plant equipment. In addition, the proposed amendment does not result in a change to a structure, system, or component (SSC), or its design function. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the FSAR. The purpose of the temperature limit for RCS oxygen control is to minimize corrosion at high temperatures on RCS components. Increasing the temperature at which oxygen levels are required to be maintained within specified limits from 180 °F to 250 °F is supported by industry and vendor data which indicates that the influence of dissolved oxygen at or below 250 °F is not significant with regard to stress corrosion cracking and general corrosion of RCS components. The proposed amendment is consistent with EPRI’s guidelines for PWR Primary Water Chemistry. This amendment places HNP in line with standard industry specifications for reactors of similar size and vintage. HNP’s proposed amendment to increase the temperature limit for applicability to 250 °F would decrease the time needed to achieve compliance with the dissolved oxygen limit and decrease the overall time to restart the plant from cold shutdown. Removing oxygen in a more expeditious fashion enhances RCS chemistry. Based on the above, RCS integrity is maintained by this amendment. Therefore, this amendment does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the PO 00000 Frm 00058 Fmt 4703 Sfmt 4703 59085 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. Nebraska Public Power District, Docket No. 50–298, Cooper Nuclear Station, Nemaha County, Nebraska Date of amendment request: June 20, 2005. Description of amendment request: The proposed amendment would revise Cooper Nuclear Station (CNS) Technical Specification (TS) 5.3, ‘‘Unit Staff Qualifications,’’ to upgrade the qualification standard for the Shift Manager, Senior Operator, Licensed Operator, and Shift Technical Engineer from Regulatory Guide (RG) 1.8, Revision 2 ‘‘Qualification and Training of Personnel for Nuclear Power Plants,’’ to RG 1.8, Revision 3. It also clarifies qualification requirements applicable to the Operations Manager position. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. These changes are administrative in nature and do not require any physical modifications, affect any plant components, or result in any changes in plant operation. They provide clarity and consistency to the CNS licensing basis. Upgrading the unit staff qualifications for the Shift Manager, Senior Operator, Licensed Operator, and Shift Technical Engineer from Regulatory Guide 1.8, Revision 2, to Regulatory Guide 1.8, Revision 3, is an administrative change that will clarify the current requirements for qualification and training of operations personnel. The changes are consistent with the application of a systems approach to training in an accredited training program. By promulgation of the 10 CFR Part 55 rule change, the NRC determined that an accredited licensed operator training program based on a systems approach to training provides an acceptable means of qualifying licensed operating personnel. The addition of qualification requirements for the Operations Manager position clarifies SRO [Senior Reactor Operator] license requirements for Operations management personnel by specifying that the Operations Supervisor is the member of Operations management required to have a current SRO license at CNS. The Operations Manager is required to hold or have previously held a E:\FR\FM\11OCN1.SGM 11OCN1 59086 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices SRO license. This will ensure an acceptable level of operations knowledge to perform in a managerial oversight role. This approach is consistent with current guidance in ANSI/ ANS [American Nuclear Standards Institute/ American Nuclear Society] 3.1–1993. This change is administrative in nature and has no impact on previously evaluated accidents. Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. These changes are administrative in nature and do not involve a physical alteration of the plant or a change to plant operations. No new failure mechanisms, malfunctions, or accident initiators are introduced. The proposed changes provide clarity and consistency to the CNS licensing basis in regard to training and qualification of operations personnel and SRO license requirements for Operations management personnel. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Response: No. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. These changes are administrative in nature and do not affect any Technical Specification safety limit or limiting condition for operation. No safety margins are affected by these changes. The proposed changes do not involve a change in plant design or operation for the mitigation of postulated accidents. The proposed changes provide clarity and consistency to the CNS licensing basis in regard to training and qualification of operations personnel and SRO license requirements for Operations management personnel. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602–0499. NRC Section Chief: David Terao. Nebraska Public Power District, Docket No. 50–298, Cooper Nuclear Station, Nemaha County, Nebraska Date of amendment request: August 25, 2005. Description of amendment request: The proposed amendment would revise the definitions of Channel Calibration, VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Channel Function Test, and Logic System Functional Test in accordance with the Technical Specification Task Force Traveler 205–A. 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: done. Revising these definitions as proposed will not result in a change to the design or operation of any plant SSC used to shutdown the plant, initiate the Emergency Core Cooling Systems, or isolate primary or secondary containment. As a result the ability of the plant to respond to and mitigate accidents is unchanged by the revised definitions. Based on the above, NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety. 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The definitions of Channel Calibration, Channel Functional Test, and Logic System Functional Test specified in Technical Specifications (TS) provide basic information regarding what the test involves, the components involved in the test, and general information regarding how the test is to be performed. These definitions and their specific wording are not precursors to any accident. As a result these revised definitions result in no increase in the probability of an accident previously evaluated. The proposed revisions of these definitions involve no changes to plant design, equipment, or operation related to mitigation of accidents. The proposed revisions of these definitions do not change their meaning or intent. The proposed revisions clarify the definitions and do not result in a reduction of required testing of instrumentation used to mitigate accidents. Based on the above NPPD [Nebraska Public Power District] concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed revisions of the definitions do not involve a change to the design or operation of any plant structure, system, or component (SSC). As a result the plant will continue to be operated in the same manner. The proposed revisions will not result in a change to how the instrumentation used to monitor plant operation and to mitigate accidents is tested. Operating the plant and testing the plant’s instrumentation in the same manner as is currently done will not create the possibility of a new or different kind of accident. Based on the above NPPD concludes that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The affected definitions involve testing of instrumentation used in the mitigation of accidents to ensure that the instrumentation will perform as assumed in safety analyses. The proposed revisions of these definitions will not change their meaning or intent. As a result, the instrumentation will continue to be tested in the same manner as is currently The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602–0499. NRC Section Chief: David Terao. PO 00000 Frm 00059 Fmt 4703 Sfmt 4703 Pacific Gas and Electric Company, Docket Nos. 50–275 and 50–323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of amendment requests: July 29, 2005. Description of amendment requests: The proposed amendments would revise Technical Specification 3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ to change the frequency of Surveillance Requirement 3.7.5.6 from 92 days to 24 months. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change to increase [the] frequency interval for Surveillance Requirement (SR) 3.7.5.6 from 92 days to 24 months has no impact on the probability of accidents previously evaluated. The valves controlled by SR 3.7.5.6 are used to provide an alternate supply of water to the auxiliary feedwater (AFW) system from the fire water storage tank (FWST) and are only operated after an accident has occurred. They are not accident initiators. Misoperation, or failure of a[n] FWST supply to be correctly positioned following an accident, could result in an inadequate supply of water to the AFW system. Failure to provide adequate core cooling could increase the radiological consequences of an accident. However, operating and maintenance histories of the FWST supply valves show that these valves have been E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices capable of full stroke cycling each time they have been tested. There is no evidence of any time-related degradation mechanism that would prevent the valves from performing their design function. Thus[,] the proposed change has no impact on the consequences of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different [kind of] accident from any accident previously evaluated? Response: No. The proposed change to increase frequency interval for SR 3.7.5.6 from 92 days to 24 months has no impact on the probability of accidents of the type evaluated in the Final Safety Analysis Report, as updated. The valves are used to provide an alternate supply of water to the AFW system from the FWST, and are only operated after an accident has occurred. They are not accident initiators. Review of the operating and maintenance histories of the FWST supply valves show that they are highly reliable in maintaining their capability to perform their design function. Therefore, the proposed change does not create the possibility of a new or different [kind of] accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change to SR 3.7.5.6 involves only an increase in the frequency interval. No physical changes are required to the facility or to the plant operating or emergency procedures as a result of the change. Based on review of the operating and maintenance histories of the FWST supply valves, they have been capable of full stroke cycling each time they have been tested. There is no evidence of any time-related degradation mechanism that would prevent the valves from performing their design function. This evidence supports the conclusion that there will be no impact in the operation of these valves following an accident. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120. NRC Section Chief: Daniel S. Collins (Acting). VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Pacific Gas and Electric Company, Docket Nos. 50–275 and 50–323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of amendment requests: August 23, 2005. Description of amendment requests: The proposed amendments would revise the expiration dates of the Units 1 and 2 facility-operating licenses to recapture low-power testing time, and to reflect a 40-year term measured from the date of issuance of each unit’s fullpower operating license. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed additional operating license periods do not affect the probability or consequences of an accident previously evaluated since they require no physical change in the plant equipment or operating procedures and the Final Safety Analysis Report (FSAR) Update safety analyses are based on [a] 40-year full[-]power operation. Surveillance and maintenance practices, as well as other programs such as environmental qualification of equipment, ensure timely identification and correction of any degradation of safety-related plant equipment. The long-term integrity of the reactor vessels has been evaluated using currently acceptable NRC calculational methods and best available Diablo Canyon Power Plant (DCPP) specific data. The evaluation results demonstrate that both reactor vessels are safe for normal operations in excess of 40 years. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different [kind of] accident from any accident previously evaluated? Response: No. The possibility of a new or different kind of accident is not created by the proposed additional operating periods since at least 40 years of full[-]power operation was assumed in the design and construction of DCPP Units 1 and 2. The plant maintenance programs are also designed to both maintain and determine the need to replace safety-related components. These programs will continue to be applied as they are presently to assure safe operation. Therefore, the proposed change does not create the possibility of a new or different [kind of] accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? PO 00000 Frm 00060 Fmt 4703 Sfmt 4703 59087 Response: No. The proposed additional operating periods do not involve a significant reduction in a margin of safety since, as is the case with present operation, degradation of safetyrelated equipment will be identified and corrected by ongoing surveillance and maintenance practices. Existing programs, routine maintenance, and compliance with Technical Specifications assure that an adequate margin of safety is maintained. These activities will remain in effect for the duration of the proposed additional operating periods. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120. NRC Section Chief: Daniel S. Collins (Acting). 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 amendment request: June 30, 2005. Description of amendment request: The proposed changes would revise the Administrative Control section of the Technical Specifications (TSs) to permit the Westinghouse best estimate methodology for loss-of-coolantaccident (LOCA) analysis methodology to be utilized for analyses as required by Title 10 of the Code of Federal Regulations, Part 50, Section 46, ‘‘Acceptance criteria for emergency core cooling systems [ECCS] for light water nuclear power reactors’ (10 CFR 50.46). 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. Implementation of the best-estimate large break LOCA methodology and associated TS changes is proposed to increase margin to the peak clad temperature limits defined in 10 CFR 50.46. There are no physical plant changes or changes in manner in which the plant will be operated as a result of this E:\FR\FM\11OCN1.SGM 11OCN1 59088 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices change. Since the plant conditions and ECCS performance assumed in the analysis are consistent with the plant’s current design, the proposed change in methodology will thus have no impact on the probability of a LOCA. When applied, the best estimate methodology shows that the ECCS is more effective than previously evaluated in mitigating the consequences of a LOCA, as lower peak clad temperatures are predicted relative to current 10 CFR 50.46 Appendix K results. Since the proposed best-estimate methodology is only applicable to a large break LOCA and since the application of the proposed methodology shows there is a high probability that all of the acceptance criteria contained in 10 CFR 50.46, Paragraph b are met, the proposed change does not increase the consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. There are no physical changes being made to the plant. No new modes of plant operation are being introduced. The parameters assumed in the analysis remain within the design limits of the existing plant equipment. All plant systems will perform as designed during the response to a potential accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed. 3. Does this change involve a significant reduction in a margin of safety? Response: No. It has been shown that the methodology used in the analysis would more realistically describe the expected behavior of V. C. Summer Nuclear Station systems during a postulated loss of coolant accident. Uncertainties have been accounted for as required by 10 CFR 50.46. A sufficient number of loss of coolant accidents with different break sizes, different locations and other variations in properties are analyzed to provide assurance that the most severe postulated loss of coolant accidents are calculated. It has been shown by analysis that there is a high level of probability that all criteria contained in 10 CFR 50.46, Paragraph b are met. Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed Technical Specifications change poses no significant hazard as delineated by 10 CFR 50.92. 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: Thomas G. Eppink, South Carolina Electric & Gas VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Company, Post Office Box 764, Columbia, South Carolina 29218. NRC Section Chief: Evangelos C. Marinos. STP Nuclear Operating Company, Docket Nos. 50–498 and 50–499, South Texas Project, Units 1 and 2, Matagorda County, Texas. Date of amendment request: August 30, 2005. Description of amendment request: The proposed amendment would change the Technical Specifications (TSs) to reflect the use of the Westinghouse Best Estimate Analyzer for Core Operations—Nuclear (BEACON) to augment the functional capability of the flux mapping system for the purpose of power distribution surveillances. In addition, editorial changes to the TSs are proposed. 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 PDMS [power distribution monitoring system] performs continuous core power distribution monitoring. This system utilizes the NRC-approved Westinghouse proprietary computer code BEACON to provide data reduction for incore flux maps, core parameter analysis, load follow operation simulation, and core prediction. It in no way provides any protection or control system function. Fission product barriers are not impacted by these proposed changes. The proposed changes occurring with PDMS will not result in any additional challenges to plant equipment that could increase the probability of any previously evaluated accident. The changes associated with the PDMS do not affect plant systems such that their function in the control of radiological consequences is adversely affected. These proposed changes will therefore not affect the mitigation of the radiological consequences of any accident described in the Updated Final Safety Analysis Report Update (UFSAR). Continuous on-line monitoring through the use of PDMS provides significantly more information about the power distributions present in the core than is currently available. This results in more time (i.e., earlier determination of an adverse condition developing) for operator action prior to having an adverse condition develop that could lead to an accident condition or to unfavorable initial conditions for an accident. Each accident analysis addressed in the UFSAR is examined with respect to changes in cycle-dependent parameters, which are obtained from application of the NRC- PO 00000 Frm 00061 Fmt 4703 Sfmt 4703 approved reload design methodologies, to ensure that the transient evaluations of reload cores are bounded by previously accepted analyses. This examination, which is performed in accordance with the requirements set forth in 10 CFR [Title 10 of the Code of Federal Regulations] 50.59, ensures that future reloads will not involve a significant increase in the probability or consequences of any accident previously evaluated. The three editorial changes only correct typographical errors made in previously approved TS changes. They do not affect plant operation or structures, systems, and components important to safety. Therefore, the proposed changes do not involve a significant increase in the probability or consequence 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 implementation of the PDMS has no influence or impact on plant operations or safety, nor does it contribute in any way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. The possibility for a new or different type of accident from any accident previously evaluated is not created since the changes associated with implementation of the PDMS do not result in a change to the design basis of any plant component or system. The evaluation of the effects of using the PDMS to monitor core power distribution parameters shows that all design standards and applicable safety criteria limits are met. The proposed changes do not result in any event previously deemed incredible being made credible. Implementation of the PDMS will not result in more adverse conditions and will not result in any increase in the challenges to safety systems. The cyclespecific variables required by the PDMS are calculated using NRC-approved methods. The TS will continue to require operation within the required core operating limits and appropriate actions will be taken if limits are exceeded. The three editorial changes only correct typographical errors made in previously approved TS changes. They do not affect plant operation or structures, systems, and components important to safety. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The margin of safety is not affected by implementation of the PDMS. The margin of safety provided by current TS is unchanged. The proposed changes continue to require operation within the core limits that are based on NRC-approved reload design methodologies. Appropriate measures exist to control the values of these cycle-specific limits. The proposed changes continue to ensure that appropriate actions will be taken E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices if limits are violated. These actions remain unchanged. The three editorial changes only correct typographical errors made in previously approved TS changes. They do not affect plant operation or structures, systems, and components important to safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004. NRC Section Chief: David Terao. Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. AmerGen Energy Company, LLC, Docket No. 50–461, Clinton Power Station, Unit 1, DeWitt County, Illinois Date of application for amendment: April 3, 2003, as supplemented December 23, 2003, December 9 and 17, 2004, and March 30 and August 19, 2005. Brief description of amendment: The amendment revised the Technical Specifications (TSs) to support the application of an alternative source term methodology in accordance with Title 10 of the Code of Federal Regulations, Section 50.67, ‘‘Accident Source Term,’’ with the exception that Technical Information Document 14844, ‘‘Calculation of Distance Factors for Power and Test Reactor Sites,’’ was used as the radiation dose basis for equipment qualification. Date of issuance: September 19, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 167. Facility Operating License No. NPF– 62: The amendment revised the TSs. Date of initial notice in Federal Register: September 2, 2003 (68 FR 52234). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 19, 2005. The supplements dated December 23, 2003, December 9 and 17, 2004, and March 30 and August 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. No significant hazards consideration comments received: No. PO 00000 Frm 00062 Fmt 4703 Sfmt 4703 59089 AmerGen Energy Company, LLC, Docket No. 50–461, Clinton Power Station, Unit 1, DeWitt County, Illinois Date of application for amendment: November 11, 2003, as supplemented April 16 and September 10, 2004, and March 30 and September 21, 2005. Brief description of amendment: The amendment revised the instrument channel trip setpoint allowable values for thirteen Technical Specification (TS) functions at Clinton Power Station, Unit 1. Date of issuance: September 27, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 168. Facility Operating License No. NPF– 62: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: March 16, 2004 (69 FR 12363). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2005. The supplements dated April 16 and September 10, 2004, and March 30 and September 21, 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. No significant hazards consideration comments received: No. Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50–317 and 50–318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of application for amendments: August 3, 2004, as supplemented on July 8 and August 26, 2005. Brief description of amendments: The amendments extend the surveillance frequency interval from monthly to quarterly for Technical Specification surveillance requirement (SR) 3.3.3.1, which involves a channel functional test of each reactor trip circuit breaker (RTCB). SRs 3.3.3.1 and 3.3.3.2 will be scheduled such that the RTCBs testing is performed every 6 weeks, which meets the vendor-recommended interval for cycling each RTCB. Date of issuance: September 26, 2005. Effective date: As of the date of issuance to be implemented within 60 days. Amendment Nos.: 275 and 252. Renewed Facility Operating License Nos. DPR–53 and DPR–69: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: January 4, 2005 (70 FR 400). E:\FR\FM\11OCN1.SGM 11OCN1 59090 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices The July 8 and August 26, 2005, supplemental letters 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 these amendments is contained in a Safety Evaluation dated September 26, 2005. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., Docket No. 50–245, Millstone Power Station Unit No. 1, New London County, Connecticut Date of application for amendment: September 8, 2004, as supplemented by letters dated May 5 and July 27, 2005. Brief description of amendment: The amendment revised the Millstone Power Station, Unit No. 1 Technical Specifications (TSs) to support the implementation of the proposed Dominion Nuclear Facility Quality Assurance Program (Topical Report DOM–QA–1). Implementation of this Topical Report would create a common quality assurance program for all sites owned by Dominion Nuclear Connecticut, Inc. Review of this proposed amendment was requested in concert with the review of the Topical Report. Date of issuance: September 15, 2005. Effective date: As of the date of issuance, and shall be implemented by February 28, 2006. Amendment No.: 115. Facility Operating License No. DPR– 21: The amendment revised the TSs. Date of initial notice in Federal Register: January 18, 2005 (70 FR 2888). The additional information provided in the supplemental letters dated May 5 and July 27, 2005, did not expand the scope of the application as noticed and did not change the NRC staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2005. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., Docket No. 50–336, Millstone Power Station, Unit No. 2, New London County, Connecticut Date of application for amendment: July 15, 2004, as supplemented by letter dated August 23, 2004. Brief description of amendment: The amendment revised the Facility VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Operating License DPR–65 to address the resolution of a non-conservative Technical Specifications (TSs) associated with control room isolation radiation monitoring instrumentation. Specifically, the amendment would revise the TSs to require two operable channels of control room isolation radiation monitoring instrumentation. Date of issuance: September 23, 2005. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment No.: 289. Facility Operating License No. DPR– 65: The amendment revised the TSs. Date of initial notice in Federal Register: January 18, 2005 (70 FR 2887). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 23, 2005. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., et al., Docket No. 50–423, Millstone Power Station, Unit No. 3, New London County, Connecticut Date of application for amendment: April 15, 2004, as supplemented on June 23, 2005. Brief description of amendment: The amendment approves modifications to the Fire Protection Program. Specifically, the modifications involve converting the existing automatic carbon dioxide fire suppression systems installed in the cable spreading room to manual actuation. Date of issuance: September 22, 2005. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 227. Facility Operating License No. NPF– 49: The amendment allows for conversion from an automatic to a manual carbon dioxide suppression system in the cable spreading area. Date of initial notice in Federal Register: July 6, 2004 (69 FR 40672). The supplement dated June 23, 2005, provided clarifying information and did not change the initial proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 22, 2005. No significant hazards consideration comments received: No. PO 00000 Frm 00063 Fmt 4703 Sfmt 4703 Dominion Nuclear Connecticut, Inc., Docket Nos. 50–336 and 50–423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut Date of application for amendments: September 8, 2004, as supplemented by letters dated May 5 and July 27, 2005. Brief description of amendments: The amendments revised the Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications (TSs) to support the implementation of the proposed Dominion Nuclear Facility Quality Assurance Program (Topical Report DOM–QA–1). Implementation of this Topical Report would create a common quality assurance program for all sites owned by Dominion Nuclear Connecticut, Inc. Review of these proposed amendments was requested to be done in concert with the review of the Topical Report. Date of issuance: September 15, 2005. Effective date: As of the date of issuance, and shall be implemented by February 28, 2006. Amendment Nos.: 288 and 226. Facility Operating License Nos. DPR– 65 and NPF–49: The amendments revised the TSs. Date of initial notice in Federal Register: January 18, 2005 (70 FR 2888). The additional information provided in the supplemental letters dated May 5, and July 27, 2005, did not expand the scope of the application as 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 September 15, 2005. No significant hazards consideration comments received: No. 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 6, 2004, as supplemented on February 16, and August 9, 2005. Brief description of amendment: The amendment revised Technical Specification (TS) surveillance requirement 4.5.B.1 related to air testing of the drywell spray headers and nozzles. Specifically, the amendment changes the test frequency from once every five years to following maintenance that could result in nozzle blockage. Date of Issuance: September 20, 2005. Effective date: As of the date of issuance, and shall be implemented within 60 days. E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices Amendment No.: 228. 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 supplements 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 September 20, 2005. No significant hazards consideration comments received: No. Date of issuance: September 19, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 138. Facility Operating License Nos. NPF– 72 and NPF–77: The amendments revised the Environmental Protection Plan. Date of initial notice in Federal Register: April 12, 2005 (70 FR 19115). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 19, 2005. No significant hazards consideration comments received: No. Entergy Operations, Inc., Docket No. 50– 313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Exelon Generation Company, LLC, Docket No. STN 50–455, Byron Station, Unit No. 2, Ogle County, Illinois Date of amendment request: September 30, 2004, as supplemented by letter dated May 20, 2005. Brief description of amendment: The amendment revises the Technical Specifications to allow the use of M5 fuel cladding and of Mark-B-high thermal performance fuel in Arkansas Nuclear One, Unit 1, during its fuel Cycle 20 and beyond. Date of issuance: September 12, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 226. Renewed Facility Operating License No. DPR–51: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: November 9, 2004 (69 FR 64988). The supplement dated May 20, 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 September 12, 2005. No significant hazards consideration comments received: No. Date of application for amendment: May 24, 2005. Brief description of amendment: The amendment modifies the inspection requirements for portions of the steam generator (SG) tubes within the hot leg tubesheet region of the SGs. Date of issuance: September 19, 2005. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment No.: 144. Facility Operating License No. NPF– 66: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 5, 2005 (70 FR 38718). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 19, 2005. No significant hazards consideration comments received: No. Exelon Generating Company, LLC, Docket Nos. STN 50–456 and STN 50– 457, Braidwood Station, Units 1 and 2, Will County, Illinois Date of application for amendment: December 17, 2004. Brief description of amendment: The amendments revised Appendix B, Environmental Protection Plan (nonradiological), of the Braidwood Station Facility Operating Licenses. VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Exelon Generation Company, LLC, Docket Nos. 50–237 and 50–249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Docket Nos. 50–254 and 50–265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of application for amendments: February 27, 2004, as supplemented by letters dated October 11, 2004, January 3, 2005, August 11, 2005, and September 12, 2005. Brief description of amendments: The amendments add the Oscillation Power Range Monitor (OPRM) instrumentation to the Technical Specifications. Date of issuance: September 22, 2005. Effective date: As of the date of issuance and shall be implemented by December 31, 2005. Amendment Nos.: 227, 222. Facility Operating License Nos. DPR– 19, DPR–25, DPR–29 and DPR–30. The PO 00000 Frm 00064 Fmt 4703 Sfmt 4703 59091 amendments revised the Technical Specifications. Date of initial notice in Federal Register: December 7, 2004 (69 FR 70718). The October 11, 2004, and January 3, 2005, August 11, 2005, and September 12, 2005, submittals provided clarifying information that did not change the initial proposed no significant hazards consideration. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated: September 22, 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: July 22, 2004, as supplemented December 3, 2004, and September 20, 2005. The September 20, 2005, supplement withdrew a portion of the original application from consideration. Brief description of amendments: The amendments modified the operability and surveillance requirements in Technical Specification (TS) 3/4.1.3, ‘‘Control Rods.’’ Specifically, the changes (1) exclude a fully-inserted immovable control rod from the shutdown action statement, and (2) limit the 24-hour exercise test of other control rods to a one-time occasion following detection of an immovable control rod. Date of issuance: September 27, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 178 and 140. Facility Operating License Nos. NPF– 39 and NPF–85. The amendments revised the TSs. Date of initial notice in Federal Register: May 24, 2005 (70 FR 29794). The September 20, 2005, supplement withdrew a portion of the original application from consideration and did not change the proposed no significant hazards consideration. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 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: June 1, 2004. E:\FR\FM\11OCN1.SGM 11OCN1 59092 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices Brief description of amendments: The amendments relocate the operability and surveillance requirements for the reactor coolant system safety/relief valve position instrumentation from the Limerick Generating Station (LGS) Technical Specifications (TSs) to the LGS Technical Requirements Manual (TRM) and plant procedures. Specifically, the amendments relocate TSs 3.4.2.c, 4.4.2.1, and the associated footnotes to the TRM. Additionally, the ‘‘Safety/Relief Valve Position Indicators’’ instrumentation is relocated from Tables 3.3.7.5–1 and 4.3.7.5–1 of TSs 3.3.7.5 and 4.3.7.5, respectively, to the TRM. Date of issuance: September 27, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 179 and 141. 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 62475). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 2005. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50–334 and 50–412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS–1 and 2), Beaver County, Pennsylvania Date of application for amendments: June 2, 2004, as supplemented February 23 and August 19, 2005. Brief description of amendments: The amendments revised the BVPS–1 and 2, Technical Specifications (TSs) 3/4 3.1, ‘‘Reactor Trip System (RTS) Instrument,’’ and 3/4 3.2, ‘‘Engineered Safety Features Actuation System (ESFAS) Instrument,’’ to increase the surveillance interval from monthly to quarterly for certain RTS and ESFAS instrument channel functional tests. Date of issuance: September 19, 2005. Effective date: September 19, 2005. Amendment Nos.: 267 and 149. Facility Operating License Nos. DPR– 66 and NPF–73: Amendments revised the TSs. Date of initial notice in Federal Register: July 6, 2004 (69 FR 40674). The supplements dated February 23 and August 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 Nuclear Regulatory Commission (NRC) staff’s original proposed no significant hazards consideration determination. VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 19, 2005. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50–334 and 50–412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS–1 and 2), Beaver County, Pennsylvania Date of application for amendments: May 26, 2004, as supplemented by letters dated October 29 and December 3, 2004, and January 18, June 15, and August 15, 2005. Brief description of amendments: The amendments extended the allowable outage time for the BVPS–1 and 2 emergency diesel generators (EDGs) from 72 hours to 14 days. The amendments also deleted surveillance requirement (SR) 4.8.1.1.2.b.1 concerning periodic EDG inspections. Requirements for periodic EDG inspections will be specified in a licensee-controlled EDG maintenance program referenced in the Updated Final Safety Analysis Report. The amendments also revised footnote (1) of TS 3.8.1.1 to clarify the wording to allow actions to be delayed for up to 7 days to allow time to restore fuel oil back to its specified limits when an EDG is inoperable solely due to failure to meet fuel oil property limits of SR 4.8.1.1.2.d.2 or SR 4.8.1.1.2.e. Date of issuance: September 29, 2005. Effective date: Upon issuance to be implemented within 60 days. The implementation shall include the commitments as described in the licensee’s submittals dated May 26 and December 3, 2004, and January 18 and June 15, 2005, and as described in the NRC staff’s safety evaluation related to this amendment. Amendment Nos.: 268 and 150. Facility Operating License Nos. DPR– 66 and NPF–73: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: July 6, 2004 (69 FR 40673). The supplements dated October 29 and December 3, 2004, and January 18, June 15, and August 15, 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 September 29, 2005. No significant hazards consideration comments received: No. PO 00000 Frm 00065 Fmt 4703 Sfmt 4703 Florida Power and Light Company, Docket No. 50–335, St. Lucie Plant, Unit No. 1, St. Lucie County, Florida Date of application for amendment: December 20, 2004. Brief description of amendment: This amendment revises Technical Specifications Figures 3.1–1b, 3.4–2a, 3.4–2b and 3.4–3 to reflect an extension in the effectiveness of the pressure/ temperature (P/T) limit curves from 23.6 to 35 effective full power years (EFPY). The low temperature overpressure protection requirements, which are based on the P/T limits, are also extended to 35 EFPY. Date of Issuance: September 21, 2005. Effective Date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 196. Renewed Facility Operating License No. DPR–67: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: March 1, 2005 (70 FR 9993). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 2005. No significant hazards consideration comments received: No. Florida Power and Light Company, et al., Docket Nos. 50–335 and 50–389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of application for amendments: September 18, 2003, as supplemented on August 25 and September 15, 2005. Brief description of amendments: The amendments revise Technical Specifications (TSs) for the control room ventilation systems to model the Combustion Engineering Standard Technical Specifications, NUREG–1432. In addition, Table 3.3–6, Radiation Monitoring Instrumentation, in each unit’s TSs is revised to resolve minor inconsistencies that resulted from changes associated with previously issued Amendments 184 (Unit 1) and 127 (Unit 2). The amendments also correct some minor typographical errors. Date of Issuance: September 27, 2005. Effective Date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: 197 and 139. Renewed Facility Operating License Nos. DPR–67 and NPF–16: Amendments revised the TSs. Date of initial notice in Federal Register: October 28, 2003 (68 FR 61478). The August 25 and September 15, 2005, supplements did not affect the original proposed no significant hazards E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices determination, or expand the scope of the request as noticed in the Federal Register. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 2005. No significant hazards consideration comments received: No. Florida Power Corporation, et al., Docket No. 50–302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida Date of application for amendment: January 13, 2005, as supplemented by letters dated February 11, May 6, and June 9, 2005. Brief description of amendment: The amendment allows a one-time extended allowed outage time (AOT) change to Improved Technical Specifications (ITS) 3.5.2, Emergency Core Cooling Systems (ECCS)—Operating; 3.6.6, Reactor Building Spray and Containment Cooling Systems; 3.7.8, Decay Heat Closed Cycle Cooling Water System (DC); and 3.7.10, Decay Heat Seawater System to allow the refurbishment of Decay Heat Seawater System Pump RWP–3B online. Date of issuance: September 15, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 221. Facility Operating License No. DPR–72: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: February 1, 2005 (70 FR 5246). The February 11, May 6, and June 9, 2005, supplements contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2005. No significant hazards consideration comments received: No. Nuclear Management Company, LLC, Docket Nos. 50–266 and 50–1, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of application for amendments: July 24, 2005. Brief description of amendments: The amendments incorporated a Point Beach Nuclear Plant (PBNP), Unit 1 reactor vessel head (RVH) drop accident analysis into the PBNP Final Safety Analysis Report and revised the PBNP, Unit 2 RVH drop accident analysis. Date of issuance: September 23, 2005. VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 220, 226. Facility Operating License Nos. DPR– 24 and DPR–27: Amendments revised the License. Date of initial notice in Federal Register: August 16, 2005 (70 FR 48198). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 23, 2005. No significant hazards consideration comments received: No. Nuclear Management Company, LLC, Docket Nos. 50–266 and 50–301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin Date of amendment request: April 8, 2004, as supplemented by letters dated November 15, 2004, July 15 and August 8, 2005. Description of amendment request: The amendments revised technical specification surveillance requirements (SR) 3.8.4.6 and SR 3.8.4.7, ‘‘DC Sources—Operating.’’ Specifically, the amendments revised battery charger current values, added a new allowance for verifying battery charger capacity, and removed a restriction on the conduct of a modified performance discharge test. Date of issuance: September 27, 2005. Effective date: As of the date of issuance and shall be implemented within 45 days. Amendment Nos.: 221, 227. Facility Operating License Nos. DPR– 24 and DPR–27: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: August 19, 2004 (69 FR 51489). The November 15, 2004, July 15 and August 8, 2005, supplemental letters provided additional information that clarified the application, did not expand the scope of the application originally noticed, and did not change the NRC staff’s original proposed no significant hazards consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 27, 2005. No significant hazards consideration comments received: No. R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50–244, R.E. Ginna Nuclear Power Plant, Wayne County, New York Date of application for amendment: September 30, 2004, and May 28, 2005. PO 00000 Frm 00066 Fmt 4703 Sfmt 4703 59093 Brief description of amendment: The amendment revises information in the Updated Final Safety Analysis Report (UFSAR) regarding the application of ‘‘leak-before-break’’ methodology for the emergency core cooling system accumulator lines A and B and the pressurizer surge line. The amendment permits the exclusion of these lines from the evaluation of the dynamic effects associated with postulated highenergy line breaks in the analyzed segments of the accumulator lines piping system and the pressurizer surge line piping system. Date of issuance: September 22, 2005. Effective date: As of the date of issuance and shall be implemented with the next update of the UFSAR in accordance with 10 CFR 50.71(e). Amendment No.: 92. Renewed Facility Operating License No. DPR–18: Amendment revised the UFSAR. Date of initial notice in Federal Register: July 5, 2005 (70 FR 38721). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 22, 2005. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of application for amendments: August 12, 2005, as supplemented by letter dated August 24, 2005. Brief description of amendments: The amendments revised the Technical Specifications to incorporate changes in the steam generator (SG) inspection scope for Vogtle, Unit 2 during Refueling Outage 11 and 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 Vogtle, Unit 1 is affected only due to the fact that Unit 1 and Unit 2 use common Technical Specifications. Date of issuance: September 21, 2005. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: 138/117. Facility Operating License Nos. NPF–68 and NPF–81: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: August 22, 2005 (70 FR 48985). The supplement dated August 24, 2005, provided clarifying information that did not change the scope of the August 12, 2005, application nor the E:\FR\FM\11OCN1.SGM 11OCN1 59094 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices initial proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 21, 2005. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of application for amendments: August 13, 2004, as supplemented by letters dated May 3 and July 7, 2005. Brief description of amendments: The amendments revised the Technical Specifications (TSs) to reflect updated spent fuel rack criticality analyses for Units 1 and 2. The amendments also corrected a typographical error on Page vi of the TSs Table of Contents associated with the issuance of Amendments 130 and 109, for Units 1 and 2 TSs, respectively. Date of issuance: September 22, 2005. Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment Nos.: 139/118. Facility Operating License Nos. NPF–68 and NPF–81: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: November 9, 2004 (69 FR 64990). The supplements dated May 3 and July 7, 2005, provided clarifying information that did not change the scope of the August 13, 2004, application nor the initial proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 22, 2005. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50–390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee Date of application for amendment: November 21, 2003, as supplemented by letters dated May 5 and August 19, 2004, and July 11, 2005. Brief description of amendment: The amendment allows the position of the control and shutdown rods to be monitored by a means other than the movable incore detectors. Date of issuance: September 20, 2005. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment No.: 58. VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 Facility Operating License No. NPF– 90: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: December 23, 2003 (68 FR 74267). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 20, 2005. No significant hazards consideration comments received: No. TXU Generation Company LP, Docket Nos. 50–445 and 50–446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: March 24, 2005. Brief description of amendments: The amendments revise Technical Specification (TS) 3.3.1 entitled ‘‘Reactor Trip System (RTS) Instrumentation’’ and TS 3.3.2 entitled ‘‘Engineered Safety Feature Actuation System (ESFAS) Instrumentation’’, and Required Action Notes in the TSs to reflect wording in the Commissions Standard TSs incorporating the channel bypass capabilities as discussed in TS Task Force Traveler 418, Revision 2. Date of issuance: September 29, 2005. Effective date: Effective as of the date of issuance and shall be implemented in 90 days from the date of issuance. Amendment Nos.: 121 and 121. Facility Operating License Nos. NPF– 87 and NPF–89: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: April 26, 2005 (70 FR 21464). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 29, 2005. No significant hazards consideration comments received: No. Virginia Electric and Power Company, Docket Nos. 50–338 and 50–339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia Date of application for amendment: September 15, 2004, as supplemented by letter dated May 5, 2005. Brief description of amendment: These amendments revise the Technical Specifications for North Anna Power Station, Units 1 and 2 to support the implementation of the proposed Topical Report DOM–QA–1, ‘‘Dominion Nuclear Facility Quality Assurance Program Description.’’ The implementation of this topical report would create a PO 00000 Frm 00067 Fmt 4703 Sfmt 4703 common quality assurance program for North Anna, Surry, and Millstone Power Stations. The review of these proposed amendments was requested to be done in concert with the review of the Topical Report. The Topical Report was submitted to the NRC staff for review on August 24, 2004, and supplemented by letter dated May 5, 2005. By letter dated September 9, 2005, the NRC staff approved of Topical Report DOM–QA– 1. Date of issuance: September 15, 2005. Effective date: As of the date of issuance and shall be implemented within 6 months from the date of issuance. Amendment Nos.: 243 and 224. Renewed Facility Operating License Nos. NPF–4 and NPF–7: Amendments change the Technical Specifications. Date of initial notice in Federal Register: November 23, 2004 (69 FR 68187). The supplement dated May 5, 2005, contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 15, 2005. No significant hazards consideration comments received: No. Virginia Electric and Power Company, et al., Docket Nos. 50–280 and 50–281, Surry Power Station, Units 1 and 2, Surry County, Virginia Date of application for amendments: September 15, 2004, as supplemented by letter dated May 5, 2005. Brief description of amendments: These amendments revise the Technical Specifications for Surry Power Station, Units 1 and 2 to support the implementation of the proposed Topical Report DOM–QA–1, ‘‘Dominion Nuclear Facility Quality Assurance Program Description.’’ The implementation of this topical report would create a common quality assurance program for North Anna, Surry, and Millstone Power Stations. The review of these proposed amendments was requested to be done in concert with the review of the Topical Report. The Topical Report was submitted to the NRC staff for review on August 24, 2004, and supplemented by letter dated May 5, 2005. Subsequently, the NRC staff approved this Topical Report on September 9, 2005. Date of issuance: As of the date of issuance and shall be implemented within 6 months from the date of issuance. Effective date: September 15, 2005. Amendment Nos.: 244/243. E:\FR\FM\11OCN1.SGM 11OCN1 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices Renewed Facility Operating License Nos. DPR–32 and DPR–37: Amendments change the Technical Specifications. Date of initial notice in Federal Register: December 7, 2004 (69 FR 70723). The supplement dated May 5, 2005, contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 15, 2005. No significant hazards consideration comments received: No. Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances) During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing. For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee’s facility of the licensee’s application and of the Commission’s proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 appropriate and the licensee has been informed of the public comments. In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant’s licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible. Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved. The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records PO 00000 Frm 00068 Fmt 4703 Sfmt 4703 59095 will be accessible from the Agencywide Documents Access and Management System’s (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission’s ‘‘Rules of Practice for Domestic Licensing Proceedings’’ in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737, or by email to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor’s/petitioner’s right under the Act to be made a party E:\FR\FM\11OCN1.SGM 11OCN1 59096 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 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 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 1 To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant’s counsel and discuss the need for a protective order. VerDate Aug<31>2005 16:40 Oct 07, 2005 Jkt 208001 seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/ requestors with respect to that contention. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect. A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415–1101, verification number is (301) 415–1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee. Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer or the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(I)–(viii). PO 00000 Frm 00069 Fmt 4703 Sfmt 4703 Indiana Michigan Power Company, Docket Nos. 50–315 and 50–316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment request: September 12, 2005. Description of amendment request: The amendments replace the paragraph of Improved Technical Specification (ITS) Surveillance Requirement (SR) 3.8.1.18 with the wording of previous TS SR 4.8.1.1.2.e.11. Date of issuance: September 23, 2005. Effective date: Immediately. Amendment Nos.: 290, 272. Facility Operating License Nos. (DPR– 58 and DPR–74): Amendment revises the technical specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. HeraldPalladium on September 18, 2005. The notice provided an opportunity to submit comments on the Commission’s proposed NSHC determination. No comments have been received. The Commission’s related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated September 23, 2005. Attorney for licensee: James M. Petro, Jr., Esquire, One Cook Place, Bridgman, MI 49106. NRC Section Chief: L. Raghavan. Dated at Rockville, Maryland, this 3rd day of October 2005. For the Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation. [FR Doc. 05–20168 Filed 10–7–05; 8:45 am] BILLING CODE 7590–01–P SECURITIES AND EXCHANGE COMMISSION [File No. 1–31514] Issuer Delisting; Notice of Application of Meredith Enterprises, Inc. to Withdraw Its Common Stock, $.01 Par Value, From Listing and Registration on the American Stock Exchange LLC October 4, 2005. On September 15, 2005, Meredith Enterprises, Inc., a Delaware corporation (‘‘Issuer’’), filed an application with the Securities and Exchange Commission (‘‘Commission’’), pursuant to Section 12(d) of the Securities Exchange Act of 1934 (‘‘Act’’) 1 and Rule 12d2–2(d) 1 15 E:\FR\FM\11OCN1.SGM U.S.C. 78l(d). 11OCN1
[Pages 59082-59096]
[FR Doc No: 05-20168]
proposed to be issued from September 6, 2005, to September 29, 2005.
The last biweekly notice was published on September 27, 2005 (70 FR
56499).
[[Page 59083]]
[[Page 59084]]
Date of amendment request: April 6, 2005, as supplemented by letter
dated August 8, 2005.
modify Technical Specification (TS) 6.8.4.k, ``Containment Leakage Rate
Testing Program,'' and TS Surveillance Requirement (SR) 4.6.1.6.1,
``Containment Vessel Surfaces.'' The proposed amendment would modify
the TS to allow for a one-time extension of the containment Type A test
interval from once in 10 years to once in 15 years.
This change does not involve a significant hazards consideration
1. The proposed amendment does not involve a significant
The proposed change to HNP [Harris Nuclear Plant] TS 6.8.4.k and
TS SR 4.6.1.6.1 provide a one-time extension of the containment Type
A test interval from 10 years to 15 years and specifies that
additional visual inspections are done in accordance with
Subsections IWE and IWL of the ASME [American Society of Mechanical
Engineers] Section XI Code. The existing 10-year test interval is
based on past test performance. The proposed TS change does not
which the plant is operated or controlled. The containment vessel is
designed to provide a leak-tight barrier against the uncontrolled
release of radioactivity to the environment in the unlikely event of
postulated accidents. As such, the containment vessel is not
considered as the initiator of an accident. Therefore, the proposed
TS change does not involve a significant increase in the probability
The proposed change involves only a one-time change to the
interval between containment Type A tests. Type B and C leakage
testing will continue to be performed at the intervals specified in
10 CFR Part 50, Appendix J, Option A, as required by the HNP TS. As
documented in NUREG-1493, ``Performance-Based Containment Leakage-
Test Program,'' industry experience has shown that Type B and C
containment leak rate tests have identified a very large percentage
of containment leak paths, and that the percentage of containment
leak paths that are detected only by Type A testing is very small.
In fact, an analysis of 144 integrated leak rate tests, including 23
failures, found that none of the failures involved a containment
liner breach. NUREG-1493 also concluded, in part, that reducing the
frequency of containment Type A testing to once per 20 years results
in an imperceptible increase in risk. The HNP test history and risk-
based evaluation of the proposed extension to the Type A test
interval supports this conclusion. The design and construction
requirements of the containment vessel, combined with the
containment inspections performed in accordance with the American
Society of Mechanical Engineers (ASME) Code, Section XI, and the
Maintenance Rule (10 CFR 50.65) provide a high degree of assurance
that the containment vessel will not degrade in a manner that is
detectable only by Type A testing. Therefore, the proposed TS change
2. The proposed amendment does not create the possibility of a
The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1
provide a one-time extension of the containment Type A test interval
to 15 years and specifies that additional visual inspections are
done in accordance with Subsections IWE and IWL of the ASME Section
XI Code. The existing 10-year test interval is based on past test
performance. The proposed change to the Type A test interval does
not result in any physical changes to HNP. In addition, the proposed
test interval extension does not change the operation of HNP such
that a failure mode involving the possibility of a new or different
kind of accident from any accident previously evaluated is created.
3. The proposed amendment does not involve a significant
from 10 years to 15 years and specifies that additional visual
inspections are done in accordance with Subsections IWE and IWL of
the ASME Section XI Code. The existing 10-year test interval is
based on past test performance. The NUREG-1493 study of the effects
of extending containment leak rate testing found that a 20 year
extension for Type A testing resulted in an imperceptible increase
in risk to the public. NUREG-1493 found that, generically, the
design containment leak rate contributes a very small amount to the
would have a minimal affect on this risk since most potential leak
paths are detected by Type B and C testing. The proposed change
involves only a one-time extension of the interval for containment
Type A testing; the overall containment leak rate specified by the
HNP TS is being maintained. Type B and C testing will continue to be
performed at the frequency required by the HNP TS. The regular
containment inspections being performed in accordance with the ASME
Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a
manner that is only detectable by Type A testing. In addition, a
plant-specific risk evaluation has demonstrated that the one-time
extension of the Type A test interval from 10 years to 15 years
results in a very small increase in risk for those accident
sequences influenced by Type A testing.
Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham Counties,
Technical Specifications (TS) 3/4.4.7, ``Reactor Coolant System
Chemistry.'' Specifically, the proposed amendment would revise the
footnotes in Tables 3.4-2 and 4.4-3 of the TS to increase the
temperature limit from 180 [deg]F to 250 [deg]F above which reactor
coolant sampling and analysis for dissolved oxygen is required and
dissolved oxygen limits apply.
This amendment does not involve a significant hazards consideration
Operation of HNP in accordance with the proposed amendment does
[[Page 59085]]
Final Safety Analysis Report (FSAR) documents the analyses of design
basis accidents (DBA) at HNP. Any scenario or previously analyzed
accident that results in offsite dose were evaluated as part of this
analysis. The proposed amendment does not change or affect any
accident previously evaluated in the FSAR. The proposed amendment
does not modify any plant equipment. In addition, the proposed
amendment does not result in a change to a structure, system, or
component (SSC), or adversely affect its design function.
The purpose of the temperature limit for RCS [Reactor Coolant
System] oxygen control is to minimize corrosion at high temperatures
on RCS components. Increasing the temperature at which oxygen levels
are required to be maintained within specified limits from 180
[deg]F to 250 [deg]F is supported by industry and vendor data which
indicates that the influence of dissolved oxygen at or below 250
[deg]F is not significant with regard to stress corrosion cracking
and general corrosion of RCS components. The proposed amendment is
consistent with the Electric Power Research Institute's (EPRI's)
guidelines for Pressurized Water Reactor (PWR) Primary Water
Chemistry. This amendment places HNP in line with standard industry
specifications for reactors of similar size and vintage. HNP's
proposed amendment to increase the temperature limit for
applicability to 250 [deg]F would decrease the time needed to
achieve compliance with the dissolved oxygen limit and decrease the
overall time to restart the plant from cold shutdown. Removing
oxygen in a more expeditious fashion enhances RCS chemistry. Based
on the above, RCS integrity is maintained by this amendment.
Therefore, this amendment does not involve a significant
from any accident previously evaluated. The FSAR documents the
analyses of design basis accidents (DBA) at HNP. Any scenario or
previously analyzed accident that results in offsite dose were
evaluated as part of this analysis. The proposed amendment does not
change or affect any accident previously evaluated in the FSAR, and
no new or different scenarios are created by the proposed amendment
to the TS. The proposed amendment does not modify any plant
equipment. In addition, the proposed amendment does not result in a
change to an SSC [structure, system, or component] or adversely
affect its design function.
The purpose of the temperature limit for RCS oxygen control is
to minimize corrosion at high temperatures on RCS components.
Increasing the temperature at which oxygen levels are required to be
maintained within specified limits from 180 [deg]F to 250 [deg]F is
supported by industry and vendor data which indicates that the
influence of dissolved oxygen at or below 250 [deg]F is not
significant with regard to stress corrosion cracking and general
corrosion of RCS components. The proposed amendment is consistent
with EPRI's guidelines for PWR Primary Water Chemistry. This
amendment places HNP in line with standard industry specifications
for reactors of similar size and vintage. HNP's proposed amendment
to increase the temperature limit for applicability to 250 [deg]F
would decrease the time needed to achieve compliance with the
dissolved oxygen limit and decrease the overall time to restart the
plant from cold shutdown. Removing oxygen in a more expeditious
fashion enhances RCS chemistry. Based on the above, RCS integrity is
maintained by this amendment.
Therefore, this amendment does not create the possibility of a
not involve a significant reduction in a margin of safety. Existing
TS operability and surveillance requirements are not reduced by the
proposed amendment. The proposed amendment does not modify any plant
change to a structure, system, or component (SSC), or its design
function. The proposed amendment does not adversely affect existing
plant safety margins or the reliability of equipment assumed to
mitigate accidents in the FSAR.
revise Cooper Nuclear Station (CNS) Technical Specification (TS) 5.3,
``Unit Staff Qualifications,'' to upgrade the qualification standard
for the Shift Manager, Senior Operator, Licensed Operator, and Shift
Technical Engineer from Regulatory Guide (RG) 1.8, Revision 2
``Qualification and Training of Personnel for Nuclear Power Plants,''
to RG 1.8, Revision 3. It also clarifies qualification requirements
applicable to the Operations Manager position.
These changes are administrative in nature and do not require
any physical modifications, affect any plant components, or result
in any changes in plant operation. They provide clarity and
consistency to the CNS licensing basis.
Upgrading the unit staff qualifications for the Shift Manager,
Senior Operator, Licensed Operator, and Shift Technical Engineer
from Regulatory Guide 1.8, Revision 2, to Regulatory Guide 1.8,
Revision 3, is an administrative change that will clarify the
current requirements for qualification and training of operations
personnel. The changes are consistent with the application of a
systems approach to training in an accredited training program. By
promulgation of the 10 CFR Part 55 rule change, the NRC determined
that an accredited licensed operator training program based on a
systems approach to training provides an acceptable means of
qualifying licensed operating personnel.
The addition of qualification requirements for the Operations
Manager position clarifies SRO [Senior Reactor Operator] license
requirements for Operations management personnel by specifying that
the Operations Supervisor is the member of Operations management
required to have a current SRO license at CNS. The Operations
Manager is required to hold or have previously held a
[[Page 59086]]
SRO license. This will ensure an acceptable level of operations
knowledge to perform in a managerial oversight role. This approach
is consistent with current guidance in ANSI/ANS [American Nuclear
Standards Institute/American Nuclear Society] 3.1-1993. This change
is administrative in nature and has no impact on previously
These changes are administrative in nature and do not involve a
physical alteration of the plant or a change to plant operations. No
new failure mechanisms, malfunctions, or accident initiators are
introduced. The proposed changes provide clarity and consistency to
the CNS licensing basis in regard to training and qualification of
operations personnel and SRO license requirements for Operations
These changes are administrative in nature and do not affect any
Technical Specification safety limit or limiting condition for
operation. No safety margins are affected by these changes. The
proposed changes do not involve a change in plant design or
operation for the mitigation of postulated accidents. The proposed
changes provide clarity and consistency to the CNS licensing basis
in regard to training and qualification of operations personnel and
SRO license requirements for Operations management personnel.
revise the definitions of Channel Calibration, Channel Function Test,
and Logic System Functional Test in accordance with the Technical
Specification Task Force Traveler 205-A.
The definitions of Channel Calibration, Channel Functional Test,
and Logic System Functional Test specified in Technical
Specifications (TS) provide basic information regarding what the
test involves, the components involved in the test, and general
information regarding how the test is to be performed. These
definitions and their specific wording are not precursors to any
accident. As a result these revised definitions result in no
The proposed revisions of these definitions involve no changes
to plant design, equipment, or operation related to mitigation of
accidents. The proposed revisions of these definitions do not change
their meaning or intent. The proposed revisions clarify the
definitions and do not result in a reduction of required testing of
instrumentation used to mitigate accidents.
Based on the above NPPD [Nebraska Public Power District]
concludes that the proposed changes do not involve a significant
The proposed revisions of the definitions do not involve a
change to the design or operation of any plant structure, system, or
component (SSC). As a result the plant will continue to be operated
in the same manner. The proposed revisions will not result in a
change to how the instrumentation used to monitor plant operation
and to mitigate accidents is tested. Operating the plant and testing
the plant's instrumentation in the same manner as is currently done
Based on the above NPPD concludes that the proposed changes do
The affected definitions involve testing of instrumentation used
in the mitigation of accidents to ensure that the instrumentation
will perform as assumed in safety analyses. The proposed revisions
of these definitions will not change their meaning or intent. As a
result, the instrumentation will continue to be tested in the same
manner as is currently done. Revising these definitions as proposed
will not result in a change to the design or operation of any plant
SSC used to shutdown the plant, initiate the Emergency Core Cooling
Systems, or isolate primary or secondary containment. As a result
the ability of the plant to respond to and mitigate accidents is
unchanged by the revised definitions.
Based on the above, NPPD concludes that the proposed changes do
revise Technical Specification 3.7.5, ``Auxiliary Feedwater (AFW)
System,'' to change the frequency of Surveillance Requirement 3.7.5.6
from 92 days to 24 months.
The proposed change to increase [the] frequency interval for
Surveillance Requirement (SR) 3.7.5.6 from 92 days to 24 months has
no impact on the probability of accidents previously evaluated. The
valves controlled by SR 3.7.5.6 are used to provide an alternate
supply of water to the auxiliary feedwater (AFW) system from the
fire water storage tank (FWST) and are only operated after an
accident has occurred. They are not accident initiators.
Misoperation, or failure of a[n] FWST supply to be correctly
positioned following an accident, could result in an inadequate
supply of water to the AFW system. Failure to provide adequate core
cooling could increase the radiological consequences of an accident.
However, operating and maintenance histories of the FWST supply
valves show that these valves have been
[[Page 59087]]
capable of full stroke cycling each time they have been tested.
There is no evidence of any time-related degradation mechanism that
would prevent the valves from performing their design function.
Thus[,] the proposed change has no impact on the consequences of an
different [kind of] accident from any accident previously evaluated?
The proposed change to increase frequency interval for SR
3.7.5.6 from 92 days to 24 months has no impact on the probability
of accidents of the type evaluated in the Final Safety Analysis
Report, as updated. The valves are used to provide an alternate
supply of water to the AFW system from the FWST, and are only
operated after an accident has occurred. They are not accident
initiators. Review of the operating and maintenance histories of the
FWST supply valves show that they are highly reliable in maintaining
their capability to perform their design function.
of a new or different [kind of] accident from any accident
The proposed change to SR 3.7.5.6 involves only an increase in
the frequency interval. No physical changes are required to the
facility or to the plant operating or emergency procedures as a
result of the change. Based on review of the operating and
maintenance histories of the FWST supply valves, they have been
would prevent the valves from performing their design function. This
evidence supports the conclusion that there will be no impact in the
operation of these valves following an accident.
revise the expiration dates of the Units 1 and 2 facility-operating
licenses to recapture low-power testing time, and to reflect a 40-year
term measured from the date of issuance of each unit's full-power
The proposed additional operating license periods do not affect
the probability or consequences of an accident previously evaluated
since they require no physical change in the plant equipment or
operating procedures and the Final Safety Analysis Report (FSAR)
Update safety analyses are based on [a] 40-year full[-]power
operation. Surveillance and maintenance practices, as well as other
programs such as environmental qualification of equipment, ensure
timely identification and correction of any degradation of safety-
related plant equipment. The long-term integrity of the reactor
vessels has been evaluated using currently acceptable NRC
calculational methods and best available Diablo Canyon Power Plant
(DCPP) specific data. The evaluation results demonstrate that both
reactor vessels are safe for normal operations in excess of 40
The possibility of a new or different kind of accident is not
created by the proposed additional operating periods since at least
40 years of full[-]power operation was assumed in the design and
construction of DCPP Units 1 and 2. The plant maintenance programs
are also designed to both maintain and determine the need to replace
safety-related components. These programs will continue to be
applied as they are presently to assure safe operation.
The proposed additional operating periods do not involve a
significant reduction in a margin of safety since, as is the case
with present operation, degradation of safety-related equipment will
be identified and corrected by ongoing surveillance and maintenance
practices. Existing programs, routine maintenance, and compliance
with Technical Specifications assure that an adequate margin of
safety is maintained. These activities will remain in effect for the
duration of the proposed additional operating periods.
the Administrative Control section of the Technical Specifications
(TSs) to permit the Westinghouse best estimate methodology for loss-of-
coolant-accident (LOCA) analysis methodology to be utilized for
analyses as required by Title 10 of the Code of Federal Regulations,
Part 50, Section 46, ``Acceptance criteria for emergency core cooling
systems [ECCS] for light water nuclear power reactors' (10 CFR 50.46).
Implementation of the best-estimate large break LOCA methodology
and associated TS changes is proposed to increase margin to the peak
clad temperature limits defined in 10 CFR 50.46. There are no
physical plant changes or changes in manner in which the plant will
be operated as a result of this
[[Page 59088]]
change. Since the plant conditions and ECCS performance assumed in
the analysis are consistent with the plant's current design, the
proposed change in methodology will thus have no impact on the
probability of a LOCA. When applied, the best estimate methodology
shows that the ECCS is more effective than previously evaluated in
mitigating the consequences of a LOCA, as lower peak clad
temperatures are predicted relative to current 10 CFR 50.46 Appendix
K results. Since the proposed best-estimate methodology is only
applicable to a large break LOCA and since the application of the
proposed methodology shows there is a high probability that all of
the acceptance criteria contained in 10 CFR 50.46, Paragraph b are
met, the proposed change does not increase the consequences of an
There are no physical changes being made to the plant. No new
modes of plant operation are being introduced. The parameters
assumed in the analysis remain within the design limits of the
existing plant equipment. All plant systems will perform as designed
during the response to a potential accident.
of a new or different kind of accident from any previously analyzed.
It has been shown that the methodology used in the analysis
would more realistically describe the expected behavior of V. C.
Summer Nuclear Station systems during a postulated loss of coolant
accident. Uncertainties have been accounted for as required by 10
CFR 50.46. A sufficient number of loss of coolant accidents with
different break sizes, different locations and other variations in
properties are analyzed to provide assurance that the most severe
postulated loss of coolant accidents are calculated. It has been
shown by analysis that there is a high level of probability that all
criteria contained in 10 CFR 50.46, Paragraph b are met.
Pursuant to 10 CFR 50.91, the preceding analyses provide a
determination that the proposed Technical Specifications change
poses no significant hazard as delineated by 10 CFR 50.92.
this review, it appears that the three standards of 10 CFR 50.92 (c)
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
change the Technical Specifications (TSs) to reflect the use of the
Westinghouse Best Estimate Analyzer for Core Operations--Nuclear
(BEACON) to augment the functional capability of the flux mapping
system for the purpose of power distribution surveillances. In
addition, editorial changes to the TSs are proposed.
The PDMS [power distribution monitoring system] performs
continuous core power distribution monitoring. This system utilizes
the NRC-approved Westinghouse proprietary computer code BEACON to
provide data reduction for incore flux maps, core parameter
analysis, load follow operation simulation, and core prediction. It
in no way provides any protection or control system function.
Fission product barriers are not impacted by these proposed changes.
The proposed changes occurring with PDMS will not result in any
additional challenges to plant equipment that could increase the
probability of any previously evaluated accident. The changes
associated with the PDMS do not affect plant systems such that their
function in the control of radiological consequences is adversely
affected. These proposed changes will therefore not affect the
mitigation of the radiological consequences of any accident
described in the Updated Final Safety Analysis Report Update
(UFSAR).
Continuous on-line monitoring through the use of PDMS provides
significantly more information about the power distributions present
in the core than is currently available. This results in more time
(i.e., earlier determination of an adverse condition developing) for
operator action prior to having an adverse condition develop that
could lead to an accident condition or to unfavorable initial
conditions for an accident.
Each accident analysis addressed in the UFSAR is examined with
respect to changes in cycle-dependent parameters, which are obtained
from application of the NRC-approved reload design methodologies, to
ensure that the transient evaluations of reload cores are bounded by
previously accepted analyses. This examination, which is performed
in accordance with the requirements set forth in 10 CFR [Title 10 of
the Code of Federal Regulations] 50.59, ensures that future reloads
The three editorial changes only correct typographical errors
made in previously approved TS changes. They do not affect plant
operation or structures, systems, and components important to
The implementation of the PDMS has no influence or impact on
plant operations or safety, nor does it contribute in any way to the
probability or consequences of an accident. No safety-related
equipment, safety function, or plant operation will be altered as a
result of this proposed change. The possibility for a new or
different type of accident from any accident previously evaluated is
not created since the changes associated with implementation of the
PDMS do not result in a change to the design basis of any plant
component or system. The evaluation of the effects of using the PDMS
to monitor core power distribution parameters shows that all design
standards and applicable safety criteria limits are met.
will not result in more adverse conditions and will not result in
any increase in the challenges to safety systems. The cycle-specific
variables required by the PDMS are calculated using NRC-approved
methods. The TS will continue to require operation within the
required core operating limits and appropriate actions will be taken
if limits are exceeded.
The margin of safety is not affected by implementation of the
PDMS. The margin of safety provided by current TS is unchanged. The
proposed changes continue to require operation within the core
limits that are based on NRC-approved reload design methodologies.
Appropriate measures exist to control the values of these cycle-
specific limits. The proposed changes continue to ensure that
appropriate actions will be taken
[[Page 59089]]
if limits are violated. These actions remain unchanged.
Unit 1, DeWitt County, Illinois
Date of application for amendment: April 3, 2003, as supplemented
December 23, 2003, December 9 and 17, 2004, and March 30 and August 19,
Specifications (TSs) to support the application of an alternative
source term methodology in accordance with Title 10 of the Code of
Federal Regulations, Section 50.67, ``Accident Source Term,'' with the
exception that Technical Information Document 14844, ``Calculation of
Distance Factors for Power and Test Reactor Sites,'' was used as the
radiation dose basis for equipment qualification.
Date of initial notice in Federal Register: September 2, 2003 (68
FR 52234).
in a Safety Evaluation dated September 19, 2005.
The supplements dated December 23, 2003, December 9 and 17, 2004,
and March 30 and August 19, 2005 provided additional information that
Date of application for amendment: November 11, 2003, as
supplemented April 16 and September 10, 2004, and March 30 and
instrument channel trip setpoint allowable values for thirteen
Technical Specification (TS) functions at Clinton Power Station, Unit
12363).
in a Safety Evaluation dated September 21, 2005. The supplements dated
April 16 and September 10, 2004, and March 30 and September 21, 2005,
consideration determination. No significant hazards consideration
Date of application for amendments: August 3, 2004, as supplemented
on July 8 and August 26, 2005.
Brief description of amendments: The amendments extend the
surveillance frequency interval from monthly to quarterly for Technical
Specification surveillance requirement (SR) 3.3.3.1, which involves a
channel functional test of each reactor trip circuit breaker (RTCB).
SRs 3.3.3.1 and 3.3.3.2 will be scheduled such that the RTCBs testing
is performed every 6 weeks, which meets the vendor-recommended interval
for cycling each RTCB.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
[[Page 59090]]
The July 8 and August 26, 2005, supplemental letters provided
contained in a Safety Evaluation dated September 26, 2005.
Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power
Station Unit No. 1, New London County, Connecticut
Date of application for amendment: September 8, 2004, as
supplemented by letters dated May 5 and July 27, 2005.
Brief description of amendment: The amendment revised the Millstone
Power Station, Unit No. 1 Technical Specifications (TSs) to support the
implementation of the proposed Dominion Nuclear Facility Quality
Assurance Program (Topical Report DOM-QA-1). Implementation of this
Topical Report would create a common quality assurance program for all
sites owned by Dominion Nuclear Connecticut, Inc. Review of this
proposed amendment was requested in concert with the review of the
implemented by February 28, 2006.
Facility Operating License No. DPR-21: The amendment revised the
2888).
The additional information provided in the supplemental letters
dated May 5 and July 27, 2005, did not expand the scope of the
application as noticed and did not change the NRC staff's original
proposed no significant hazards consideration determination.
in a Safety Evaluation dated September 15, 2005.
Date of application for amendment: July 15, 2004, as supplemented
by letter dated August 23, 2004.
Brief description of amendment: The amendment revised the Facility
Operating License DPR-65 to address the resolution of a non-
conservative Technical Specifications (TSs) associated with control
room isolation radiation monitoring instrumentation. Specifically, the
amendment would revise the TSs to require two operable channels of
control room isolation radiation monitoring instrumentation.
Facility Operating License No. DPR-65: The amendment revised the
2887).
in a Safety Evaluation dated September 23, 2005.
Date of application for amendment: April 15, 2004, as supplemented
Brief description of amendment: The amendment approves
modifications to the Fire Protection Program. Specifically, the
modifications involve converting the existing automatic carbon dioxide
fire suppression systems installed in the cable spreading room to
Facility Operating License No. NPF-49: The amendment allows for
conversion from an automatic to a manual carbon dioxide suppression
system in the cable spreading area.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40672). The supplement dated June 23, 2005, provided clarifying
information and did not change the initial proposed no significant
hazards consideration determination.
in a Safety Evaluation dated September 22, 2005.
Date of application for amendments: September 8, 2004, as
Millstone Power Station, Unit N