Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 17944-17959 [E7-6632]

Download as PDF 17944 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices Dated: April 5, 2007. Rochelle C. Bavol, Office of the Secretary. [FR Doc. 07–1795 Filed 4–6–07; 11:49 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations cprice-sewell on PROD1PC66 with NOTICES I. Background Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 16, 2007 to March 29, 2007. The last biweekly notice was published on March 27, 2007 (72 FR 14303). 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 Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of 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 00129 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\10APN1.SGM 10APN1 cprice-sewell on PROD1PC66 with NOTICES Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / 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 Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee. Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)–(viii). For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397– 4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. AmerGen Energy Company, LLC, Docket No. 50–461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Date of amendment request: November 13, 2006. Description of amendment request: The proposed amendment changes the technical specification (TS) testing frequency for the surveillance requirement (SR) in TS 3.1.4, ‘‘Control Rod Scram Times.’’ The proposed change revises the test frequency of SR 3.1.4.2, control rod scram time testing, from ‘‘120 days cumulative operation in MODE 1’’ to ‘‘200 days cumulative operation in Mode 1.’’ AmerGen has reviewed the proposed no significant hazards consideration determination published in the Federal Register on August 23, 2004 (69 FR 51864), as part of the consolidated line item improvement process (CLIIP) and has concluded that the proposed determination presented in the notice is applicable to Clinton Power Station, Unit No. 1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant PO 00000 Frm 00130 Fmt 4703 Sfmt 4703 17945 hazards consideration is presented below. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of ‘‘no significant hazards consideration’’ is justified. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. NRC Branch Chief: Russell Gibbs. Detroit Edison Company, Docket No. 50–341, Fermi 2, Monroe County, Michigan Date of amendment request: January 26, 2007. Description of amendment request: The proposed amendment would revise technical specifications (TS) requirements for unavailable barriers by E:\FR\FM\10APN1.SGM 10APN1 17946 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices cprice-sewell on PROD1PC66 with NOTICES adding limiting condition for operation (LCO) 3.0.9. This would establish conditions under which TS systems would remain operable when required physical barriers are not capable of providing their related support function. Also, the proposed amendment would make editorial changes to LCO 3.0.8 to be consistent with the terminology in LCO 3.0.9. 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 by a reference to a generic analysis published in the Federal Register on October 3, 2006 (71 FR 58444), which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of any accident previously evaluated? Response: No. The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. 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 does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 accident from an accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee’s performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation. 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 G. Pettinari, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226–1279. NRC Branch Chief: L. Raghavan. Entergy Nuclear Operations, Inc., Docket No. 50–286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York Date of amendment request: January 18, 2007. Description of amendment request: The proposed amendment would revise the expiration limit for the reactor coolant system Pressure/Temperature (P/T) limit graphs in Technical Specifications (TS); revise the adjusted reference temperature for the reactor vessel; and revise the Low Temperature Overpressure Protection (LTOP) arming temperature value specified in TSs. It would also make editorial changes in the use of inequality signs in TSs associated with the LTOP arming temperature in order to make them consistent. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: PO 00000 Frm 00131 Fmt 4703 Sfmt 4703 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No The proposed change does not affect the accident initiators or mitigation assumptions associated with any of the accidents previously evaluated. Operating restrictions on pressure-temperature conditions for the reactor pressure vessel provide assurance that reactor vessel integrity will be maintained under accident or transient conditions. The proposed change uses approved criteria and analysis methods to update the time period for which the current operating limits remain valid. The LTOP system performs an automatic function by opening relief valves if reactor coolant system pressure reaches a temperature-dependent limit. The proposed change includes establishing a more restrictive temperature limit for when this system must be in service, to reflect the material condition of the reactor vessel at the new EFPY limit proposed for the pressuretemperature graphs. The mitigation function and capability of the LTOP system is not being changed by this request. Therefore, the proposed changes do 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 new accident initiators being introduced by this proposed change. The proposed change does not involve installation of new plant equipment, modification of existing equipment, or changes in the way that plant equipment is operated. Pressure-temperature operating limits depicted by graphs in the technical specifications will not be changed and will continue to be used by plant operators. A change in the LTOP system arming temperature will assure that the graphs remain valid for the proposed new operating period of 27.2 EFPY [effective full power years]. 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 Operating limits on pressure and temperature conditions for the reactor coolant system (RCS) are important to assure that the RCS pressure boundary stresses are within analyzed limits. Margins of safety are inherent in the analysis methods, assumptions, and limits specified in regulations and guidance documents. The proposed change is based on NRC-accepted methods, assumptions and limits and maintains the required margin of 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 E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices 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 Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Acting Branch Chief: Douglas V. Pickett. cprice-sewell on PROD1PC66 with NOTICES Entergy Nuclear Operations, Inc., Docket Nos. 50–247 and 50–286, Indian Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), Westchester County, New York Date of amendment request: March 13, 2007. Description of amendment request: The amendment would revise License Condition 2.K for IP2 and License Condition 2.H for IP3, which require the implementation and maintenance of an approved Fire Protection Program for each unit. 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 changes are strictly an administrative relocation of the specific fire protection SER [safety evaluation report] references and do not modify any requirements of the fire protection programs. Therefore, the proposed changes do 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 changes are strictly an administrative relocation of the specific fire protection SER references and do not modify any requirements of the fire protection programs. 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 proposed changes are strictly an administrative relocation of the specific fire protection SER references and do not modify any requirements of the fire protection programs. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Acting Branch Chief: Douglas V. Pickett. Entergy Nuclear Operations, Inc., Docket No. 50–247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York Date of amendment request: March 22, 2007. Description of amendment request: The proposed amendment will revise the test acceptance criteria specified in Technical Specification Surveillance Requirement (SR) 3.8.1.10 for the diesel generator endurance test. 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. The proposed change revises the acceptance criteria to be applied to an existing surveillance test of the facility emergency diesel generators (DGs). Performing a surveillance test is not an accident initiator and does not increase the probability of an accident occurring. The proposed new acceptance criteria will assure that the DGs are capable of carrying the peak electrical loading assumed in the various existing safety analyses which take credit for the operation of the DGs. Establishing acceptance criteria that bound existing analyses validates the related assumption used in those analyses regarding the capability of equipment to mitigate accident conditions. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? No. The proposed change revises the test acceptance criteria for a specific performance test conducted on the existing DGs. The proposed change does not involve installation of new equipment or modification of existing equipment, so no new equipment failure modes are introduced. The proposed revision to the DG surveillance test acceptance criteria also is not a change to the way that the equipment or facility is operated and no new accident initiators are PO 00000 Frm 00132 Fmt 4703 Sfmt 4703 17947 created. 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? No. The conduct of performance tests on safety-related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed change in the DG technical specification surveillance test acceptance criteria is consistent with values assumed in existing safety analyses and is consistent with the design rating of the DGs. 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 Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Acting Branch Chief: Douglas V. Pickett. Entergy Nuclear Operations, Inc., Docket No. 50–333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: February 15, 2007. Description of amendment request: The proposed changes would revise Technical Specification (TS) 3.10.1 to expand its scope to include provisions for reactor coolant temperature excursions greater than 212 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4, which is defined to be reactor coolant temperature less than or equal to 212 °F. This change was proposed by the industry’s TS Task Force (TSTF) and is designated TSTF–484. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 21, 2006 (71 FR 48561), on possible amendments concerning TSTF–484, including a model safety evaluation and model no significant hazards (NSHC) determination, using the consolidated line item improvement process (CLIIP). The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on October 27, E:\FR\FM\10APN1.SGM 10APN1 17948 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices cprice-sewell on PROD1PC66 with NOTICES 2006 (71 FR 63050). The licensee affirmed the applicability of the following NSHC determination in its application dated February 15, 2007. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1: The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or 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. Criterion 2: The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3: The proposed change does not involve a significant reduction in a margin of safety. Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, the NRC staff concludes that the proposed change VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified. Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Acting Branch Chief: Douglas V. Pickett. Exelon Generation Company, LLC (EGC), Docket Nos. 50–373 and 50–374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois Date of amendment request: November 17, 2006. Description of amendment request: The proposed amendments would replace references to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) with a reference to the ASME Code of Operation and Maintenance of Nuclear Power Plants (OM Code) in Technical Specification (TS) 5.5.7, ‘‘Inservice Testing Program [IST].’’ These proposed changes are consistent with the implementation of the LSCS, Units 1 and 2 third 10-year IST program in accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, ‘‘Codes and standards,’’ paragraph (f), ‘‘Inservice testing requirements.’’ The third 10-year interval for LSCS, Units 1 and 2 is scheduled to start on October 12, 2007. In addition to the replacement of the references, EGC is also adding provisions in TS 5.5.7, item b, to only apply Surveillance Requirement (SR) 3.0.2 to those inservice testing frequencies of two years or less. These proposed changes are based on TS Task Force (TSTF) Traveler No. 479–A (TSTF–479–A), Revision 0, ‘‘Changes to Reflect Revision of 10 CFR 50.55a,’’ as modified by TSTF–497, Revision 0, ‘‘Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less’’ and approved by the NRC in December 6, 2005, and October 4, 2006. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a, ‘‘Codes and PO 00000 Frm 00133 Fmt 4703 Sfmt 4703 standards,’’ paragraph (f) regarding the inservice testing of pumps and calves for the Third 10-year Interval. The current TS reference the [American Society of Mechanical Engineers] ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with 10 CFR 50.55a, paragraph (f), ‘‘Inservice testing requirements,’’ and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request. The proposed changes are administrative in nature, do not affect any accident initiators, do not affect the ability of LSCS to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations. Thus, the radiological consequences of any accident previously evaluated are not increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a(f) regarding the inservice testing of pumps and valves for the Third 10-year Interval. The current TS reference the ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request. The proposed changes to TS Section 5.5.7 do not affect the performance of any LSCS structure, system, or component credited with mitigating any accident previously evaluated and do not introduce any new modes of system operation or failure mechanisms. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? Response: No. The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a(f) regarding the inservice testing of pumps and valves for the Third 10-Year Interval. The current TS reference the ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request. The proposed changes do not modify the safety limits setpoints at which protective actions are initiated and do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of 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 three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. NRC Branch Chief: Russell Gibbs. cprice-sewell on PROD1PC66 with NOTICES Exelon Generation Company, LLC, Docket Nos. 50–352 and 50–353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: June 2, 2006. Description of amendment request: The proposed amendments incorporates revised 10 CFR Part 20 requirements for Limerick Generating Station Units 1 and 2 technical specifications (TSs). 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. Updating the Technical Specifications (TS) to be consistent with 10 CFR Part 20 has no impact on plant structures, systems, or components, does not affect any accident initiators, and does not change any safety analysis. Therefore, the proposed changes do not involve an 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. Updating the TS to be consistent with 10 CFR Part 20 will not change any equipment, VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 require new equipment to be installed, or change the way current equipment operates. No credible new failure mechanisms, malfunctions, or accident initiators are created by the proposed changes. Therefore, the changes do 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. Updating the TS to be consistent with 10 CFR Part 20 does not adversely affect existing plant safety margins or the reliability of equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. 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 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. Brad Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348. NRC Branch Chief: Harold K. Chernoff. FirstEnergy Nuclear Operating Company, et al., Docket No. 50–346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio Date of amendment request: February 12, 2007. Description of amendment request: The proposed license amendment would revise Technical Specification (TS) Limiting Condition for Operation 3.9.4, ‘‘Containment Penetrations’’, to allow penetrations included under TS 3.9.4(c) to be opened during core alterations or movement of irradiated fuel, under administrative controls. This change is based on the TS Task Force Traveler No. 312–A, Revision 1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change would allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core PO 00000 Frm 00134 Fmt 4703 Sfmt 4703 17949 alterations. These penetrations are normally closed during this time period to prevent the release of radioactive material in the event of a Fuel Handling Accident inside containment. These penetrations are not initiators of any accident. The probability of a Fuel Handling Accident is unaffected by the status of these penetrations. The Fuel Handling Accident analyses demonstrate that the maximum offsite dose is well [within] the acceptance limits specified in SRP [Standard Review Plan] 15.7.4, and the control room dose is within the acceptance criteria specified in GDC [General Design Criterion] 19. Furthermore, the existing analysis results are independent of the containment release path, and therefore are unaffected by the proposed change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change will not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a Technical Specification change that will allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core alterations. Open penetrations are not accident initiators, and will not create the possibility of a new kind of accident. Administrative controls will be implemented to ensure the capability to close the affected containment penetrations in the event of a Fuel Handling Accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change has the potential to slightly increase the post-Fuel Handling Accident dose at the site boundary and in the control room. However, the existing analyses take no credit for containment of the release, so that the existing analysis results will remain bounding. 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 W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A–GO–18, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Russell Gibbs. E:\FR\FM\10APN1.SGM 10APN1 17950 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices cprice-sewell on PROD1PC66 with NOTICES FirstEnergy Nuclear Operating Company, et al., Docket No. 50–440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio Date of amendment request: January 19, 2007. Description of amendment request: The proposed amendment would modify Technical Specification (TS) 5.5.9, ‘‘Diesel Fuel Oil Testing Program,’’ by relocating a reference to a specific American Society for Testing and Materials (ASTM) international standard for fuel oil testing to licenseecontrolled documents, and by adding an alternate criteria to the ‘‘clear and bright’’ acceptance test for new fuel oil, per the consolidated line item improvement process (CLIIP). The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on February 22, 2006 (71 FR 9179), on possible amendments concerning the CLIIP, including a model safety evaluation and a model no significant hazards consideration determination. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 21, 2006 (71 FR 20735), as part of the CLIIP. In its application dated January 19, 2007, the licensee affirmed the applicability of the following determination. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of any accident previously evaluated? Response: No. The proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Requirements to perform testing in accordance with applicable ASTM standards are retained in the TS as are requirements to perform surveillances of both new and stored diesel fuel oil. Future changes to the licenseecontrolled document will be evaluated pursuant to the requirements of 10 CFR 50.59, ‘‘Changes, tests and experiments,’’ to ensure that such changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated. In addition, the ‘‘clear and bright’’ test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to recognize more rigorous testing of water and sediment content. Relocating the specific ASTM standard references from the TS to a licensee-controlled document and allowing a VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 water and sediment content test to be performed to establish the acceptability of new fuel oil will not affect nor degrade the ability of the emergency diesel generators (DGs) to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. Therefore, the changes do 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 changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. In addition, the ‘‘clear and bright’’ test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs. Therefore, the changes do 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 changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Instituting the proposed changes will continue to ensure the use of applicable ASTM standards to evaluate the quality of both new and stored fuel oil designated for use in the emergency DGs. Changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there PO 00000 Frm 00135 Fmt 4703 Sfmt 4703 is no significant reduction in a margin of safety. The ‘‘clear and bright’’ test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The margin of safety provided by the DGs is unaffected by the proposed changes since there continue to be TS requirements to ensure fuel oil is of the appropriate quality for emergency DG use. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies while maintaining sufficient controls to preserve the current margins of safety. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Russell A. Gibbs Florida Power Corporation, et al., Docket No. 50–302, Crystal River Unit 3 Nuclear Generating Plant (CR–3), Citrus County, Florida Date of amendment request: December 14, 2006, as supplemented by letter dated March 14, 2007. Description of amendment request: The proposed amendment would modify the technical specification (TS) requirements for inoperable snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The changes are consistent with NRC approved Industry/ Technical Specification Task Force (TSTF) standard TS change TSTF–372, Revision 4. The proposed amendment includes an administrative change to LCO 3.0.1 that will clarify that LCO 3.0.7 allows specified TS requirements to be suspended during physics tests performed in accordance with TSs 3.1.8 and 3.1.9. This administrative change will make the CR–3 TSs more consistent with the standard TSs and with TSTF– 372, Revision 4. The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed the applicability of the model NSHC determination in its application dated April 26, 2006. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices cprice-sewell on PROD1PC66 with NOTICES 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. Entrance into Actions or delaying entrance into Actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on the delay time allowed before declaring a TS supported system inoperable and taking its Conditions and Required Actions are no different than the consequences of an accident under the same plant conditions while relying on the existing TS supported system Conditions and Required Actions. Therefore, the consequences of an accident previously evaluated are not significantly increased by this change. Therefore, this 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 allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. Thus, this 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 allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change restores an allowance in the pre-ISTS [improved Standard Technical Specifications] conversion TS that was unintentionally eliminated by the conversion. The pre-ISTS TSs were considered to provide an adequate margin of safety for plant operation, as does the postISTS conversion TS. Therefore, this change does not involve a significant reduction in a margin of safety. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David T. Conley, Associate General Counsel II— Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Branch Chief: Thomas H. Boyce. VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 PSEG Nuclear LLC, Docket Nos. 50–272 and 50–311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of amendment request: March 16, 2007. Description of amendment request: The proposed amendment would add new Technical Specification (TS) requirements for the response times associated with a steam generator feedwater pump (SGFP) trip and feedwater isolation valve (FIV) closure. The amendment would also revise the TS requirements for the containment fan cooler unit (CFCU) cooling water flow rate. These changes are associated with a revised containment response analysis that credits a SGFP trip and FIV closure (on a feedwater regulator valve failure) to reduce the mass/energy release to the containment during a main steam line break (MSLB). The containment analysis also credits a reduced heat removal capability for the CFCUs, allowing a reduction in the required service water (SW) flow to the CFCUs. 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 does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change establishes response time requirements for feedwater isolation and reduced CFCU flow rates to support containment analyses to accommodate reduced CFCU heat removal capacity. The changes in analysis input assumptions affect plant response to an accident and are not accident initiators; therefore, they have no bearing on the probability of an accident. The Salem FSAR [Final Safety Analysis Report] Chapter 15 accidents which are impacted by a change in the CFCU modeling parameters are LOCA [loss-of-coolant accident] and MSLB mass and energy release Containment analyses. The consequences of these postulated accidents are shown to be acceptable using assumptions consistent with the proposed changes. For the LOCA transients, the containment cooling systems are considered for three aspects: core response, containment response and dose. The core response is most limiting when the containment conditions minimize back pressure since this increases the blowdown and reduces the effectiveness of the ECCS [emergency core cooling system]. The LOCA core response (10 CFR 50.46 [Section 50.46 of Title 10 of the Code of Federal Regulations]—PCT [peak cladding PO 00000 Frm 00136 Fmt 4703 Sfmt 4703 17951 temperature]) is conservatively biased to minimize the containment backpressure such that any safety injection effectiveness is minimized (the core becomes the highest resistance flow path). Thus, any reduction in the accident capability of the CFCUs has no bearing on the LOCA core response. The bounding containment integrity analyses are the LBLOCA [large-break LOCA] and the MSLB Inside Containment events. The containment integrity analysis relies on two heat removal paths to maintain containment pressure and temperature conditions. The CFCU air-to-water heat exchangers reject containment energy to the SW System and the Containment Spray System removes containment energy by using spray droplet direct contact heat exchange to transfer the energy from the containment ambient to the containment sump, where it is transferred out of containment via the RHR [residual heat removal] heat exchanger and CCW [component cooling water]/SW Systems. Containment integrity analyses for both LOCA and MSLB, using input assumptions consistent with the proposed changes, show that containment integrity is maintained with reduced CFCU heat removal capacity. The potential dose impacts due to reduced CFCU heat removal capacity are bounded as the design basis assumptions concerning the number of operating CFCUs (three of five), and the thermal-hydraulic transient operation of the Containment Spray System are unchanged. The Salem design basis only credits Containment Spray iodine removal effectiveness during the LOCA injection and recirculation phases based on a single failure of an entire ESF [engineered safety features] train. This assumption results in 3 of 5 CFCUs being available to ensure adequate mixing of the containment ambient air as well as operation of a single Containment Spray Train, which controls containment spray droplet size and pH, as described in UFSAR [updated FSAR] Section 6.2.3. As a further conservatism, the current LOCA Alternate Source Term (AST) analysis (Calculation S–C–ZZ–MDC–1945, an interim revision of which was sent to the NRC [Nuclear Regulatory Commission] staff for review via letter dated September 16, 2004) only credits two CFCUs for mixing. The Containment Building and Auxiliary Building leakage rates are unaffected by the revised containment analysis as the peak containment pressure and temperatures are less than the design basis values described in the Salem UFSAR. Therefore, there is no impact on offsite dose rates due to the reduced CFCU heat removal capacity. One other high energy line break for consideration is the rupture of a feedwater line break. From a containment response aspect, this event is bounded by the MSLB event, so it is not explicitly analyzed (or even discussed in the Salem UFSAR). A review of the Salem design basis for AST dose calculations shows that the revised Containment Integrity Analysis, WCAP– 16503, does not challenge any of the assumptions that are part of the AST design basis. Section 6.2 of the UFSAR indicates that the Appendix J Type A containment leak rate test E:\FR\FM\10APN1.SGM 10APN1 cprice-sewell on PROD1PC66 with NOTICES 17952 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices pressure is based on the containment design pressure of 47.0 psig, not the calculated accident pressure. Since the design pressure value bounds the peak pressure calculated in WCAP–16503 and is not being changed, the Appendix J testing requirements are not impacted. Thus, in conclusion, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change modifies response time requirements for feedwater isolation, and reduces CFCU flow rates and heat removal requirements consistent with the new containment analysis. 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 does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes support revised containment analysis to accommodate the reduced CFCU heat removal capacity. The response time-related changes impose new surveillance acceptance criteria to existing plant equipment that actuates to isolate feedwater following a safety injection signal. There is no change in actuation logic associated with the addition of response time criteria; therefore no new accident sequences would result from the imposition of response time test criteria to existing plant equipment. The reduction in minimum service water system flow to the CFCUs is supported by analyses demonstrating acceptable system performance and containment integrity following a demand for system operation. The post-accident conditions resulting from the proposed reduction in flow do not adversely impact the environmental qualification of equipment, such that no new consequential failures are introduced to any design basis accident scenario. CFCU operation with the proposed reduction in minimum required accident flow would not result in the progression of any design basis event into a previously unanalyzed accident. Therefore, no new accident scenarios are created from the CFCU flowrate reduction. 3. Does the proposed change involve a significant reduction in [a] margin of safety? Response: No. The proposed change does not involve a significant reduction in the margin of safety. The revised containment analyses accommodate reduced CFCU heat removal capacity using input assumptions consistent with the proposed changes. The proposed change involves the addition of feedwater isolation response time surveillance criteria and reduction in minimum service water system flows to CFCUs. These changes affect input to the analyses of mass/energy releases and containment response to a design basis main steam line break or loss of coolant accident. The analyses, consistent with the proposed changes, demonstrate that the acceptance criteria continue to be met, and the postaccident conditions do not adversely affect containment integrity or otherwise challenge VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 any safety limit. The margin of safety with respect to containment pressure is preserved by demonstrating that the calculated pressures do not exceed the design limit of 47 psig. 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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: Harold K. Chernoff. TXU Generation Company LP, Docket Nos. 50–445 and 50–446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas Date of amendment request: December 19, 2006. Brief description of amendments: The amendments requested would revise Technical Specifications (TS) requirement 3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ TS 3.8.1, ‘‘AC Sources—Operating,’’ TS 3.8.9, ‘‘Distribution Systems—Operating,’’ and TS Example 1.3–3. 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. D[o] the proposed change[s] involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes eliminate certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Time are no different than the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released PO 00000 Frm 00137 Fmt 4703 Sfmt 4703 offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes are consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change[s] d[o] not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. D[o] the proposed change[s] create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The changes do not alter any assumptions made in the safety analysis. Therefore, the proposed change[s] d[o] not create the possibility of a new or different accident from any accident previously evaluated. 3. D[o] the proposed change[s] involve a significant reduction in a margin of safety? Response: No. The proposed change to delete the second Completion Time does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside of the design basis. Therefore, the proposed change[s] d[o] 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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036. NRC Branch Chief: David Terao TXU Generation Company LP, Docket Nos. 50–445 and 50–446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas Date of amendment request: January 18, 2007. Brief description of amendments: The amendments requested would revise Technical Specifications (TS) requirement 3.8.1, ‘‘AC Sources— Operating,’’ Extension of Completion Times for Diesel Generators. 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: E:\FR\FM\10APN1.SGM 10APN1 cprice-sewell on PROD1PC66 with NOTICES Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed Technical Specification (TS) changes do not significantly increase the probability of occurrence of a previously evaluated accident because the Diesel Generators (DGs) are not initiators of previously evaluated accidents involving a loss of offsite power (LOOP). The proposed changes to the TS Required Actions and Completion Times (CT) do not affect any of the assumptions used in the deterministic or the Probabilistic Safety Assessment (PSA) analysis. Implementation of the proposed changes does not result in a risk significant impact. The onsite AC [alternating current] power sources will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by the increase in core damage frequency (CDF) is less than 1E–06 per year and the increase in large early release frequency (LERF) is less than 1E–07 per year. In addition, for the CT changes, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5E–07 and 5E–08, respectively. These changes meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, since the onsite AC power sources will continue to perform their functions with high reliability as originally assumed and the increase in risk as measured by DCDF, DLERF, ICCDP, and ICLERP risk metrics is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes are consistent with safety analysis assumptions and resultant consequences. The proposed TS changes will continue to ensure the DGs perform their function when called upon. Extending the TS CT to 14 days, when an AACPS [alternate AC power source] is available, does not affect the design, the operational characteristics, the function, or the reliability of the DGs. Additionally, the CT extension to 14 days does not affect the interfaces between the DGs and other plant systems. Conversely, in the absence of an AACPS, the DG 72-hour CT will be applied. The availability of the onsite AC power system to perform its accident mitigation function is not affected by the proposed VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 activity and thus there is no impact to the radiological consequences of any accident analysis. To fully evaluate the effect of the changes to the CT, PSA methods were utilized. The results of this analysis show no significant increase in the CDF and LERF. The Configuration Risk Management Program (CRMP) in TS 5.5.18 is an administrative program that assesses risk based on plant status. The risk-informed CT will be implemented consistent with the CRMP and approved plant procedures. When utilizing the 14-day extension, requirements of the CRMP per TS 5.5.18 call for the consideration of other measures to mitigate the consequences of an accident occurring while a DG is inoperable. Furthermore, administrative controls will be applied when exercising the 14-day CT extension and are adequate to maintain defense-in-depth and sufficient safety margins. 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 do not result in a change in the manner in which the electrical distribution subsystems provide plant protection. The changes to the CT do not change any existing accident scenarios, nor create any new or different accident scenarios. In addition, the changes do not impose any new or different accident mitigation requirements or eliminate any existing requirements. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does 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 proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are impacted by these changes. The proposed changes will not result in plant operation in a configuration outside the current design basis. The proposed activities only involve changes to certain TS CTs. Therefore the proposed change does not involve a reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036. PO 00000 Frm 00138 Fmt 4703 Sfmt 4703 17953 NRC Branch 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. Duke Power Company LLC, Docket Nos. 50–369 and 50–370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: March 8, 2007. Brief description of amendment request: The proposed amendments would revise the McGuire Nuclear Station, Units 1 and 2, Technical Specification 3.5.2.8, and the associated Bases and authorizes changes to the Updated Final Safety Analysis Reports concerning modifications to the emergency core cooling system sump. Date of publication of individual notice in Federal Register: March 19, 2007. Expiration date of individual notice: Comments April 18, 2007; Hearing May 18, 2007. 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, E:\FR\FM\10APN1.SGM 10APN1 17954 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices cprice-sewell on PROD1PC66 with NOTICES and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. Carolina Power & Light Company, et al., Docket No. 50–400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of application for amendment: May 23, 2006, as supplemented by letters dated October 3, 2006, and October 24, 2006. Brief description of amendment: This amendment revises Technical Specification by modifying the steam generator tube surveillance program at Shearon Harris Nuclear Power Plant, Unit 1. Date of issuance: March 16, 2007. Effective date: This amendment is effective as of the date of issuance and shall be implemented within 90 days of issuance. Amendment No. 124. Facility Operating License No. NPF– 63: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: December 19, 2006 (71 FR 75991). The supplemental letters VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 provided additional 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 the Safety Evaluation dated: March 16, 2007. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., Docket No. 50–423, Millstone Power Station, Unit No. 3 New London County, Connecticut Date of application for amendment: July 19, 2006. Brief description of amendment: The proposed amendment changed the Millstone Power Station, Unit No. 3 (MPS3) reactor core safety limits Technical Specification (TS) and relocated the reactor core safety limit figure to the Core Operating Limits Report in the MPS3 Technical Requirements Manual. Date of issuance: March 14, 2007 Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 236 Facility Operating License No. NPF– 49: The amendment revised the TSs. Date of initial notice in Federal Register: August 29, 2006 (71 FR 51227). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 14, 2007. No significant hazards consideration comments received: No. Duke Power Company LLC, et al., Docket Nos. 50–413 and 50–414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of application for amendments: April 11, 2006. Brief description of amendments: (TSTF–372, Rev. 4) The amendments added Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed with an approved Bases Control Program that is consistent with the TS Bases Control Program described in Section 5.5 of the applicable vendor’s Standard Technical Specifications. The amendment also made an administrative change, renumbering existing LCO 3.0.8 to LCO 3.0.9. Date of issuance: March 19, 2007 Effective date: As of the date of issuance and shall be implemented PO 00000 Frm 00139 Fmt 4703 Sfmt 4703 within 120 days from the date of issuance. Amendment Nos.: 235, 231 Renewed Facility Operating License Nos. NPF–35 and NPF–52: Amendments revised the licenses and the technical specifications. Date of initial notice in Federal Register: December 5, 2006 (71 FR 70555). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007. No significant hazards consideration comments received: No Duke Power Company LLC, Docket Nos. 50–369 and 50–370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of application for amendments: April 11, 2006. Brief description of amendments: (TSTF–372, Rev. 4) The amendments added Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed with an approved Bases Control Program that is consistent with the TS Bases Control Program described in Section 5.5 of the applicable vendor’s Standard Technical Specifications. The amendment also made an administrative change, renumbering existing LCO 3.0.8 to LCO 3.0.9. Date of issuance: March 29, 2007. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: 238, 220. Renewed Facility Operating License Nos. NPF–9 and NPF–17: Amendments revised the licenses and the technical specifications. Date of initial notice in Federal Register: December 5, 2006 (71 FR 70556). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 29, 2007. No significant hazards consideration comments received: No. Energy Northwest, Docket No. 50–397, Columbia Generating Station, Benton County, Washington Date of application for amendment: May 22, 2006, as supplemented by letter dated February 5, 2007. Brief description of amendment: The amendment revised Technical Specification Surveillance Requirements 3.8.1.11, 3.8.1.12, 3.8.1.16, and 3.8.1.19 to eliminate the specific test-performance mode E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices restrictions for the High-Pressure Core Spray Division 3 diesel generator. Date of issuance: March 23, 2007. Effective date: As of the date of issuance and shall be implemented within 45 days from the date of issuance. Amendment No.: 203. Facility Operating License No. NPF– 21: The amendment revised the Technical Specifications and Facility Operating License. Date of initial notice in Federal Register: July 18, 2006 (71 FR 40745). The supplemental letter dated February 5, 2007, 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 amendment is contained in a Safety Evaluation dated March 23, 2007. No significant hazards consideration comments received: No. Entergy Nuclear Operations, Inc., Docket No. 50–293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts. cprice-sewell on PROD1PC66 with NOTICES Date of application for amendment: December 27, 2006. Brief description of amendment: The amendment revised the Technical Specification Limiting Condition for Operation 3.14.A to adopt the Technical Specification Task Force 484, Revision 0, ‘‘Use of Technical Specification 3.10.1 for Scram Time Testing Activities.’’ Date of issuance: March 26, 2007. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 226. Facility Operating License No. DPR– 35: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 20, 2007 (72 FR 7776). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2007. No significant hazards consideration comments received: No Entergy Nuclear Operations, Inc., Docket No. 50–293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts. Date of application for amendment: January 15, 2007. Brief description of amendment: The amendment revised the Technical Specifications (TS) to extend the use of VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 17955 the current pressure-temperature limits as specified in TS Figures 3.6.1, 3.6.2, and 3.6.3 through the end of operating cycle 18. Date of issuance: March 26, 2007. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 227. Facility Operating License No. DPR– 35: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 12, 2007 (72 FR 6609). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2007. No significant hazards consideration comments received: No. Assemblies,’’ to allow the use of hafnium as an additional type of control material. Date of issuance: March 16, 2007. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment No: 174. Facility Operating License No. NPF– 29: The amendment revises the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: February 13, 2007 (72 FR 6782). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2007. 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 Entergy Operations, Inc., Docket No. 50– 382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of application for amendment: April 22, 2006. Brief description of amendment: The amendment revised Technical Specification (TS) requirements for inoperable snubbers by relocating the current TS requirements Limiting Condition for Operation (LCO) 3.6.I and Surveillance Requirement (SR) 4.6.I to the Technical Requirements Manual and adding LCO 3.0.8 to the TSs. The associated TS Bases section has also been relocated. Date of Issuance: March 26, 2007. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 230. Facility Operating License No. DPR– 28: The amendment revised the License and TSs. Date of initial notice in Federal Register: June 6, 2006 (71 FR 32604). The Commission’s related evaluation of this amendment is contained in a Safety Evaluation dated March 26, 2007. 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: January 18, 2007. Brief description of amendment: The amendment revised the description of the control rod assemblies in Grand Gulf Nuclear Station, Unit 1, Technical Specification 4.2.2, ‘‘Control Rod PO 00000 Frm 00140 Fmt 4703 Sfmt 4703 Date of amendment request: September 26, 2006. Brief description of amendment: The amendment deleted reference to the containment fan cooler condensate flow switch from Technical Specification 3.4.5.1, ‘‘Reactor Coolant System Leakage—Leakage Detection Instrumentation,’’ and modified or deleted associated actions. The Nuclear Regulatory Commission staff had determined that the remaining leak detection methods provided adequate means for detecting, and to the extent practical, identifying the location of the source of potential reactor coolant leakage. Date of issuance: March 19, 2007. Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance. Amendment No.: 212. Facility Operating License No. NPF– 38: The amendment revised the Operating License and the Technical Specifications. Date of initial notice in Federal Register: February 13, 2007 (72 FR 6782). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 19, 2007. No significant hazards consideration comments received: No. 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 application for amendments: May 26, 2006, as supplemented on December 26, 2006, and March 14, 2007. Brief description of amendments: The amendments revised the existing steam E:\FR\FM\10APN1.SGM 10APN1 17956 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices generator (SG) tube surveillance program. The changes are modeled after Technical Specifications Task Force (TSTF) traveler TSTF–449, Revision 4, ‘‘Steam Generator Tube Integrity,’’ and the model safety evaluation prepared by the Nuclear Regulatory Commission staff and published in the Federal Register on March 2, 2005 (70 FR 10298). In this regard, the scope of the amendments includes changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the SG tube surveillance program (SG tube integrity), and changes to the SG reporting requirements. Date of issuance: March 14, 2007. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 298 and 279. Facility Operating License Nos. DPR– 58 and DPR–74: Amendments revise the Technical Specifications. Date of initial notice in Federal Register: July 5, 2006 (71 FR 38183). The 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 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 March 14, 2007. No significant hazards consideration comments received: No. cprice-sewell on PROD1PC66 with NOTICES Nuclear Management Company, LLC, Docket No. 50–263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of application for amendment: March 7, 2006, as supplemented by letters dated May 30, September 7, December 15, 2006, and January 2, 2007. Brief description of amendment: The amendment revised Section 4.3, ‘‘Fuel Storage,’’ of the Monticello Nuclear Generating Plant, technical specifications to allow for installation of an additional temporary 8x8 (64-cell) high-density spent fuel storage rack in the spent fuel pool to maintain full core off-load capability. Date of issuance: March 9, 2007. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 150. Facility Operating License No. DPR– 22. Amendment revised the Technical Specifications. Date of initial notice in Federal Register: April 3, 2006 (71 FR 16599). The supplemental letters contained clarifying information and did not VerDate Aug<31>2005 16:29 Apr 09, 2007 Jkt 211001 change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2007. 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 application for amendments: March 23, 2006, as supplemented on December 19, 2006. Brief description of amendments: The amendments revise Technical Specification (TS) 3.3.4, ‘‘Loss of Power (LOP) Diesel Generator (DG) Start and Load Sequence Instrumentation,’’ and surveillance requirement 3.3.4.3.b to modify the TS title and correct nonconservatisms in the allowable values for the degraded voltage time delay. Date of issuance: March 21, 2007. Effective date: As of the date of issuance and shall be implemented within 45 days. Amendment Nos.: 225 & 231. Renewed Facility Operating License Nos. DPR–24 and DPR–27: Amendments revised the Technical Specifications and License. Date of initial notice in Federal Register: April 25, 2006 (71 FR 23958). The December 19, 2006, supplement, contained clarifying information and did not change the staff’s initial proposed finding of no significant hazards consideration. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 21, 2007. No significant hazards consideration comments received: No. Nuclear Management Company, LLC, Docket Nos. 50–282 and 50–306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of application for amendments: February 16, 2006, supplemented by letters dated July 21, and December 27, 2006. Brief description of amendments: The amendments consist of changes to the Technical Specifications (TSs) related to steam generator tube integrity. The amendments are modeled after the U.S. Nuclear Regulatory Commission approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF– 449, ‘‘Steam Generator Tube Integrity,’’ Revision 4 (ML0510902003). PO 00000 Frm 00141 Fmt 4703 Sfmt 4703 Date of issuance: March 20, 2007. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment Nos.: 177 and 167. Facility Operating License Nos. DPR– 42 and DPR–60: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: April 11, 2006 (71 FR 18376) The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 20, 2007. No significant hazards consideration comments received: No. 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 application for amendments: May 30, 2006, as supplemented by letters dated November 22, 2006, and January 11, 2007. Brief description of amendments: The amendments revised the existing steam generator (SG) tube surveillance program. The changes were modeled after Technical Specification Task Force (TSTF) traveler TSTF–449, Revision 4, ‘‘Steam Generator Tube Integrity,’’ and the model safety evaluation prepared by the U.S. Nuclear Regulatory Commission and published in the Federal Register on March 2, 2005 (70 FR 10298). The scope of the application included changes to the definition of leakage, changes to the primary-tosecondary leakage requirements, changes to the SG tube surveillance program (SG tube integrity), changes to the SG reporting requirements, and associated changes to the Technical Specification Bases. Date of issuance: March 21, 2007. Effective date: As of its date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: Unit 1—194; Unit 2—195. Facility Operating License Nos. DPR– 80 and DPR–82: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: July 18, 2006 (71 FR 40751). The supplemental letters dated November 22, 2006, and January 11, 2007, provided additional information that clarified the application, did not expand the scope of the application as E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices 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 amendments is contained in a Safety Evaluation dated March 21, 2007. No significant hazards consideration comments received: No. 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 application for amendments: December 14, 2006. Brief description of amendments: The amendments deleted Section 2.G of Facility Operating License Nos. DPR–80 and DPR–82, which require reporting of violations of the requirements of Sections 2.C, 2.E, and 2.F of the operating license. This operating license improvement was made available by the U.S. Nuclear Regulatory Commission on November 4, 2005, as part of the consolidated line item improvement process (CLIIP). Date of issuance: March 19, 2007. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment Nos.: Unit 1–193; Unit 2–194. Facility Operating License Nos. DPR– 80 and DPR–82: The amendments revised the Facility Operating Licenses. Date of initial notice in Federal Register: January 3, 2007 (72 FR 154). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007. No significant hazards consideration comments received: No. cprice-sewell on PROD1PC66 with NOTICES PPL Susquehanna, LLC, Docket No. 50– 387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania Date of application for amendments: November 18, 2005, as supplemented on November 29, 2006, December 1, 2006, December 15, 2006, January 9, 2007, and March 12, 2007 (PLA–6168 and PLA– 6169). Brief description of amendments: The amendments change the SSES 1 and 2 Technical Specifications (TSs) to implement the Average Power Range Monitor/Rod Block Monitor/TSs/ Maximum Load Line Limit Analysis by revising TS 1.1, ‘‘Definitions,’’ TS 5.6.5, ‘‘Core Operating Limits Report,’’ and the surveillance requirement sections of TS 3.3.1.1, ‘‘Reactor Protection System Instrumentation,’’ and TS 3.3.2.1, VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 ‘‘Control Rod Block Instrumentation.’’ The amendments also delete TS 3.2.4, ‘‘Average Power Range Monitor Gain and Setpoints,’’ and its associated references in the TSs. Additionally, the amendments change the method of evaluation for the postulated recirculation line break in the reactor pressure vessel shield annulus region. Date of issuance: March 23, 2007. Effective date: As of the date of issuance and to be implemented prior to the startup following the SSES 1 spring 2008 15th refueling outage for Unit 1 and prior to the startup following the SSES 2 spring 2007 13th refueling outage for Unit 2. Amendment Nos.: 242 and 220. Facility Operating License Nos. NPF– 14 and NPF–22: The amendments revised the TSs and the License. Date of initial notice in Federal Register: February 14, 2006 (71 FR 7810). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 23, 2007. No significant hazards consideration comments received: No. PPL Susquehanna, LLC, Docket No. 50– 387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania Date of application for amendments: September 7, 2006. Brief description of amendments: The amendments revise the SSES 1 and 2 Technical Specifications (TSs) Section 5.5.6, ‘‘Inservice Testing Program,’’ and TS 5.5.12, ‘‘Primary Containment Leakage Rate Testing Program,’’ to be consistent with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(f)(4) and 10 CFR 50.55a(g)(4), respectively. The amendments implement TS Task Force (TSTF)–343, Revision 1 and TSTF–479, Revision 0. Date of issuance: March 19, 2007. Effective date: As of the date of issuance and to be implemented within 30 days. Amendment Nos.: 241 and 219. Facility Operating License Nos. NPF– 14 and NPF–22: The amendments revised the License and Technical Specifications. Date of initial notice in Federal Register: December 19, 2006 (71 FR 75997). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007. No significant hazards consideration comments received: No. PO 00000 Frm 00142 Fmt 4703 Sfmt 4703 17957 PPL Susquehanna, LLC, Docket No. 50– 388, Susquehanna Steam Electric Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania Date of application for amendment: November 16, 2006, as supplemented on February 15, 2007. Brief description of amendment: The amendment changes the SSES 2 Technical Specification (TS) Section 2.1.1.2 by revising the Unit 2 Cycle 14 Minimum Critical Power Ratio Safety Limit for two-loop and single-loop operation and the references listed in TS 5.6.5.b. Date of issuance: March 19, 2007. Effective date: As of the date of issuance and to be implemented within 30 days. Amendment No.: 218. Facility Operating License No. NPF– 22: The amendment revised the License and Technical Specifications. Date of initial notice in Federal Register: December 19, 2006 (71 FR 75998). The supplement dated February 15, 2007, 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 March 19, 2007. No significant hazards consideration comments received: No. PSEG Nuclear LLC, Docket Nos. 50–272 and 50–311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of application for amendments: May 1, 2006. Brief description of amendments: The amendments eliminate the requirement for a power range neutron flux high negative rate trip and delete the references to this trip in Salem Unit Nos. 1 and 2 Technical Specification (TS) Table 2.2–1, ‘‘Reactor Trip System Instrumentation Trip Setpoints,’’ TS Table 3.3–1, ‘‘Reactor Trip System Instrumentation,’’ TS Table 3.3–2, ‘‘Reactor Trip System Instrumentation Response Times,’’ and TS Table 4.3–1, ‘‘Reactor Trip System Instrumentation Surveillance Requirements.’’ The amendments also incorporate administrative and editorial changes to correct miscellaneous errors in the TSs for Salem Units Nos. 1 and 2. Date of issuance: March 19, 2007. Effective date: As of the date of issuance, to be implemented within 60 days. E:\FR\FM\10APN1.SGM 10APN1 17958 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices Amendment Nos.: 278 and 261 Facility Operating License Nos. DPR– 70 and DPR–75: The amendments revised the TSs and the License. Date of initial notice in Federal Register: July 18, 2006 (71 FR 40752). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007. No significant hazards consideration comments received: No. PSEG Nuclear LLC, Docket Nos. 50–272 and 50–311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of application for amendments: August 4, 2006, as supplemented by letter dated February 20, 2007. Brief description of amendments: The amendments allow the use of blind flanges for containment isolation in the containment purge system supply and exhaust lines, and make corresponding changes to the Technical Specifications (TSs). The amendments also consolidate the containment isolation requirements by moving the requirements of TS 3/4 6.1.7, ‘‘Containment Ventilation System,’’ to TS 3/4 6.3.1 (TS 3/4 6.3 for Unit No. 2), ‘‘Containment Isolation Valves.’’ Date of issuance: March 19, 2007. Effective date: As of the date of issuance, to be implemented within 60 days. Amendment Nos.: 277 and 260. Facility Operating License Nos. DPR– 70 and DPR–75: The amendments revised the License and the TSs. cprice-sewell on PROD1PC66 with NOTICES PSEG Nuclear LLC, Docket No. 50–272, Salem Nuclear Generating Station, Unit No. 1, Salem County, New Jersey Date of application for amendment: January 18, 2007, as supplemented on February 23, March 9, and March 22, 2007. Brief description of amendment: The amendment approves a one-time change to the Technical Specifications (TSs) regarding the steam generator (SG) tube inspection and repair required for the portion of the SG tubes passing through the tubesheet region. Specifically, for Salem Unit No. 1 refueling outage 18 (planned for spring 2007) and the subsequent operating cycle, the TS changes limit the required inspection (and repair if degradation is found) to the portions of the SG tubes passing through the upper 17 inches of the approximate 21-inch tubesheet region. Date of issuance: March 27, 2007. Effective date: As of the date of issuance, to be implemented within 60 days. Amendment No.: 279. VerDate Aug<31>2005 15:22 Apr 09, 2007 Jkt 211001 Facility Operating License No. DPR– 70: The amendment revised the TSs and the License. Date of initial notice in Federal Register: January 25, 2007 (72 FR 3427). The letters dated February 23, March 9, and March 22, 2007, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 27, 2007. 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: March 28, 2006, as supplemented by letter dated October 24, 2006. Brief description of amendment: The amendment revises Technical Specification Surveillance Requirement 3.5.1.4 to change the method and frequency for verifying emergency core cooling system accumulator boric acid concentration. Date of issuance: March 28, 2007. Effective date: As of the date of issuance to be implemented within 45 days. Amendment No.: 101. Renewed Facility Operating License No. DPR–18: Amendment revised the License and Technical Specifications. Date of initial notice in Federal Register: April 25, 2006 (71 FR 23960) The October 24, 2006, letter 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 March 28, 2007. 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: August 22, 2005, as supplemented by letters dated September 18, 2006, October 23, 2006, and February 16, 2007. Brief description of amendments: These amendments modified Technical Specification (TS) requirements related to control room envelope habitability in PO 00000 Frm 00143 Fmt 4703 Sfmt 4703 TS 3.7.10, ‘‘Control Room Emergency Filtration/Pressurization System (CREFS)’’ and TS Section 5.5, ‘‘Administrative Controls—Programs and Manuals.’’ Date of issuance: March 26, 2007. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: 136/136. Facility Operating License Nos. NPF– 87 and NPF–89: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: November 8, 2005 (70 FR 67754). The supplemental letters dated September 18 and October 23, 2006, and February 16, 2007, 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 amendments is contained in a Safety Evaluation dated March 26, 2007. 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 31, 2006. Brief description of amendments: The amendments revised Technical Specification 5.0 entitled, ‘‘ADMINISTRATIVE CONTROLS.’’ Specifically, the change deleted the Vice President, Nuclear Operations, as an alternative to the Plant Manager for certain functions. Date of Issuance: March 20, 2007. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: Unit 1–134; Unit 2–134. Facility Operating License Nos. NPF– 87 and NPF–89: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: September 12, 2006 (71 FR 53722). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated March 20, 2007. No significant hazards consideration comments received: No. E:\FR\FM\10APN1.SGM 10APN1 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices TXU Generation Company LP, Docket Nos. 50–445 and 50–446, Comanche Peak Steam Electric Station (CPSES), Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: February 21, 2006, as supplemented by letter dated March 19, 2007. Brief description of amendments: The amendments revise TS 5.6.5 entitled, ‘‘Core Operating Limits Report (COLR),’’ by adding two reports providing Lossof-Coolant Accident (LOCA) and nonLOCA analysis methodologies for CPSES Unit 1. Date of issuance: March 26, 2007. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance, but no later than the entry into Mode 5 in the restart of Unit 1 from its spring 2007 refueling outage. Amendment Nos.: 135/135. Facility Operating License Nos. NPF– 87 and NPF–89: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: June 6, 2006 (71 FR 32609). The supplemental letter dated March 19, 2007, 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 amendments is contained in a Safety Evaluation dated March 26, 2007. No significant hazards consideration comments received: No. cprice-sewell on PROD1PC66 with NOTICES Union Electric Company, Docket No. 50–483, Callaway Plant, Unit 1, Callaway County, Missouri Date of application for amendment: May 25, 2006, as supplemented by letter dated March 12, 2007. Brief description of amendment: The amendment revised Technical Specifications 3.1.7, ‘‘Rod Position Indication,’’ 3.2.1, ‘‘Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology),’’ 3.2.4, ‘‘Quadrant Power Tilt Ratio (QPTR),’’ and 3.3.1, ‘‘Reactor Trip System (RTS) Instrumentation,’’ to allow use of the Westinghouse proprietary computer code, the Best Estimate Analyzer for Core Operations— Nuclear (BEACON). Certain required actions, for when a limiting condition for operation is not met, and certain surveillance requirements are being changed to refer to power distribution measurements or measurement information of the core. VerDate Aug<31>2005 16:29 Apr 09, 2007 Jkt 211001 Date of issuance: March 21, 2007. Effective date: As of its date of issuance and shall be implemented before entry into Mode 2 in the plant restart from the refueling outage scheduled for the spring of 2007. This includes the incorporation of the identified changes to the Final Safety Analysis Report (FSAR) in Attachment 6 of the licensee’s application dated May 25, 2006, into the FSAR. Amendment No.: 182. Facility Operating License No. NPF– 30: The amendment revised the Operating License and Technical Specifications. Date of initial notice in Federal Register: July 18, 2006 (71 FR 40756) The supplemental letter dated March 12, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff’s original proposed no significant hazards consideration determination published in the Federal Register on July 18, 2006. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated March 21, 2007. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 3rd day of April 2007. For the Nuclear Regulatory Commission. Catherine Haney, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. E7–6632 Filed 4–9–07; 8:45 am] BILLING CODE 7590–01–P OVERSEAS PRIVATE INVESTMENT CORPORATION Submission for OMB Review; Comment Request Overseas Private Investment Corporation (OPIC) ACTION: Request for comments. AGENCY: SUMMARY: Under the provision of the Paperwork Reduction Act (44 U.S.C. Chapter 35), agencies are required to publish a Notice in the Federal Register notifying the public that Agency is preparing an information collection request for OMB review and approval and to request public review and comment on the submission. Comments are being solicited on the need for the information, its practical utility, the accuracy of the Agency’s burden estimate, and on ways to minimize the reporting burden, including automated collection techniques and uses of other forms of PO 00000 Frm 00144 Fmt 4703 Sfmt 4703 17959 technology. The proposed form under review is summarized below. DATES: Comments must be received within 30 calendar days of this notice. ADDRESSES: Copies of the subject form and the request for review prepared for submission to OMB may be obtained from the Agency submitting officer. Comments on the form should be submitted to the Agency Submitting Officer. FOR FURTHER INFORMATION CONTACT: OPIC Agency Submitting Officer: Essie Bryant, Record Manager, Overseas Private Investment Corporation, 1100 New York Avenue, NW., Washington, DC 20527; 202–336–8563. Summary Form Under Review Type of Request: Revised form. Title: OPIC Self-Monitoring Questionnaire. Form Number: OPIC–162. Frequency of Use: Annually for duration of project. Type of Respondents: Business or other institution (except farms); individuals. Standard Industrial Classification Codes: All. Description of Affected Public: U.S. companies or citizens investing overseas. Reporting Hours: 6.5 hours per project. Number of Responses: 350 per year. Federal Cost: $35,000. Authority for Information Collection: Sections 231, 234(a), 239(d), and 240A of the Foreign Assistance Act of 1961, as amended. Abstract (Needs and Uses): The questionnaire is completed by OPICassisted investors annually. The questionnaire allows OPIC’s assessment of effects of OPIC-assisted projects on the U.S. economy and employment, as well as on the environment and economic development abroad. Dated: April 5, 2007. John P. Crowley, III, Senior Administrative Counsel, Department of Legal Affairs. [FR Doc. 07–1771 Filed 4–9–07; 8:45 am] BILLING CODE 3210–01–M PENSION BENEFIT GUARANTY CORPORATION Approval of Exemption From the Bond/ Escrow Requirement Relating to the Sale of Assets by an Employer Who Contributes to a Multiemployer Plan; Washington Nationals Baseball Club, LLC Pension Benefit Guaranty Corporation. AGENCY: E:\FR\FM\10APN1.SGM 10APN1

Agencies

[Federal Register Volume 72, Number 68 (Tuesday, April 10, 2007)]
[Notices]
[Pages 17944-17959]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-6632]


-----------------------------------------------------------------------

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 March 16, 2007 to March 29, 2007. The last 
biweekly notice was published on March 27, 2007 (72 FR 14303).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of

[[Page 17945]]

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 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: November 13, 2006.
    Description of amendment request: The proposed amendment changes 
the technical specification (TS) testing frequency for the surveillance 
requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The proposed 
change revises the test frequency of SR 3.1.4.2, control rod scram time 
testing, from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in Mode 1.''
    AmerGen has reviewed the proposed no significant hazards 
consideration determination published in the Federal Register on August 
23, 2004 (69 FR 51864), as part of the consolidated line item 
improvement process (CLIIP) and has concluded that the proposed 
determination presented in the notice is applicable to Clinton Power 
Station, Unit No. 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below.

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    Based on the above, the proposed change presents no significant 
hazards consideration under the standards set forth in 10 CFR 50.92(c), 
and accordingly, a finding of ``no significant hazards consideration'' 
is justified.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Russell Gibbs.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: January 26, 2007.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) requirements for unavailable 
barriers by

[[Page 17946]]

adding limiting condition for operation (LCO) 3.0.9. This would 
establish conditions under which TS systems would remain operable when 
required physical barriers are not capable of providing their related 
support function. Also, the proposed amendment would make editorial 
changes to LCO 3.0.8 to be consistent with the terminology in LCO 
3.0.9.
    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 by a reference to a generic analysis published in the 
Federal Register on October 3, 2006 (71 FR 58444), which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. 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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG 1.177. A 
bounding risk assessment was performed to justify the proposed TS 
changes. This application of LCO 3.0.9 is predicated upon the 
licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant 
as indicated by the anticipated low levels of associated risk (ICCDP 
and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety 
Evaluation.
    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 G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: L. Raghavan.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: January 18, 2007.
    Description of amendment request: The proposed amendment would 
revise the expiration limit for the reactor coolant system Pressure/
Temperature (P/T) limit graphs in Technical Specifications (TS); revise 
the adjusted reference temperature for the reactor vessel; and revise 
the Low Temperature Overpressure Protection (LTOP) arming temperature 
value specified in TSs. It would also make editorial changes in the use 
of inequality signs in TSs associated with the LTOP arming temperature 
in order to make them consistent.
    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 does not affect the accident initiators or 
mitigation assumptions associated with any of the accidents 
previously evaluated. Operating restrictions on pressure-temperature 
conditions for the reactor pressure vessel provide assurance that 
reactor vessel integrity will be maintained under accident or 
transient conditions. The proposed change uses approved criteria and 
analysis methods to update the time period for which the current 
operating limits remain valid.
    The LTOP system performs an automatic function by opening relief 
valves if reactor coolant system pressure reaches a temperature-
dependent limit. The proposed change includes establishing a more 
restrictive temperature limit for when this system must be in 
service, to reflect the material condition of the reactor vessel at 
the new EFPY limit proposed for the pressure-temperature graphs. The 
mitigation function and capability of the LTOP system is not being 
changed by this request.
    Therefore, the proposed changes do 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 new accident initiators being introduced by this 
proposed change. The proposed change does not involve installation 
of new plant equipment, modification of existing equipment, or 
changes in the way that plant equipment is operated. Pressure-
temperature operating limits depicted by graphs in the technical 
specifications will not be changed and will continue to be used by 
plant operators. A change in the LTOP system arming temperature will 
assure that the graphs remain valid for the proposed new operating 
period of 27.2 EFPY [effective full power years].
    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
    Operating limits on pressure and temperature conditions for the 
reactor coolant system (RCS) are important to assure that the RCS 
pressure boundary stresses are within analyzed limits. Margins of 
safety are inherent in the analysis methods, assumptions, and limits 
specified in regulations and guidance documents. The proposed change 
is based on NRC-accepted methods, assumptions and limits and 
maintains the required margin of 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

[[Page 17947]]

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 Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), Westchester 
County, New York

    Date of amendment request: March 13, 2007.
    Description of amendment request: The amendment would revise 
License Condition 2.K for IP2 and License Condition 2.H for IP3, which 
require the implementation and maintenance of an approved Fire 
Protection Program for each unit.
    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 changes are strictly an administrative relocation 
of the specific fire protection SER [safety evaluation report] 
references and do not modify any requirements of the fire protection 
programs.
    Therefore, the proposed changes do 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 changes are strictly an administrative relocation 
of the specific fire protection SER references and do not modify any 
requirements of the fire protection programs.
    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 proposed changes are strictly an administrative relocation 
of the specific fire protection SER references and do not modify any 
requirements of the fire protection programs.
    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 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 Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: March 22, 2007.
    Description of amendment request: The proposed amendment will 
revise the test acceptance criteria specified in Technical 
Specification Surveillance Requirement (SR) 3.8.1.10 for the diesel 
generator endurance test.
    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. The proposed change revises the acceptance criteria to be 
applied to an existing surveillance test of the facility emergency 
diesel generators (DGs). Performing a surveillance test is not an 
accident initiator and does not increase the probability of an 
accident occurring. The proposed new acceptance criteria will assure 
that the DGs are capable of carrying the peak electrical loading 
assumed in the various existing safety analyses which take credit 
for the operation of the DGs. Establishing acceptance criteria that 
bound existing analyses validates the related assumption used in 
those analyses regarding the capability of equipment to mitigate 
accident conditions. Therefore the proposed change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed change revises the test acceptance criteria for 
a specific performance test conducted on the existing DGs. The 
proposed change does not involve installation of new equipment or 
modification of existing equipment, so no new equipment failure 
modes are introduced. The proposed revision to the DG surveillance 
test acceptance criteria also is not a change to the way that the 
equipment or facility is operated and no new accident initiators are 
created. 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?
    No. The conduct of performance tests on safety-related plant 
equipment is a means of assuring that the equipment is capable of 
maintaining the margin of safety established in the safety analyses 
for the facility. The proposed change in the DG technical 
specification surveillance test acceptance criteria is consistent 
with values assumed in existing safety analyses and is consistent 
with the design rating of the DGs. 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 Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: February 15, 2007.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) 3.10.1 to expand its scope to include 
provisions for reactor coolant temperature excursions greater than 212 
[deg]F as a consequence of inservice leak and hydrostatic testing, and 
as a consequence of scram time testing initiated in conjunction with an 
inservice leak or hydrostatic test, while considering operational 
conditions to be in Mode 4, which is defined to be reactor coolant 
temperature less than or equal to 212 [deg]F.
    This change was proposed by the industry's TS Task Force (TSTF) and 
is designated TSTF-484. The NRC staff issued a notice of opportunity 
for comment in the Federal Register on August 21, 2006 (71 FR 48561), 
on possible amendments concerning TSTF-484, including a model safety 
evaluation and model no significant hazards (NSHC) determination, using 
the consolidated line item improvement process (CLIIP). The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 27,

[[Page 17948]]

2006 (71 FR 63050). The licensee affirmed the applicability of the 
following NSHC determination in its application dated February 15, 
2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1: The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or 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.
    Criterion 2: The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3: The proposed change does not involve a significant 
reduction in a margin of safety.
    Technical Specifications currently allow for operation at 
greater than 212 [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    Based on the above, the NRC staff concludes that the proposed 
change presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 
no significant hazards consideration is justified.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Douglas V. Pickett.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-373 and 50-374, 
LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: November 17, 2006.
    Description of amendment request: The proposed amendments would 
replace references to Section XI of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (Code) with a 
reference to the ASME Code of Operation and Maintenance of Nuclear 
Power Plants (OM Code) in Technical Specification (TS) 5.5.7, 
``Inservice Testing Program [IST].'' These proposed changes are 
consistent with the implementation of the LSCS, Units 1 and 2 third 10-
year IST program in accordance with the requirements of Title 10 of the 
Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes and 
standards,'' paragraph (f), ``Inservice testing requirements.'' The 
third 10-year interval for LSCS, Units 1 and 2 is scheduled to start on 
October 12, 2007.
    In addition to the replacement of the references, EGC is also 
adding provisions in TS 5.5.7, item b, to only apply Surveillance 
Requirement (SR) 3.0.2 to those inservice testing frequencies of two 
years or less. These proposed changes are based on TS Task Force (TSTF) 
Traveler No. 479-A (TSTF-479-A), Revision 0, ``Changes to Reflect 
Revision of 10 CFR 50.55a,'' as modified by TSTF-497, Revision 0, 
``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies 
of 2 Years or Less'' and approved by the NRC in December 6, 2005, and 
October 4, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a, ``Codes and 
standards,'' paragraph (f) regarding the inservice testing of pumps 
and calves for the Third 10-year Interval. The current TS reference 
the [American Society of Mechanical Engineers] ASME Boiler and 
Pressure Vessel Code, Section XI, requirements for the inservice 
testing of ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed changes would reference the ASME OM Code, which is 
consistent with 10 CFR 50.55a, paragraph (f), ``Inservice testing 
requirements,'' and approved for use by the NRC. In addition, 
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only 
applied to those inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes are administrative in nature, do not affect 
any accident initiators, do not affect the ability of LSCS to 
successfully respond to previously evaluated accidents and do not 
affect radiological assumptions used in the evaluations. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a(f) regarding the 
inservice testing of pumps and valves for the Third 10-year 
Interval. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes 
would reference the ASME OM Code, which is consistent with the 10 
CFR 50.55a(f) and approved for use by the NRC. In addition, 
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only 
applied to those inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes to TS Section 5.5.7 do not affect the 
performance of any LSCS structure, system, or component credited 
with mitigating any accident previously evaluated and do not 
introduce any new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to 
conform to the requirements of 10 CFR 50.55a(f) regarding the 
inservice testing of pumps and valves for the Third 10-Year 
Interval. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2,

[[Page 17949]]

and 3 pumps and valves. The proposed changes would reference the 
ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and 
approved for use by the NRC. In addition, provisions modifying TS 
5.5.7, item b, clarify that SR 3.0.2 is only applied to those 
inservice testing frequencies of two years or less. The definitions 
of the frequencies are not changed by this license amendment 
request.
    The proposed changes do not modify the safety limits setpoints 
at which protective actions are initiated and do not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of 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 three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Russell Gibbs.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: June 2, 2006.
    Description of amendment request: The proposed amendments 
incorporates revised 10 CFR Part 20 requirements for Limerick 
Generating Station Units 1 and 2 technical specifications (TSs).
    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.
    Updating the Technical Specifications (TS) to be consistent with 
10 CFR Part 20 has no impact on plant structures, systems, or 
components, does not affect any accident initiators, and does not 
change any safety analysis. Therefore, the proposed changes do not 
involve an 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.
    Updating the TS to be consistent with 10 CFR Part 20 will not 
change any equipment, require new equipment to be installed, or 
change the way current equipment operates. No credible new failure 
mechanisms, malfunctions, or accident initiators are created by the 
proposed changes.
    Therefore, the changes do 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.
    Updating the TS to be consistent with 10 CFR Part 20 does not 
adversely affect existing plant safety margins or the reliability of 
equipment assumed to operate in the safety analysis. As such, there 
are no changes being made to safety analysis assumptions, safety 
limits or limiting safety system settings that would adversely 
affect plant safety as a result of the proposed changes. 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 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. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: February 12, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) Limiting Condition for 
Operation 3.9.4, ``Containment Penetrations'', to allow penetrations 
included under TS 3.9.4(c) to be opened during core alterations or 
movement of irradiated fuel, under administrative controls. This change 
is based on the TS Task Force Traveler No. 312-A, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow containment penetrations 
identified under Technical Specification 3.9.4(c) to remain open 
during fuel movement and core alterations. These penetrations are 
normally closed during this time period to prevent the release of 
radioactive material in the event of a Fuel Handling Accident inside 
containment. These penetrations are not initiators of any accident. 
The probability of a Fuel Handling Accident is unaffected by the 
status of these penetrations.
    The Fuel Handling Accident analyses demonstrate that the maximum 
offsite dose is well [within] the acceptance limits specified in SRP 
[Standard Review Plan] 15.7.4, and the control room dose is within 
the acceptance criteria specified in GDC [General Design Criterion] 
19. Furthermore, the existing analysis results are independent of 
the containment release path, and therefore are unaffected by the 
proposed change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change will 
not alter the design, configuration, or method of operation of the 
plant beyond the standard functional capabilities of the equipment. 
The proposed change involves a Technical Specification change that 
will allow containment penetrations identified under Technical 
Specification 3.9.4(c) to remain open during fuel movement and core 
alterations. Open penetrations are not accident initiators, and will 
not create the possibility of a new kind of accident. Administrative 
controls will be implemented to ensure the capability to close the 
affected containment penetrations in the event of a Fuel Handling 
Accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change has the potential to slightly increase the 
post-Fuel Handling Accident dose at the site boundary and in the 
control room. However, the existing analyses take no credit for 
containment of the release, so that the existing analysis results 
will remain bounding.

    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 W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.

[[Page 17950]]

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: January 19, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 5.5.9, ``Diesel Fuel Oil Testing 
Program,'' by relocating a reference to a specific American Society for 
Testing and Materials (ASTM) international standard for fuel oil 
testing to licensee-controlled documents, and by adding an alternate 
criteria to the ``clear and bright'' acceptance test for new fuel oil, 
per the consolidated line item improvement process (CLIIP).
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on February 22, 2006 
(71 FR 9179), on possible amendments concerning the CLIIP, including a 
model safety evaluation and a model no significant hazards 
consideration determination. The NRC staff subsequently issued a notice 
of availability of the models for referencing in license amendment 
applications in the Federal Register on April 21, 2006 (71 FR 20735), 
as part of the CLIIP.
    In its application dated January 19, 2007, the licensee affirmed 
the applicability of the following determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. Future changes to the licensee-controlled document 
will be evaluated pursuant to the requirements of 10 CFR 50.59, 
``Changes, tests and experiments,'' to ensure that such changes do 
not result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do 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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. The requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs.
    Therefore, the changes do 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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs.
    Changes to the licensee-controlled document are performed in 
accordance with the provisions of 10 CFR 50.59. This approach 
provides an effective level of regulatory control and ensures that 
diesel fuel oil testing is conducted such that there is no 
significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell A. Gibbs

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: December 14, 2006, as supplemented by 
letter dated March 14, 2007.
    Description of amendment request: The proposed amendment would 
modify the technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The 
changes are consistent with NRC approved Industry/Technical 
Specification Task Force (TSTF) standard TS change TSTF-372, Revision 
4.
    The proposed amendment includes an administrative change to LCO 
3.0.1 that will clarify that LCO 3.0.7 allows specified TS requirements 
to be suspended during physics tests performed in accordance with TSs 
3.1.8 and 3.1.9. This administrative change will make the CR-3 TSs more 
consistent with the standard TSs and with TSTF-372, Revision 4.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated April 26, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:


[[Page 17951]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. Entrance into Actions 
or delaying entrance into Actions is not an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on the delay time allowed 
before declaring a TS supported system inoperable and taking its 
Conditions and Required Actions are no different than the 
consequences of an accident under the same plant conditions while 
relying on the existing TS supported system Conditions and Required 
Actions. Therefore, the consequences of an accident previously 
evaluated are not significantly increased by this change. Therefore, 
this 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 allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. Thus, this 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 allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
restores an allowance in the pre-ISTS [improved Standard Technical 
Specifications] conversion TS that was unintentionally eliminated by 
the conversion. The pre-ISTS TSs were considered to provide an 
adequate margin of safety for plant operation, as does the post-ISTS 
conversion TS. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: March 16, 2007.
    Description of amendment request: The proposed amendment would add 
new Technical Specification (TS) requirements for the response times 
associated with a steam generator feedwater pump (SGFP) trip and 
feedwater isolation valve (FIV) closure. The amendment would also 
revise the TS requirements for the containment fan cooler unit (CFCU) 
cooling water flow rate. These changes are associated with a revised 
containment response analysis that credits a SGFP trip and FIV closure 
(on a feedwater regulator valve failure) to reduce the mass/energy 
release to the containment during a main steam line break (MSLB). The 
containment analysis also credits a reduced heat removal capability for 
the CFCUs, allowing a reduction in the required service water (SW) flow 
to the CFCUs.
    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 does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change establishes response time requirements for 
feedwater isolation and reduced CFCU flow rates to support 
containment analyses to accommodate reduced CFCU heat removal 
capacity. The changes in analysis input assumptions affect plant 
response to an accident and are not accident initiators; therefore, 
they have no bearing on the probability of an accident. The Salem 
FSAR [Final Safety Analysis Report] Chapter 15 accidents which are 
impacted by a change in the CFCU modeling parameters are LOCA [loss-
of-coolant accident] and MSLB mass and energy release Containment 
analyses. The consequences of these postulated accidents are shown 
to be acceptable using assumptions consistent with the proposed 
changes.
    For the LOCA transients, the containment cooling systems are 
considered for three aspects: core response, containment response 
and dose. The core response is most limiting when the containment 
conditions minimize back pressure since this increases the blowdown 
and reduces the effectiveness of the ECCS [emergency core cooling 
system]. The LOCA core response (10 CFR 50.46 [Section 50.46 of 
Title 10 of the Code of Federal Regulations]--PCT [peak cladding 
temperature]) is conservatively biased to minimize the containment 
backpressure such that any safety injection effectiveness is 
minimized (the core becomes the highest resistance flow path). Thus, 
any reduction in the accident capability of the CFCUs has no bearing 
on the LOCA core response.
    The bounding containment integrity analyses are the LBLOCA 
[large-break LOCA] and the MSLB Inside Containment events. The 
containment integrity analysis relies on two heat removal paths to 
maintain containment pressure and temperature conditions. The CFCU 
air-to-water heat exchangers reject containment energy to the SW 
System and the Containment Spray System removes containment energy 
by using spray droplet direct contact heat exchange to transfer the 
energy from the containment ambient to the containment sump, where 
it is transferred out of containment via the RHR [residual heat 
removal] heat exchanger and CCW [component cooling water]/SW 
Systems. Containment integrity analyses for both LOCA and MSLB, 
using input assumptions consistent with the proposed changes, show 
that containment integrity is maintained with reduced CFCU heat 
removal capacity.
    The potential dose impacts due to reduced CFCU heat removal 
capacity are bounded as the design basis assumptions concerning the 
number of operating CFCUs (three of five), and the thermal-hydraulic 
transient operation of the Containment Spray System are unchanged. 
The Salem design basis only credits Containment Spray iodine removal 
effectiveness during the LOCA injection and recirculation phases 
based on a single failure of an entire ESF [engineered safety 
features] train. This assumption results in 3 of 5 CFCUs being 
available to ensure adequate mixing of the containment ambient air 
as well as operation of a single Containment Spray Train, which 
controls containment spray droplet size and pH, as described in 
UFSAR [updated FSAR] Section 6.2.3. As a further conservatism, the 
current LOCA Alternate Source Term (AST) analysis (Calculation S-C-
ZZ-MDC-1945, an interim revision of which was sent to the NRC 
[Nuclear Regulatory Commission] staff for review via letter dated 
September 16, 2004) only credits two CFCUs for mixing. The 
Containment Building and Auxiliary Building leakage rates are 
unaffected by the revised containment analysis as the peak 
containment pressure and temperatures are less than the design basis 
values described in the Salem UFSAR. Therefore, there is no impact 
on offsite dose rates due to the reduced CFCU heat removal capacity.
    One other high energy line break for consideration is the 
rupture of a feedwater line break. From a containment response 
aspect, this event is bounded by the MSLB event, so it is not 
explicitly analyzed (or even discussed in the Salem UFSAR).
    A review of the Salem design basis for AST dose calculations 
shows that the revised Containment Integrity Analysis, WCAP-16503, 
does not challenge any of the assumptions that are part of the AST 
design basis.
    Section 6.2 of the UFSAR indicates that the Appendix J Type A 
containment leak rate test

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pressure is based on the containment design pressure of 47.0 psig, 
not the calculated accident pressure. Since the design pressure 
value bounds the peak pressure calculated in WCAP-16503 and is not 
being changed, the Appendix J testing requirements are not impacted.
    Thus, in conclusion, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The proposed change modifies response 
time requirements for feedwater isolation, and reduces CFCU flow 
rates and heat removal requirements consistent with the new 
containment analysis.
    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 does no