Source: https://regulations.justia.com/regulations/fedreg/2005/08/02/E5-4067.html
Timestamp: 2020-01-20 21:00:08
Document Index: 489622785

Matched Legal Cases: ['art 2', 'art 50', 'art 50', 'art 50', 'art] 50', 'art 50', 'arts 30', 'arts 30', 'art 50', 'art 50', 'art 50', 'art] 50', 'art 50']

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 44400-44407 [E5-4067] :: 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, 44400-44407 [E5-4067]
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 44400-44407 [E5-4067]
Download as PDF 44400 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices Dated: July 28, 2005. Sandy Joosten, Office of the Secretary. [FR Doc. 05–15283 Filed 7–29–05; 11:08 am] BILLING CODE 7590–01–M NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC 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 July 8, 2005, to July 21, 2005. The last biweekly notice was published on July 19, 2005 (70 FR 41442). 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 VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, 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 PO 00000 Frm 00078 Fmt 4703 Sfmt 4703 for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission’s ‘‘Rules of Practice for Domestic Licensing Proceedings’’ in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System’s (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor’s/petitioner’s right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor’s/petitioner’s property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor’s/petitioner’s interest. The petition must also set forth the specific contentions which the petitioner/ requestor seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of E:\FR\FM\02AUN1.SGM 02AUN1 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/ requestor to relief. A petitioner/ requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 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 VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 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. Duke Energy Corporation, et al., Docket Nos. 50–413 and 50–414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina and Docket Nos. 50–369 and 50–370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: July 7, 2005. Description of amendment request: The amendments would revise Technical Specification 3.9.1, ‘‘Boron Concentration,’’ to clarify the technical requirements for boron concentration when the refueling canal and the refueling cavity are not connected to the reactor coolant system. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Would implementation of the changes proposed in this LAR [License Amendment Request] involve a significant increase in the probability or consequences of an accident previously evaluated? PO 00000 Frm 00079 Fmt 4703 Sfmt 4703 44401 No. This LAR clarifies Technical Specification [TS] 3.9.1 regarding the applicability of boron concentration limits when the refueling canal and refueling cavity are not connected to the reactor coolant system [RCS]. When the refueling canal and the refueling cavity are isolated from the RCS, no potential path for boron dilution of the RCS exists, thus there is no significant increase in the probability of an accident that has been previously evaluated, nor would there be a significant increase in the consequences of an accident that has been previously evaluated. 2. Would implementation of the changes proposed in this LAR create the possibility of a new or different kind of accident from any accident previously evaluated? No. The change proposed in this LAR clarifies the applicability of TS 3.9.1 when the refueling canal and refueling cavity are not connected to the reactor coolant system. When the refueling canal and the refueling cavity are isolated from the RCS, no potential path for boron dilution of the RCS exists, thus there is no means to initiate an accident that is new or different from any accident that has been previously evaluated. 3. Would implementation of the changes proposed in this LAR involve a significant reduction in a margin of safety? No. The change proposed in this LAR only clarifies the applicability of TS 3.9.1 when the refueling canal and the refueling cavity are not connected to the reactor coolant system. [TS 3.9.1 limits the boron concentrations of the reactor coolant system], the refueling canal, and the refueling cavity to ensure that the reactor remains subcritical during Mode 6 plant conditions. However, when the refueling canal and the refueling cavity are isolated from the reactor coolant system, no potential for boron dilution of the RCS exists. Therefore, in this condition it is not necessary to place a limit on the boron concentration in the refueling canal and the refueling cavity, thus there is no significant reduction in a margin of safety since no specific boron limits are being changed. 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: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201–1006. NRC Section Chief: Evangelos C. Marinos. Entergy Operations, Inc., Docket No. 50– 368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas Date of amendment request: January 31, 2005. Description of amendment request: Entergy Operations, Inc. (EOI) has requested a change which would revise E:\FR\FM\02AUN1.SGM 02AUN1 44402 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices the requirements associated with the Arkansas Nuclear One, Unit 2 (ANO–2) containment overcurrent protection devices. EOI proposes to amend Operating License NPF–6 to eliminate Technical Specifications (TSs) section 3.8.2.5, ELECTRICAL POWER SYSTEMS-Containment Penetration Conductor Overcurrent Protection Devices. The proposed change would relocate the requirements for containment penetration conductor overcurrent protective devices to the Technical Requirements Manual (TRM). 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? The proposed changes to relocate the requirements for containment penetration conductor overcurrent protective devices from Technical Specifications to the TRM will have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design basis accidents will not change. Operation of the containment penetration conductor overcurrent protective devices is not an accident initiator and can not cause an accident. Whether the requirements for the containment penetration conductor overcurrent protective devices are located in Technical Specifications or the TRM will have no effect on the probability or consequences of any accident previously evaluated. Therefore, the removal of overcurrent protection devices from the TS does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes to relocate the requirements from Technical Specifications to the TRM will not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. The proposed changes will not introduce any new failure modes that could result in a new accident. Also, the response of the plant and the operators following the design basis accidents is unaffected by the changes. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? The proposed changes will relocate the requirements for containment penetration conductor overcurrent protective devices from Technical Specifications to the TRM. VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 Any future changes to the relocated requirements will be in accordance with 10 CFR 50.59 and approved station procedures. The proposed changes will have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. The plant response to the design basis accidents will not change. In addition, the relocated requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on items for which Technical Specifications must be established. Therefore, this change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1700 K Street, NW., Washington, DC 20006–3817. NRC Section Chief: David Terao. Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50–278, Peach Bottom Atomic Power Station, Unit 3, York and Lancaster Counties, Pennsylvania Date of application for amendment: July 6, 2005. Description of amendment request: The proposed changes extend the use of the Peach Bottom Atomic Power Station, Unit 3, pressure-temperature (P–T) limits specified in the Technical Specifications (TSs) from 22 to 32 effective full power 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: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to the technical specifications to extend the use of the existing pressure-temperature (P–T) limits does not affect the operation or configuration of any plant equipment. Thus, no new accident initiators are created by this change. The proposed P–T limits are based on the projected reactor vessel neutron fluence at 32 effective full power years (EFPY) of operation. A bounding calculation of reactor vessel 32 EFPY fast neutron fluence has been completed for Peach Bottom Atomic Power Station (PBAPS), Unit 3, using the methodology described in a General Electric (GE) Company Licensing Topical Report (LTR), which adheres to the guidance in Regulatory Guide 1.190, ‘‘Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.’’ The three- PO 00000 Frm 00080 Fmt 4703 Sfmt 4703 dimensional spatial distribution of neutron flux was modeled by combining the results of two separate two-dimensional neutron transport calculations. The latest available cross section libraries for the important components of Boiling Water Reactor (BWR) neutron flux calculations, i.e., oxygen, hydrogen and individual iron isotopes, were included. The resulting reactor vessel fast neutron fluence value was then used in concert with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), Section XI, Case –640 and ASME Code, Section XI, Appendix G, paragraph G–2214.1 to develop updated P–T curves. A comparison of the updated P– T curves with the existing PBAPS, Unit 3 curves indicates that the existing curves are bounding through 32 EFPY. This provides sufficient assurance that the PBAPS, Unit 3, reactor vessel will be operated in a manner that will protect it from brittle fracture under all operating conditions. Therefore, 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 changes to the technical specifications to extend the use of the existing P–T limits do not affect the operation or configuration of any plant equipment. The proposed P–T limits will remain valid and conservative throughout the proposed extension. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The proposed changes extend the use of the existing P–T limits. The proposed P–T limits are based on the projected reactor vessel neutron fluence at 32 EFPY of operation. A bounding calculation of reactor vessel 32 EFPY fast neutron fluence has been completed for PBAPS, Unit 3, using the NRC approved methodology in a GE LTR, which adheres to the guidance in Regulatory Guide 1.190. The three-dimensional spatial distribution of neutron flux was modeled by combining the results of two separate twodimensional neutron transport calculations. The latest available cross section libraries for the important components of BWR neutron flux calculations, i.e., oxygen, hydrogen and individual iron isotopes, were included. The resulting reactor vessel fast neutron fluence value was then used in concert with ASME Code Case –640 and ASME Code, Section XI, Appendix G, paragraph G–2214.1 to develop updated P–T curves. A comparison of the updated P–T curves with the existing PBAPS, Unit 3 curves indicates that the existing curves are bounding through 32 EFPY. This provides sufficient margin such that the PBAPS, Unit 3, reactor vessel will be operated in a manner that will protect it from brittle fracture under all operating conditions. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. E:\FR\FM\02AUN1.SGM 02AUN1 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for Licensee: Thomas S. O’Neill, Associate and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555. NRC Section Chief: Darrell J. Roberts. Omaha Public Power District, Docket No. 50–285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: July 1, 2005. Description of amendment request: The proposed change will amend the design and licensing basis of the Fort Calhoun Station, Unit 1, by revising the updated safety analysis report (USAR) to describe an existing Emergency Operating Procedure (EOP) operator action to isolate steam generator blowdown within 15 minutes of reactor trip during a loss of main feedwater event. 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 the USAR clarifies reliance on operator action which has been utilized since implementation of the EOPs. It does not affect an accident initiator previously evaluated in the USAR or Technical Specifications and will not prevent safety systems from performing their accident mitigating function as discussed in the USAR or Technical Specifications. Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change provides clarification to the existing USAR accident analysis of record. The change does not modify or install any safety related equipment. It does not alter any design or licensing basis assumptions and does not alter any operating procedures other than the explicit specification [of] the time constraint of the 15 minutes. Presently the action is included in EOP–00 without a time constraint. VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change provides clarification to the USAR section 14.10.1 and has no effect on safety margins. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005– 3502. NRC Section Chief: Daniel S. Collins, Acting. 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: July 4, 2005. Description of amendment request: The proposed changes would extend the allowed outage time for Technical Specification (TS) 3/4.7.4, ‘‘Essential Cooling Water System,’’ and the associated TSs for those systems supported by Essential Cooling Water, from 7 days to 14 days. 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. Since only one train of components is affected by the condition and single failure is not considered while a plant is in an LCO [Limiting Condition for Operation] ACTION, the operable ESF [Engineered Safety Feature] trains are adequate to maintain the plant’s design basis. Thus, this condition will not alter assumptions relative to the mitigation of an accident or transient event. Considering compensatory action and risks involved in a plant shutdown, STPNOC [STP Nuclear Operating Company] has determined that there is no significant risk associated with extending the Allowed Outage Time for the Essential Cooling Water System and the systems it supports for an additional 7 days. Additionally, the proposed change to remove the one-time note from TS 3.7.4 is considered PO 00000 Frm 00081 Fmt 4703 Sfmt 4703 44403 an administrative change and does not impact the probability or consequences of any accident previously evaluated. Based on this evaluation, there is no 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 accident from any accident previously evaluated? Response: No. This proposed change only extends an Allowed Outage Time and will not physically alter the plant. No new or different type of equipment will be installed by this action. The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. No change to the system[s] as evaluated in the South Texas Project safety analysis is proposed. The proposed change to remove the one-time note from TS 3.7.4 is considered an administrative change and does not create the possibility of a new or different kind of accident previously evaluated. Therefore, this 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. Considering compensatory action and risks involved in a plant shutdown, STPNOC has determined that there is no significant risk associated with extending the Allowed Outage Time for the Essential Cooling Water System and the systems it supports for an additional 7 days. Based on the availability of redundant systems, the compensatory actions that will be taken, and the extremely low probability of an accident that could not be mitigated by the available systems, STPNOC concludes that there is no significant reduction in the margin of safety. The proposed change to remove the one-time note from TS 3.7.4 is considered an administrative change and does not impact any 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. 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: July 4, 2005. Description of amendment request: The proposed change to Technical Specification 4.0.5 would add a reference to the NRC-approved E:\FR\FM\02AUN1.SGM 02AUN1 44404 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices exemption of selected pumps, valves, and other components from special treatment requirements. As an editorial change, references to Title 10, Code of Federal Regulations (10 CFR) Part 50, Section 50.55a(f) and 10 CFR Part 50, Section 50.55a(f)(6)(i) would be added to the paragraph for inservice testing, similar to the existing references for inservice inspection. In addition, ‘‘inservice testing’’ and ‘‘inservice inspection’’ would be reordered for consistency with the sequence of the regulations in 10 CFR Part 50, Section 50.55a. 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? No. Including the reference to the exemption in the Technical Specifications establishes consistency between the surveillance requirements for inservice inspection and testing and the exemption as approved by the NRC. There are no changes in the inspection and testing procedures as a result of adding the reference because the exemption already removes low safety significance and non-risk significant components from the requirements for special treatment. The proposed changes are administrative in nature and do not have a significant adverse effect on plant operation or personnel safety. Consequently, the changes will not affect 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? No. Including the reference to the exemption in the Technical Specifications establishes consistency between the surveillance requirements for inservice inspection and testing and the exemption as approved by the NRC. There are no changes in the inspection and testing procedures as a result of adding the reference because the exemption already removes low safety significance and non-risk significant components from the requirements for special treatment. The proposed changes are administrative in nature and do not have a significant adverse effect on plant operation or personnel safety. Consequently, the 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? No. Including the reference to the exemption in the Technical Specifications establishes consistency between the surveillance requirements for inservice inspection and testing and the exemption as VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 approved by the NRC. There are no changes in the inspection and testing procedures as a result of adding the reference because the exemption already removes low safety significance and non-risk significant components from the requirements for special treatment. The proposed changes are administrative in nature and do not have a significant adverse effect on plant operation or personnel safety. Consequently, the changes do not significantly reduce 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. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration. For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice. Tennessee Valley Authority, Docket No. 50–259 , Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama Date of application for amendments: June 28, 2004, as supplemented February 23 and April 25, 2005. Description of amendments request: The proposed amendment would change the operating license to increase the maximum authorized power level from 3293 megawatts thermal (MWt) to 3952 MWt; an increase of approximately 20 percent. The amendment would also change the licensing bases and any associated Technical Specifications for containment overpressure, the maximum ultimate heat sink temperature, and the upper bound peak cladding temperature. PO 00000 Frm 00082 Fmt 4703 Sfmt 4703 Date of publication of individual notice in the Federal Register: July 11, 2005 (70 FR 39803). Expiration date of individual notice: August 10, 2005 (Public comments) and September 9, 2005 (Hearing requests). Tennessee Valley Authority, Docket Nos. 50–260 and 50–296, Browns Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama Date of application for amendments: June 25, 2004, as supplemented February 23 and April 25, 2005. Description of amendments request: The proposed amendments would change the operating licences to increase the maximum authorized power level from 3458 megawatts thermal (MWt) to 3952 MWt; an increase of approximately 15 percent. The amendment would also change the licensing bases and any associated Technical Specifications for containment overpressure. Date of publication of individual notice in the Federal Register: July 12, 2005 (70 FR 40064). Expiration date of individual notice: August 11, 2005 (Public comments) and September 12, 2005 (Hearing requests). Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has E:\FR\FM\02AUN1.SGM 02AUN1 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. AmerGen Energy Company, LLC, et al., Docket No. 50–219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Date of application for amendment: March 25, 2005, as supplemented on June 10, 2005. Brief description of amendment: The amendment revised Section 3.7, ‘‘Auxiliary Electrical Power,’’ of the Technical Specifications to reflect the capability upgrade of one of the offsite power supply lines from 69 kilovolts (KV) to 230 KV. Date of Issuance: July 14, 2005. Effective date: July 14, 2005 and shall be implemented as soon as the upgraded offsite supply line is placed in service. Amendment No.: 256. Facility Operating License No. DPR– 16: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: April 12, 2005 (70 FR 19113). The June 10, 2005, letter provided clarifying information within the scope of the original application and did not change the staff’s initial proposed no significant hazards consideration determination. The Commission’s related evaluation of this amendment is contained in a Safety Evaluation dated July 14, 2005. No significant hazards consideration comments received: No. 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 application for amendment: October 15, 2004. VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 Brief description of amendment: This amendment revises Technical Specifications by extending the inspection interval for reactor coolant pump flywheels to 20 years. Date of issuance: June 21, 2005. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 119. Facility Operating License No. NPF– 63. Amendment revises the Technical Specifications. Date of initial notice in Federal Register: March 1, 2005 (70 FR 9988). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated June 21, 2005. No significant hazards consideration comments received: No. 44405 Renewed Facility Operating License Nos. DPR–38, DPR–47, and DPR–55: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: May 10, 2005 (70 FR 24649) The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated July 12, 2005. No significant hazards consideration comments received: No. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50–416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of application for amendment: December 17, 2004. Brief description of amendment: The proposed change revises the air lock Duke Energy Corporation, Docket Nos. surveillance test acceptance criteria to 50–269, 50–270, and 50–287, Oconee be consistent with the NRC approved Nuclear Station, Units 1, 2, and 3, Industry Technical Specification Task Oconee County, South Carolina Force (TSTF) change to the Standard Date of application of amendments: Technical Specifications TSTF–52, February 14, 2005. entitled, ‘‘Implement 10 CFR [Part] 50, Brief description of amendments: The Appendix J, Option B.’’ By letter dated amendments revised the Technical April 6, 1998, the NRC Staff issued Specification Surveillance Requirement amendment number 135 to the Grand 3.3.7.1 to extend the frequency of the Gulf Nuclear Station license permitting channel functional test for the the implementation of the containment Engineered Safeguards Protective leak rate testing provisions of 10 CFR System digital actuation logic channels Part 50, Appendix J, Option B. from once every 31 days to once every Date of issuance: July 12, 2005. 92 days. Effective date: As of the date of Date of Issuance: May 19, 2005. issuance and shall be implemented Effective date: As of the date of within 60 days of issuance. issuance and shall be implemented Amendment No: 168. within 90 days. Facility Operating License No. NPF– Amendment Nos.: 345, 347 and 346. 29: The amendment revises the Renewed Facility Operating License Technical Specifications. Nos. DPR–38, DPR–47, and DPR–55: Date of initial notice in Federal Amendments revised the Technical Register: February 1, 2005 (70 FR 5242). Specifications. The Commission’s related evaluation Date of initial notice in Federal Register: March 15, 2005 (70 FR 12745). of the amendment is contained in a Safety Evaluation dated July 12, 2005. The Commission’s related evaluation No significant hazards consideration of the amendments is contained in a comments received: No. Safety Evaluation dated May 19, 2005. No significant hazards consideration FPL Energy Seabrook, LLC, Docket No. comments received: No. 50–443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire Duke Energy Corporation, Docket Nos. 50–269, 50–270, and 50–287, Oconee Date of amendment request: February Nuclear Station, Units 1, 2, and 3, 4, 2004, as supplemented by letter dated Oconee County, South Carolina March 16, 2005. Description of amendment request: Date of application of amendments: The amendment modified the Seabrook March 14, 2005. Brief description of amendments: The Station Technical Specification (TS) Index; TS Table 3.3–10, ‘‘Accident amendments deleted Technical Monitoring Instrumentation’’; TS Table Specification 5.5.4, ‘‘Post Accident 4.4–2, ‘‘Steam Generator Tube Sampling.’’ Inspection’’; TS 6.0, ‘‘Administrative Date of Issuance: July 12, 2005. Controls’’; and Appendix B to Facility Effective date: As of the date of Operating License (FOL) No. NPF–86, issuance and shall be implemented ‘‘Environmental Protection Plan’’. within 180 days. Date of issuance: July 18, 2005. Amendment Nos.: 346, 348, and 347. PO 00000 Frm 00083 Fmt 4703 Sfmt 4703 E:\FR\FM\02AUN1.SGM 02AUN1 44406 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices Facility Operating License Nos. NPF– Effective date: As of its date of 10 and NPF–15: The amendments issuance, and shall be implemented revised the Technical Specifications. within 90 days. Amendment No.: 104. Date of initial notice in Federal Facility Operating License No. NPF– Register: August 3, 2004 (69 FR 46588). 86: The amendment revised the TSs and The supplemental letter dated June 14, Appendix B to the FOL. 2005, provided additional information Date of initial notice in Federal that clarified the application, did not Register: March 2, 2004 (69 FR 9861). Theexpand the scope of the application as March 16, 2005, supplement provided originally noticed, and did not change clarifying information that did not the NRC staff’s original proposed no change the scope of the proposed significant hazards consideration amendment as described in the original determination. notice of proposed action published in The Commission’s related evaluation the Federal Register, and did not of the amendments is contained in a change the initial proposed no Safety Evaluation dated July 19, 2005. significant hazards consideration No significant hazards consideration determination. The Commission’s comments received: No. related evaluation of the amendment is TXU Generation Company LP, Docket contained in a Safety Evaluation dated Nos. 50–445 and 50–446, Comanche July 18, 2005. Peak Steam Electric Station, Unit Nos. No significant hazards consideration 1 and 2, Somervell County, Texas comments received: No. Date of amendment request: October Southern California Edison Company, et 13, 2004. al., Docket Nos. 50–361 and 50–362, Brief description of amendments: The San Onofre Nuclear Generating Station, amendments revise Technical Units 2 and 3, San Diego County, Specification (TS) 5.6.5b by adding two California topical reports (TRs) into the list of Date of application for amendments: approved analytical methods used to June 29, 2004, as supplemented by letter determine the core operating limits, dated June 14, 2005. deleting four TRs for analytical methods Brief description of amendments: The no longer used to determine the core proposed changes revise the Technical operating limits, and sequentially Specifications (TSs) to implement the renumbering the remaining approved following miscellaneous TS changes: analytical methods in TS 5.6.5b. Revise TS 2.2.5 Safety Limit Violations Date of issuance: July 13, 2005. Licensee Event Report reporting period Effective date: As of the date of from 30 days to 60 days; revise 3.4.3.1.2 issuance and shall be implemented Pressurizer Heatup/Cooldown Limits within 60 days from the date of Surveillance Requirements frequency to issuance. reflect pressurizer spray cyclic limits Amendment Nos.: 119, 119. being governed by the temperature Facility Operating License Nos. NPF– differentials between the spray nozzle 87 and NPF–89: The amendments and the spray line; revise TS 5.5.2.11 revised the Technical Specifications. Steam Generator Tube Surveillance Date of initial notice in Federal requirements to correct typographical Register: December 21, 2004 (69 FR errors; remove TS 5.5.2.14 Configuration 76495). The Commission’s related evaluation Risk Management Program in of the amendments is contained in a accordance with Federal Register Safety Evaluation dated July 13, 2005. Notice Vol. 64, No. 137 (64 FR 38551, No significant hazards consideration July 19, 1999); and revise TS 5.7.1.5 Core Operating Limits Report (COLR) to comments received: No. delete revision numbers and dates from Virginia Electric and Power Company, the referenced documents in this Docket Nos. 50–338 and 50–339, North section, consistent with the NRC Anna Power Station, Units 1 and 2, approved industry Technical Louisa County, Virginia Specifications Task Force (TSTF) Date of application for amendment: Standard Technical Specifications July 1, 2004, as supplemented by letters Traveler number TSTF–363, ‘‘Revise dated and October 28, 2004, and Topical Report References in ITS November 16, 2004. (Improved Technical Specifications) Brief description of amendment: 5.6.5 COLR.’’ These amendments revise the reactor Date of issuance: July 19, 2005. coolant pressure and temperature limits, Effective date: As of the date of low-temperature overpressure issuance and shall be implemented protection system (LTOPS) setpoint within 60 days from the date of values, and LTOPS enable temperatures issuance. that are valid for 50.3 effective fullAmendment Nos.: 197, 188. VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 PO 00000 Frm 00084 Fmt 4703 Sfmt 4703 power years (EFPY) and 52.3 EFPY of operation for North Anna, Units 1 and 2, respectively. Date of issuance: July 8, 2005. Effective date: As of the date of issuance and shall be implemented within 6 months from the date of issuance. Amendment Nos.: 242 and 223. Renewed Facility Operating License Nos. NPF–4 and NPF–7: Amendments change the Technical Specifications. Date of initial notice in Federal Register: August 31, 2004 (69 FR 53114). The supplements dated October 28, 2004, and November 16, 2004, 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 July 8, 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: November 4, 2004, as supplemented on February 21 and June 2, 2005. Brief Description of amendments: These amendments revise the Technical Specifications (TS) to delete the Inservice Inspection (ISI) and Inservice Testing (IST) requirements in TS 4.0.5; relocate the IST requirements to the administrative section of the TS as a program; revise the TS to reference the IST program instead of TS 4.0.5; delete the individual TS references to the ISI program; and add a TS Bases Control Program to the TS Administrative Controls section. Date of issuance: July 15, 2005. Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment Nos.: 243 and 242. Renewed Facility Operating License Nos. DPR–32 and DPR–37: Amendments change the Technical Specifications. Date of initial notice in Federal Register: February 15, 2005 (70 FR 7771). The February 21 and June 2, 2005, supplements contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated July 15, 2005. No significant hazards consideration comments received: No. E:\FR\FM\02AUN1.SGM 02AUN1 Federal Register / Vol. 70, No. 147 / Tuesday, August 2, 2005 / Notices Dated at Rockville, Maryland, this 25th day of July 2005. For the Nuclear Regulatory Commission Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation. [FR Doc. E5–4067 Filed 8–1–05; 8:45 am] II. On September 11, 2001, terrorists simultaneously attacked targets in New York, NY, and Washington, DC, utilizing large commercial aircraft as weapons. In response to the attacks and intelligence information subsequently obtained, the Commission issued a number of Safeguards and Threat Advisories to Licensees in order to strengthen Licensees’ capabilities and readiness to respond to a potential attack on this regulated activity. The Commission has also communicated with other Federal, State and local government agencies and industry representatives to discuss and evaluate the current threat environment in order to assess the adequacy of the current security measures. In addition, the Commission commenced a comprehensive review of its safeguards and security programs and requirements. As a result of its initial consideration of current safeguards and security requirements, as well as a review of information provided by the intelligence community, the Commission has determined that certain security measures are required to be implemented by Licensees as prudent, interim measures to address the current threat environment in a consistent manner. Therefore, the Commission is imposing requirements, as set forth in Attachment B 2 of this Order, on all Licensees identified in Attachment A of this Order. These additional security measures, which supplement existing regulatory requirements, will provide the Commission with reasonable assurance that the common defense and security continue to be adequately protected in the current threat environment. These additional security measures will remain in effect until the Commission determines otherwise. The Commission recognizes that Licensees may have already initiated many of the measures set forth in Attachment B to this Order in response to previously issued Safeguards and Threat Advisories or on their own. It is also recognized that some measures may not be possible or necessary for all shipments of radioactive material quantities of concern, or may need to be tailored to accommodate the Licensees’ specific circumstances to achieve the intended objectives and avoid any unforeseen effect on the safe transport of radioactive material quantities of concern. Although the security measures implemented by Licensees in response to the Safeguards and Threat Advisories have been adequate to provide reasonable assurance of adequate protection of common defense and security, in light of the continuing threat environment, the Commission concludes that the security measures must be embodied in an Order, consistent with the established regulatory framework. The Commission has determined that the security measures contained in Attachment B of this Order contains Safeguards Information and will not be released to the public as per Order entitled, ‘‘Issuance of Order Imposing Requirements for Protecting Certain Safeguards Information,’’ issued on 1 Attachment A contains sensitive unclassified information and will not be released to the public. 2 Attachment B contains Safeguards Information and will not be released to the public. BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION [EA–05–006] In the Matter of Certain Licensees Authorized To Possess and Transfer Items Containing Radioactive Material Quantities of Concern; Order Imposing Additional Security Measures (Effective Immediately) I. The Licensees identified in Attachment A 1 to this Order, hold licenses issued by the U.S. Nuclear Regulatory Commission (NRC or Commission) or an Agreement State, in accordance with the Atomic Energy Act of 1954, as amended, and 10 CFR parts 30, 32, 70 and 71, or equivalent Agreement State regulations. The licenses authorize them to possess and transfer items containing radioactive material quantities of concern. This Order is being issued to all such Licensees who may transport radioactive material quantities of concern under the NRC’s authority to protect the common defense and security, which has not been relinquished to the Agreement States. The Orders require compliance with specific additional security measures to enhance the security for transport of certain radioactive material quantities of concern. VerDate jul<14>2003 17:21 Aug 01, 2005 Jkt 205001 PO 00000 Frm 00085 Fmt 4703 Sfmt 4703 44407 November 5, 2004. To provide assurance that Licensees are implementing prudent measures to achieve a consistent level of protection to address the current threat environment, all licensees identified in Attachment A to this Order shall implement the requirements identified in Attachment B to this Order. In addition, pursuant to 10 CFR 2.202, I find that in light of the common defense and security matters identified above, which warrant the issuance of this Order, the public health and safety require that this Order be immediately effective. III. Accordingly, pursuant to Sections 53, 63, 81, 161b, 161i, 161o, 182 and 186 of the Atomic Energy Act of 1954, as amended, and the Commission’s regulations in 10 CFR 2.202 and 10 CFR parts 30, 32, 70 and 71, it is hereby ordered, effective immediately, that all licensees identified in attachment a to this order shall comply with the following: A. All Licensees shall, notwithstanding the provisions of any Commission or Agreement State regulation or license to the contrary, comply with the requirements described in Attachment B to this Order. The Licensees shall immediately start implementation of the requirements in Attachment B to the Order and shall complete implementation by January 17, 2006, or before the licensee’s next shipment after the 180 day implementation period of this Order. This Order supersedes the additional transportation security measures prescribed in the Manufacturer and Distributor Order issued January 12, 2004. B.1. All Licensees shall, within twenty (20) days of the date of this Order, notify the Commission, (1) if they are unable to comply with any of the requirements described in Attachment B, (2) if compliance with any of the requirements is unnecessary in their specific circumstances, or (3) if implementation of any of the requirements would cause the Licensee to be in violation of the provisions of any Commission or Agreement State regulation or its license. The notification shall provide the Licensees’ justification for seeking relief from or variation of any specific requirement. 2. Any Licensee that considers that implementation of any of the requirements described in Attachment B to this Order would adversely impact the safe transport of radioactive material quantities of concern must notify the Commission, within twenty (20) days of E:\FR\FM\02AUN1.SGM 02AUN1
[Pages 44400-44407]
[FR Doc No: E5-4067]
proposed to be issued from July 8, 2005, to July 21, 2005. The last
biweekly notice was published on July 19, 2005 (70 FR 41442).
[[Page 44401]]
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina and Docket
Technical Specification 3.9.1, ``Boron Concentration,'' to clarify the
technical requirements for boron concentration when the refueling canal
and the refueling cavity are not connected to the reactor coolant
1. Would implementation of the changes proposed in this LAR
[License Amendment Request] involve a significant increase in the
No. This LAR clarifies Technical Specification [TS] 3.9.1
regarding the applicability of boron concentration limits when the
refueling canal and refueling cavity are not connected to the
reactor coolant system [RCS]. When the refueling canal and the
refueling cavity are isolated from the RCS, no potential path for
boron dilution of the RCS exists, thus there is no significant
increase in the probability of an accident that has been previously
evaluated, nor would there be a significant increase in the
consequences of an accident that has been previously evaluated.
2. Would implementation of the changes proposed in this LAR
any accident previously evaluated?
No. The change proposed in this LAR clarifies the applicability
of TS 3.9.1 when the refueling canal and refueling cavity are not
connected to the reactor coolant system. When the refueling canal
and the refueling cavity are isolated from the RCS, no potential
path for boron dilution of the RCS exists, thus there is no means to
initiate an accident that is new or different from any accident that
has been previously evaluated.
3. Would implementation of the changes proposed in this LAR
involve a significant reduction in a margin of safety?
No. The change proposed in this LAR only clarifies the
applicability of TS 3.9.1 when the refueling canal and the refueling
cavity are not connected to the reactor coolant system. [TS 3.9.1
limits the boron concentrations of the reactor coolant system], the
refueling canal, and the refueling cavity to ensure that the reactor
remains subcritical during Mode 6 plant conditions. However, when
the refueling canal and the refueling cavity are isolated from the
reactor coolant system, no potential for boron dilution of the RCS
exists. Therefore, in this condition it is not necessary to place a
limit on the boron concentration in the refueling canal and the
refueling cavity, thus there is no significant reduction in a margin
of safety since no specific boron limits are being changed.
Description of amendment request: Entergy Operations, Inc. (EOI)
has requested a change which would revise
[[Page 44402]]
the requirements associated with the Arkansas Nuclear One, Unit 2 (ANO-
2) containment overcurrent protection devices. EOI proposes to amend
Operating License NPF-6 to eliminate Technical Specifications (TSs)
section 3.8.2.5, ELECTRICAL POWER SYSTEMS-Containment Penetration
Conductor Overcurrent Protection Devices. The proposed change would
relocate the requirements for containment penetration conductor
overcurrent protective devices to the Technical Requirements Manual
The proposed changes to relocate the requirements for
containment penetration conductor overcurrent protective devices
from Technical Specifications to the TRM will have no adverse effect
on plant operation, or the availability or operation of any accident
mitigation equipment. The plant response to the design basis
accidents will not change. Operation of the containment penetration
conductor overcurrent protective devices is not an accident
initiator and can not cause an accident. Whether the requirements
for the containment penetration conductor overcurrent protective
devices are located in Technical Specifications or the TRM will have
no effect on the probability or consequences of any accident
Therefore, the removal of overcurrent protection devices from
the TS does not involve a significant increase in the probability or
The proposed changes to relocate the requirements from Technical
Specifications to the TRM will not alter the plant configuration (no
new or unusual operator actions. The proposed changes will not
design basis accidents is unaffected by the changes.
The proposed changes will relocate the requirements for
from Technical Specifications to the TRM. Any future changes to the
relocated requirements will be in accordance with 10 CFR 50.59 and
approved station procedures. The proposed changes will have no
adverse effect on plant operation, or the availability or operation
of any accident mitigation equipment. The plant response to the
design basis accidents will not change. In addition, the relocated
requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on
items for which Technical Specifications must be established.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station, Unit 3, York and Lancaster
Description of amendment request: The proposed changes extend the
use of the Peach Bottom Atomic Power Station, Unit 3, pressure-
temperature (P-T) limits specified in the Technical Specifications
(TSs) from 22 to 32 effective full power years.
Response: No. The proposed changes to the technical
specifications to extend the use of the existing pressure-
temperature (P-T) limits does not affect the operation or
configuration of any plant equipment. Thus, no new accident
initiators are created by this change. The proposed P-T limits are
based on the projected reactor vessel neutron fluence at 32
effective full power years (EFPY) of operation. A bounding
calculation of reactor vessel 32 EFPY fast neutron fluence has been
completed for Peach Bottom Atomic Power Station (PBAPS), Unit 3,
using the methodology described in a General Electric (GE) Company
Licensing Topical Report (LTR), which adheres to the guidance in
Regulatory Guide 1.190, ``Calculational and Dosimetry Methods for
Determining Pressure Vessel Neutron Fluence.'' The three-dimensional
spatial distribution of neutron flux was modeled by combining the
results of two separate two-dimensional neutron transport
calculations. The latest available cross section libraries for the
important components of Boiling Water Reactor (BWR) neutron flux
calculations, i.e., oxygen, hydrogen and individual iron isotopes,
were included. The resulting reactor vessel fast neutron fluence
value was then used in concert with the American Society of
Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),
Section XI, Case -640 and ASME Code, Section XI, Appendix G,
paragraph G-2214.1 to develop updated P-T curves. A comparison of
the updated P-T curves with the existing PBAPS, Unit 3 curves
indicates that the existing curves are bounding through 32 EFPY.
This provides sufficient assurance that the PBAPS, Unit 3, reactor
vessel will be operated in a manner that will protect it from
brittle fracture under all operating conditions.
specifications to extend the use of the existing P-T limits do not
affect the operation or configuration of any plant equipment. The
proposed P-T limits will remain valid and conservative throughout
Response: No. The proposed changes extend the use of the
existing P-T limits. The proposed P-T limits are based on the
projected reactor vessel neutron fluence at 32 EFPY of operation. A
bounding calculation of reactor vessel 32 EFPY fast neutron fluence
has been completed for PBAPS, Unit 3, using the NRC approved
methodology in a GE LTR, which adheres to the guidance in Regulatory
Guide 1.190. The three-dimensional spatial distribution of neutron
flux was modeled by combining the results of two separate two-
dimensional neutron transport calculations. The latest available
cross section libraries for the important components of BWR neutron
flux calculations, i.e., oxygen, hydrogen and individual iron
isotopes, were included. The resulting reactor vessel fast neutron
fluence value was then used in concert with ASME Code Case -640 and
ASME Code, Section XI, Appendix G, paragraph G-2214.1 to develop
updated P-T curves. A comparison of the updated P-T curves with the
existing PBAPS, Unit 3 curves indicates that the existing curves are
bounding through 32 EFPY. This provides sufficient margin such that
the PBAPS, Unit 3, reactor vessel will be operated in a manner that
will protect it from brittle fracture under all operating
Attorney for Licensee: Thomas S. O'Neill, Associate and General
Description of amendment request: The proposed change will amend
the design and licensing basis of the Fort Calhoun Station, Unit 1, by
revising the updated safety analysis report (USAR) to describe an
existing Emergency Operating Procedure (EOP) operator action to isolate
steam generator blowdown within 15 minutes of reactor trip during a
loss of main feedwater event.
The proposed change to the USAR clarifies reliance on operator
action which has been utilized since implementation of the EOPs. It
does not affect an accident initiator previously evaluated in the
USAR or Technical Specifications and will not prevent safety systems
from performing their accident mitigating function as discussed in
the USAR or Technical Specifications.
The proposed change provides clarification to the existing USAR
accident analysis of record. The change does not modify or install
any safety related equipment. It does not alter any design or
licensing basis assumptions and does not alter any operating
procedures other than the explicit specification [of] the time
constraint of the 15 minutes. Presently the action is included in
EOP-00 without a time constraint.
The proposed change provides clarification to the USAR section
14.10.1 and has no effect on safety margins.
Description of amendment request: The proposed changes would extend
the allowed outage time for Technical Specification (TS) 3/4.7.4,
``Essential Cooling Water System,'' and the associated TSs for those
systems supported by Essential Cooling Water, from 7 days to 14 days.
Since only one train of components is affected by the condition
and single failure is not considered while a plant is in an LCO
[Limiting Condition for Operation] ACTION, the operable ESF
[Engineered Safety Feature] trains are adequate to maintain the
plant's design basis. Thus, this condition will not alter
Considering compensatory action and risks involved in a plant
shutdown, STPNOC [STP Nuclear Operating Company] has determined that
there is no significant risk associated with extending the Allowed
Outage Time for the Essential Cooling Water System and the systems
it supports for an additional 7 days. Additionally, the proposed
change to remove the one-time note from TS 3.7.4 is considered an
administrative change and does not impact the probability or
Based on this evaluation, there is no significant increase in
the probability or consequence of an accident previously evaluated.
This proposed change only extends an Allowed Outage Time and
will not physically alter the plant. No new or different type of
equipment will be installed by this action. The changes in methods
governing normal plant operation are consistent with current safety
analysis assumptions. No change to the system[s] as evaluated in the
South Texas Project safety analysis is proposed. The proposed change
to remove the one-time note from TS 3.7.4 is considered an
administrative change and does not create the possibility of a new
or different kind of accident previously evaluated.
Therefore, this proposed change[ does not] create the
shutdown, STPNOC has determined that there is no significant risk
associated with extending the Allowed Outage Time for the Essential
Cooling Water System and the systems it supports for an additional 7
Based on the availability of redundant systems, the compensatory
actions that will be taken, and the extremely low probability of an
accident that could not be mitigated by the available systems,
STPNOC concludes that there is no significant reduction in the
margin of safety. The proposed change to remove the one-time note
from TS 3.7.4 is considered an administrative change and does not
impact any margin of safety.
Description of amendment request: The proposed change to Technical
Specification 4.0.5 would add a reference to the NRC-approved
exemption of selected pumps, valves, and other components from special
treatment requirements. As an editorial change, references to Title 10,
Code of Federal Regulations (10 CFR) Part 50, Section 50.55a(f) and
10 CFR Part 50, Section 50.55a(f)(6)(i) would be added to the
paragraph for inservice testing, similar to the existing references for
inservice inspection. In addition, ``inservice testing'' and
``inservice inspection'' would be reordered for consistency with the
sequence of the regulations in 10 CFR Part 50, Section 50.55a.
No. Including the reference to the exemption in the Technical
Specifications establishes consistency between the surveillance
requirements for inservice inspection and testing and the exemption
as approved by the NRC. There are no changes in the inspection and
testing procedures as a result of adding the reference because the
exemption already removes low safety significance and non-risk
significant components from the requirements for special treatment.
The proposed changes are administrative in nature and do not have a
significant adverse effect on plant operation or personnel safety.
Consequently, the changes will not affect the probability or
Consequently, the changes do not create the possibility of a new or
Consequently, the changes do not significantly reduce a margin of
Tennessee Valley Authority, Docket No. 50-259 , Browns Ferry Nuclear
Date of application for amendments: June 28, 2004, as supplemented
February 23 and April 25, 2005.
change the operating license to increase the maximum authorized power
level from 3293 megawatts thermal (MWt) to 3952 MWt; an increase of
approximately 20 percent. The amendment would also change the licensing
bases and any associated Technical Specifications for containment
overpressure, the maximum ultimate heat sink temperature, and the upper
bound peak cladding temperature.
July 11, 2005 (70 FR 39803).
Expiration date of individual notice: August 10, 2005 (Public
comments) and September 9, 2005 (Hearing requests).
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: June 25, 2004, as supplemented
Description of amendments request: The proposed amendments would
change the operating licences to increase the maximum authorized power
level from 3458 megawatts thermal (MWt) to 3952 MWt; an increase of
approximately 15 percent. The amendment would also change the licensing
July 12, 2005 (70 FR 40064).
Expiration date of individual notice: August 11, 2005 (Public
comments) and September 12, 2005 (Hearing requests).
10 CFR 51.12(b) and has
Date of application for amendment: March 25, 2005, as supplemented
Brief description of amendment: The amendment revised Section 3.7,
``Auxiliary Electrical Power,'' of the Technical Specifications to
reflect the capability upgrade of one of the offsite power supply lines
from 69 kilovolts (KV) to 230 KV.
Effective date: July 14, 2005 and shall be implemented as soon as
the upgraded offsite supply line is placed in service.
Facility Operating License No. DPR-16: Amendment revised the
The June 10, 2005, letter provided clarifying information within
the scope of the original application and did not change the staff's
The Commission's related evaluation of this amendment is contained in a
Safety Evaluation dated July 14, 2005.
Specifications by extending the inspection interval for reactor coolant
pump flywheels to 20 years.
Facility Operating License No. NPF-63. Amendment revises the
9988).
in a Safety Evaluation dated June 21, 2005.
Technical Specification Surveillance Requirement 3.3.7.1 to extend the
frequency of the channel functional test for the Engineered Safeguards
Protective System digital actuation logic channels from once every 31
days to once every 92 days.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12745).
in a Safety Evaluation dated May 19, 2005.
Specification 5.5.4, ``Post Accident Sampling.''
24649)
in a Safety Evaluation dated July 12, 2005.
Brief description of amendment: The proposed change revises the air
lock surveillance test acceptance criteria to be consistent with the
NRC approved Industry Technical Specification Task Force (TSTF) change
to the Standard Technical Specifications TSTF-52, entitled, ``Implement
10 CFR [Part] 50, Appendix J, Option B.'' By letter dated April 6,
1998, the NRC Staff issued amendment number 135 to the Grand Gulf
Nuclear Station license permitting the implementation of the
containment leak rate testing provisions of 10 CFR Part 50, Appendix J,
Date of initial notice in Federal Register: February 1, 2005 (70 FR
Date of amendment request: February 4, 2004, as supplemented by
letter dated March 16, 2005.
Description of amendment request: The amendment modified the
Seabrook Station Technical Specification (TS) Index; TS Table 3.3-10,
``Accident Monitoring Instrumentation''; TS Table 4.4-2, ``Steam
Generator Tube Inspection''; TS 6.0, ``Administrative Controls''; and
Appendix B to Facility Operating License (FOL) No. NPF-86,
``Environmental Protection Plan''.
Date of issuance: July 18, 2005.
TSs and Appendix B to the FOL.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9861). The March 16, 2005, supplement provided clarifying information
that did not change the scope of the proposed amendment as described in
the original notice of proposed action published in the Federal
Register, and did not change the initial proposed no significant
hazards consideration determination. The Commission's related
Date of application for amendments: June 29, 2004, as supplemented
by letter dated June 14, 2005.
Brief description of amendments: The proposed changes revise the
Technical Specifications (TSs) to implement the following miscellaneous
TS changes: Revise TS 2.2.5 Safety Limit Violations Licensee Event
Report reporting period from 30 days to 60 days; revise 3.4.3.1.2
Pressurizer Heatup/Cooldown Limits Surveillance Requirements frequency
to reflect pressurizer spray cyclic limits being governed by the
temperature differentials between the spray nozzle and the spray line;
revise TS 5.5.2.11 Steam Generator Tube Surveillance requirements to
correct typographical errors; remove TS 5.5.2.14 Configuration Risk
Management Program in accordance with Federal Register Notice Vol. 64,
No. 137 (64 FR 38551, July 19, 1999); and revise TS 5.7.1.5 Core
Operating Limits Report (COLR) to delete revision numbers and dates
from the referenced documents in this section, consistent with the NRC
Technical Specifications Traveler number TSTF-363, ``Revise Topical
Report References in ITS (Improved Technical Specifications) 5.6.5
COLR.''
46588). The supplemental letter dated June 14, 2005, provided
in a Safety Evaluation dated July 19, 2005.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Specification (TS) 5.6.5b by adding two topical reports (TRs) into the
list of approved analytical methods used to determine the core
operating limits, deleting four TRs for analytical methods no longer
used to determine the core operating limits, and sequentially
renumbering the remaining approved analytical methods in TS 5.6.5b.
FR 76495).
in a Safety Evaluation dated July 13, 2005.
Date of application for amendment: July 1, 2004, as supplemented by
letters dated and October 28, 2004, and November 16, 2004.
Brief description of amendment: These amendments revise the reactor
coolant pressure and temperature limits, low-temperature overpressure
protection system (LTOPS) setpoint values, and LTOPS enable
temperatures that are valid for 50.3 effective full-power years (EFPY)
and 52.3 EFPY of operation for North Anna, Units 1 and 2, respectively.
within 6 months from the date of issuance.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53114). The supplements dated October 28, 2004, and November 16, 2004,
contained clarifying information only and did not change the initial no
significant hazards consideration determination or expand the scope of
in a Safety Evaluation dated July 8, 2005.
Date of application for amendments: November 4, 2004, as
supplemented on February 21 and June 2, 2005.
Technical Specifications (TS) to delete the Inservice Inspection (ISI)
and Inservice Testing (IST) requirements in TS 4.0.5; relocate the IST
requirements to the administrative section of the TS as a program;
revise the TS to reference the IST program instead of TS 4.0.5; delete
the individual TS references to the ISI program; and add a TS Bases
Control Program to the TS Administrative Controls section.
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7771). The February 21 and June 2, 2005, supplements contained
clarifying information only and did not change the initial proposed no
in a Safety Evaluation dated July 15, 2005.