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BILLING CODE 7590-01-M NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to section 189a.
(2)of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from October 13, 2006, to October 26, 2006. The last biweekly notice was published on October 24, 2006 (71 FR 62306). 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 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.* Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona *Date of amendments request:* September 28, 2006. *Description of amendments request:* The proposed amendments would revise certain Technical Specification
(TS)requirements for mode change limitations in Limiting Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4. This request is consistent with NRC-approved Industry/Technical Specification Task Force
(TSTF)Traveler number TSTF-359, Revision 9, “Increase Flexibility in Mode Restraints.” In addition, the proposed amendments would correct TS Example 1.4-1, “Surveillance Requirements,” to accurately reflect the changes made by TSTF-359, which is consistent with NRC-approved TSTF-485, Revision 0. *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 revises Section 1.4, Frequency, “Example 1.4-1,” to be consistent with Surveillance Requirement
(SR)3.0.4 and Limiting Condition for Operation
(LCO)3.0.4. This change is considered administrative in that it modifies the example to demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly explained in the associated Bases. As a result, modifying the example will not result in a change in usage of the Technical Specifications (TS). The proposed change does not adversely affect accident initiators or precursors, the ability of structures, systems, and components
(SSCs)to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Therefore, this change is considered administrative and will have no effect on the probability or consequences of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. No new or different accidents result from utilizing the proposed change. The change does 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 change does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis. The proposed change is 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. 3. Does the proposed change involve a significant reduction in a margin of safety? *Response:* No. The proposed change is administrative and will have no effect on the application of the Technical Specification requirements. Therefore, the margin of safety provided by the Technical Specification requirements is unchanged. 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 that review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. *Attorney for licensee:* Michael G. Green, Senior Regulatory Counsel, Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, Phoenix, Arizona 85072-2034. *NRC Branch Chief:* David Terao. Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2 New London County, Connecticut *Date of amendment request:* March 17, 2006. *Description of amendment request:* The proposed amendment would revise Millstone Power Station, Unit No. 2 Technical Specification
(TS)3.4.4 to replace the existing maximum and minimum pressurizer water volume and water level limits with a maximum water level limit. The associated TS bases will be updated to address the proposed change. *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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below. 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change does not change the accident analysis of record, maintains the current maximum operating pressurizer level at its present value, does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change has no effect on the consequences of any analyzed accident since the change does not affect the function of any equipment credited for accident mitigation. Therefore, the proposed amendment does not increase the probability or consequences of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. Since the proposed change does not modify any plant equipment and there is no impact on the capability of existing equipment to perform its intended functions and no new failure modes are introduced by the proposed change, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Involve a significant reduction in a margin of safety? The proposed change maintains the current maximum operating pressurizer level at its present value, and the acceptance criterion for the maximum pressurizer level is unchanged. Since there are no changes, the proposed change does not involve a reduction in a margin of safety. 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:* Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385. *NRC Branch Chief:* Harold K. Chernoff. Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana *Date of amendment request:* October 16, 2006. *Description of amendment request:* The proposed change will add an NRC previously approved topical report to the analytical methods referenced in Technical Specification
(TS)Section 5.6.5, “Core Operating Limits Report (COLR).” The current method of performing the loss-of-coolant accident (LOCA ) analyses will be replaced by an updated method described in AREVA NP (formerly known as Framatome or Siemens) topical report, “EXEM BWR-2000 [Boiling-Water Reactor—2000] ECCS [Emergency Core Cooling System] Evaluation Model.” *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. Core operating limits are established each operating cycle in accordance with TS 3.2, “Power Distribution” and TS 5.6.5, “Core Operating Limits Report (COLR)”. These core operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). In addition, the Average Planar Linear Heat Generation Rate (APLHGR) operating limits imposed by Technical Specification 3.2.1 also ensure that the peak cladding temperature
(PCT)during the postulated design basis LOCA does not exceed the 2200 °F limit specified in 10 CFR 50.46. The APLHGR is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location. The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS section 5.6.5.b. A change to TS section 5.6.5.b is requested to include an updated LOCA analysis method, EXEM BWR-2000. The updated method will be used to determine the APLHGR operating limits imposed by Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and approved by the NRC and is applicable to the RBS [River Bend Station] plant design and the AREVA NP fuel being used at RBS. The application of the LOCA analytical model will continue to ensure that the APLHGR operating limits are established to protect the fuel cladding integrity during normal operation, AOOs, and the design basis LOCA. The requested TS changes concern the use of analytical methods and do not involve any plant modifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. The proposed TS amendment will not change the design function, reliability, performance, or operation of any plant systems, components, or structures. It does not create the possibility of a new failure mechanism, malfunction, or accident initiators not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC approved methods that are applicable to the RBS design and the RBS fuel. 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 ECCS performance analysis methods are used to establish the APLHGR limits required by Technical Specification 3.2.1. The APLHGR limits are specified in the COLR and are the result of fuel design, design basis accident (DBA), and transient analyses. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during anticipated operational occurrences
(AOOs)and that the peak cladding temperature
(PCT)during the postulated design basis LOCA does not exceed the 2200 °F limit specified in 10 CFR 50.46. The EXEM BWR-2000 evaluation model is an updated LOCA analytical method that has been approved by the NRC and is applicable to the RBS plant design and the fuel being used at RBS. A RBS plant specific ECCS performance analysis has been performed with the EXEM BWR-2000 evaluation model. This evaluation concluded that the resulting PCT still afforded adequate margin to the 2200 °F limit of 10 CFR 50.46. 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:* Mark Wetterhahn, Esq., Winston & Strawn LLP, 1700 K Street, NW., Washington, DC 20006. *NRC Branch Chief:* David Terao. Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York *Date of amendment request:* September 25, 2006. *Description of amendment request:* The amendment proposes revisions to the Technical Specifications that are editorial in nature and consist of typographical corrections, update of references, and deletion of obsolete notes. *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 editorial in nature and have no affect on accident scenarios previously evaluated. The proposed changes include typographical corrections, consistent with the current version of the Standard Technical Specifications (NUREG 1431, Revision 3); updated references, consistent with the current version of the Entergy Quality Assurance Program Manual (Revision 13); and deletion of notes that provided one-time allowances or are otherwise now obsolete. The proposed changes do not affect initiating events for accidents previously evaluated and do not affect or modify plants systems or procedures used to mitigate the progression or outcome of those accident scenarios. 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 do not involve the installation of new plant equipment or modification of existing plant equipment. No system or component setpoints are being changed and there are no changes being proposed for the way that the plant is operated. There are no new accident initiators or equipment failure modes resulting from the proposed changes. The proposed changes are editorial in nature, consisting of typographical corrections, reference updates, and deletion of obsolete notes. 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 editorial in nature and do not change setpoints or limiting parameters specified in the plant Technical Specifications. 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 Branch Chief:* Richard J. Laufer. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas *Date of amendment request:* August 31, 2006. *Description of amendment request:* Entergy Operations, Inc., proposes to relocate Technical Specification
(TS)3.8.7 requirements associated with 120 Volt Inverter Y-28 and TS 3.8.9 requirements associated with 120 VAC electrical power distribution subsystem panel C-540 to the Technical Requirements Manual (TRM). *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? *Response:* No. The proposed change does not physically alter any plant structures, systems, or components and does not affect or create new accident initiators or precursors. The loss of Y-28, in itself, has no significant impact on station operation because its associated instrument panel, C-540, remains energized from an Emergency Diesel Generator
(EDG)backed vital AC source. A potential loss of vital instrument panel C-540 does not prevent the fulfillment of a safety function and does not cause Emergency Safeguard Features
(ESF)systems actuations that could render multiple ESF-related trains incapable of performing their intended safety function. Therefore, there is no effect on probability of accidents previously evaluated. The proposed change relocates operability requirements for Y-28 and C-540 to the TRM. The TRM is part of the Safety Analysis Report
(SAR)and is controlled under 10 CFR 50.59. In addition, TS-related components powered by C-540 continue to be governed by other TSs that limit the time in which the components can be out of service or provide compensatory measures during the out-of-service period. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. The proposed change does not physically alter any structures, systems, or components, and does not affect or create new accident initiators or precursors. The accident analysis assumptions and results are unchanged. No new failures or interactions have been created. In addition, the proposed change does not introduce new failure modes or mechanisms associated with plant operation and will not create a new accident type. 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 applicable margin of safety is the period of time that equipment important to safety is inoperable. There is no increase in risk that is a result of the proposed change because
(1)affected non-TS components are not safety significant,
(2)compensatory measures are procedurally established for those components governed by other regulation (i.e., 10 CFR [Part] 50, Appendix R), and
(3)TS-related component out-of-service time or related compensatory actions are governed by other existing TSs. The proposed change does not affect any safety limits, other operational parameters, or setpoints in the TS, nor does it affect any margins assumed in the accident analyses. In addition, Y-28 and C-540 operability requirements will be relocated to the TRM, which is part of the Safety Analysis Report
(SAR)and controlled by 10 CFR 50.59. 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:* Nicholas S. Reynolds, Esquire, Winston and Strawn, 1700 K Street, NW., Washington, DC 20006-3817. *NRC Branch Chief:* David Terao. 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 amendments request:* June 30, 2006. *Description of amendments request:* The amendments would relocate the movable incore detectors and radioactive gaseous effluent oxygen monitoring instrumentation from the Technical Specifications to the Updated Final Safety Analysis Report (UFSAR). *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed change[s] involve a significant increase in the probability or consequences of an accident previously evaluated? *Response:* No. The proposed amendment would relocate Technical Specification
(TS)3/4.3.3.2, “Movable Incore Detectors,” and TS 3/4.3.3.9 from the TS to the UFSAR. Movable Incore Detectors and Radioactive Gaseous Effluent Oxygen Monitoring Instrumentation are not initiators to any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. Movable Incore Detectors and Radioactive Gaseous Effluent Oxygen Monitoring Instrumentation are not accident mitigating structures, systems, or components. No impact on the plant response to accidents will be created. Thus the consequences of accidents previously analyzed are unchanged between the existing TS requirements and the proposed changes. The proposed revision to TS SR [Surveillance Requirement] 4.11.2.5 is an administrative change to a reference necessitated by the proposed relocation of TS Table 3.3-13 from the TS to the UFSAR. The proposed revision to the TS Index, page renumbering, and minor format changes to improve consistency are also administrative changes necessitated by the proposed relocation of TS 3/4.3.3.2 and TS 3/4.3.3.9 from the TS to the UFSAR. Therefore, the proposed changes do not involve a significant increase in the probability or radiological consequences of an accident previously evaluated. 2. Do the proposed change[s] create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Specifically, no new hardware is being added to the plant as part of the proposed changes, no existing equipment is being modified, and no significant changes in operations are being introduced. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Do the proposed change[s] involve a significant reduction in a margin of safety? *Response:* No. The proposed changes will not alter any assumptions, initial conditions, or results of any accident analyses. The Movable Incore Detectors and oxygen monitoring instrumentation will continue to perform as before. The proposed changes relocate TS 3/4.3.3.2 and TS 3/4.3.3.9 from the TS to the UFSAR consistent with the guidance in NRC Generic Letter 95-10 and 10 CFR 50.36, and make conforming administrative changes to the TS Index, page renumbering, and minor format changes to improve consistency and to TS SR 4.11.2.5 to reflect the relocation of TS 3/4.3.3.9 from the TS to the Salem UFSAR. 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:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Harold K. Chernoff. 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 amendments request:* September 26, 2006. *Description of amendments request:* The amendments would revise Technical Specification 6.9.1.9 to remove the revision number and date for the topical reports that contain the analytical methods used in the Core Operating Limits Report (COLR). The effect of this change is to allow the licensee to use current topical reports, as long as they have been approved by the NRC. The amendments would also add an NCR-approved topical report to the Salem Nuclear Generating Station, Unit No. 2, COLR methods. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? *Response:* No. The proposed changes affect the administrative controls section of Technical Specifications
(TS)that govern the analytical methods used to determine core operating limits. Removal of revision levels and dates from NRC-approved methods listed in TS is an administrative change that has no impact on the probability or consequences of an accident. TS 6.9.1.9.b will still require these methods to be reviewed and approved by [the] NRC. The proposed change does not affect the required TS actions to be taken in the event that any core operating limits are exceeded. The proposed use of WCAP-10054-P-A, Addendum 2 for the Salem Unit 2 Small Break Loss of Coolant Accident (SBLOCA) analysis is consistent with the limitations and conditions of NRC approval. The parameters assumed in the analysis are within the design limits of the plant equipment. Therefore, there will be no increase in the probability of a loss of coolant accident. The consequences of a LOCA are not being increased, since it is shown that the Emergency Core Cooling System
(ECCS)is designed so that its calculated cooling performance conforms to the criteria contained in 10 CFR 50.46, Paragraph b. No other accident is potentially affected by this change. Therefore, the proposed change does 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. No new modes of plant operation are being introduced. The parameters assumed in the analysis are within the design limits of the plant equipment. TS will continue to require operation within the core operating limits determined using NRC-approved analytical methods and the proposed change does not affect any actions required in the event the core operating limits are exceeded. Therefore, the proposed change does not involve an increase in the probability or consequences of an accident previously evaluated. 3. Do the proposed change[s] involve a significant reduction in a margin of safety? *Response:* No. The proposed changes do not have any impact on plant equipment or safety analysis acceptance criteria. Core operating limits will continue to be determined using NRC-approved analytical methods. The ECCS acceptance criteria of 10 CFR 50.46 will continue to be met following the proposed use of WCAP-10054-P-A, Addendum 2 for the Salem Unit 2 SBLOCA analysis[.] Therefore, the proposed change[s] do[es] 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:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Harold K. Chernoff. PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, Unit No. 2, Salem County, New Jersey *Date of amendment request:* April 6, 2006. *Description of amendment request:* The amendment would change the Technical Specifications to reduce the maximum allowable reactor power when two main steam safety valves (MSSVs) are inoperable. *Basis for proposed no significant hazards consideration determination:* As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do[es] the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? *Response:* No. The proposed change reduces the power level at which Salem Unit 2 may be operated with a maximum of two inoperable MSSVs in any steam generator. This change is consistent with analyses of the limiting transients for secondary system pressure (loss of load/turbine trip and rod withdrawal at power), performed to demonstrate the acceptance criterion of 110% of design pressure will continue to be met following steam generator replacement. The proposed change does not involve any changes to the MSSV actuation setpoints; they remain well above the Main Steam System operating pressures. The proposed change does not challenge the relief capacity of the MSSVs. Therefore, the probability of an event associated with mis-operation of the MSSVs (e.g., inadvertent depressurization of the Main Steam System) is not impacted by the proposed change. The proposed reduction in allowable power level establishes initial conditions consistent with the safety analyses, and does not affect the probability of any previously evaluated accident. The proposed change is necessitated by analyses of limiting secondary system pressure transients, whose acceptance criteria continue to be met provided that plant operation is restricted to 58% RTP [rated thermal power] with a maximum of two inoperable MSSVs in any steam generator. There is no impact on any radiological consequences of an accident associated with the proposed reduction in maximum power level. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do[es] the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. Reducing the allowable power level per the proposed change does not introduce any new accident scenarios or malfunctions. The proposed change establishes an operating restriction consistent with current safety analysis methodology. It represents a change to an input assumption used in analyses of limiting secondary pressurization transients to ensure plant operation does not challenge any design limits. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do[es] the proposed change involve a significant reduction in a margin of safety? *Response:* No. Acceptable margins of safety are inherent in the safety analysis acceptance criteria, including the limit on secondary system pressure to 110% of design pressure during a loss of load/turbine trip (LOL/TT) or rod withdrawal at power
(RWAP)transient. The purpose of the proposed change is to limit operation with a maximum of two inoperable MSSVs for any steam generator, such that the acceptance criterion for secondary pressure continues to be met. The proposed change does not modify any acceptance criteria, nor would it cause any design limit to be exceeded. 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:* Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. *NRC Branch Chief:* Harold K. Chernoff. R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York *Date of amendment request:* September 29, 2006. *Description of amendment request:* The proposed amendment would revise Technical Specification 3.7.8, “Service Water
(SW)System,” to change the limiting conditions for operation (LCOs), Actions, Completion Times, and Surveillance Requirements (SRs). Specifically, the proposed amendment would change the LCO to require a specific number of SW pumps to be operable rather than the current SW train operability. The LCO Actions, Completion Times, and SRs would also be revised based on pump operability. *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 safety related function of the Service Water
(SW)System is to provide cooling for safety related equipment, mitigate the containment response effects of a Main Steam Line Break
(MSLB)and design basis Loss of Coolant Accident (LOCA), and provide long term containment and core cooling in the event of a LOCA. The operation of the SW system, including the number of pumps operating or available, has no affect on the probability of these accidents. The probability of a loss of SW event is not increased. The proposed TS provides for more restrictive actions for pump inoperability than the existing TS, thereby reducing the probability of this event. The consequences of a[n] MSLB or LOCA or other design basis accidents are not increased beyond that assumed in the accident analysis. Two service water pumps are sufficient for all accident mitigation functions. The change provides for adequate service water supply (2 pumps) for both normal and accident conditions. The availability of associated power supplies is also considered. For a reduction in the total number of available pumps, appropriate LCO action statements ensure that the pumps are returned to service within a time limit commensurate with an acceptable level of plant safety and risk, or the plant is placed in a safe mode. The loss of SW has been previously evaluated and measures implemented to mitigate the event. Since a loss of SW assumes no SW pumps are operating, the proposed amendment has no affect on consequences of this event. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. The only accidents directly initiated from this system are the loss of SW or flooding concerns. Both of these accidents have been previously evaluated with acceptable results. Therefore, 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. This change will ensure that sufficient SW pumps are available for accident mitigation at any one time while still providing the appropriate operational flexibility. A risk determination demonstrates that any increase in risk associated with this change is within the established regulatory guidelines. The technical analysis shows that appropriate action statements exist to ensure adequate SW is available for accident mitigation, considering emergency power supply availability. Therefore, this 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:* Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005. *NRC Branch Chief:* Richard J. Laufer. R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York *Date of amendment request:* October 12, 2006. *Description of amendment request:* The proposed amendment would revise Technical Specification
(TS)4.3.3, “Capacity,” to change the limit on the number of fuel assemblies in the spent fuel pool. The proposed amendment would also revise TS 3.7.13, “Spent Fuel Pool Storage,” to remove the references to Type 4 spent fuel pool storage racks, which are not currently installed. *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 reduces the total number of fuel assemblies that can be stored in the current spent fuel pool storage locations and reduces the number of available locations. This will limit the potential inventory of spent fuel in the pool. The probability of an accident has not changed since the number of stored fuel assemblies is not a precursor for a spent fuel handling accident. A comparison of the criticality analysis of fuel assemblies to be used in subsequent Extended Power Uprate core reloads to the current criticality analysis has been performed. The design parameter assumptions used in the licensing basis criticality analyses are bounding. There are no new components or new functions associated with the spent fuel cooling system so the probability of an accident has not changed. The effect of a single failure on the spent fuel pool system's capability to provide for heat removal from the fuel pool has been analyzed. The analysis concluded that the system remains within the parameters previously evaluated. The implementation of the Extended Power Uprate does not affect the capability of the system to perform its function. The Extended Power Uprate conditions do not add any new or previously unevaluated materials to the spent fuel pool storage system and do not include any reductions in the boron concentration requirements so the probability of an accident has not changed. The total soluble boron concentration required to maintain the spent fuel pool in a subcritical condition with the transition to the new fuel has not changed. The conclusions in the Ginna UFSAR [Updated Final Safety Analysis Report], assuming the most limiting accident, remain valid. Therefore, the consequences of a fuel handling accident, a loss of spent fuel cooling, and a boron reduction concentration event previously evaluated have not increased. 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 do not alter the function of the spent fuel pool or any related equipment, nor cause it to operate differently than it was designed to operate. All equipment required to mitigate the consequences of an accident would continue to operate as before. The proposed changes reduce the maximum number of fuel assemblies that can be stored in the spent fuel pool and the number of storage locations. Therefore, 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 changes reduce the maximum number of fuel assemblies that can be stored in the spent fuel pool and the number of storage locations. The changes are in accordance with conclusions supporting Extended Power Uprate and have been determined to be acceptable. The design parameter assumptions used in the licensing basis criticality analysis bound those of the new fuel assemblies. Although the individual heat load per assembly has increased due to the changed fuel design, the maximum spent fuel pool heat load has decreased due to the reduction in the number of fuel assemblies that will be stored based on future plans to use dry cask storage. Therefore, this proposed change does not reduce the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005. *NRC Branch Chief:* Richard J. Laufer. Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia *Date of amendment request:* October 3, 2006. *Description of amendment request:* The proposed amendment would revise the Technical Specifications surveillance requirements related to inspection of the containment sump trash racks and screens, inside recirculation spray
(RS)pump wells, and outside RS and low head safety injection pump suction inlets resulting from Nuclear Regulatory Commission's (NRC's) Generic Safety Issue
(GSI)191 and Generic Letter
(GL)2004-02. *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? The proposed change does not impact the condition or performance of any plant structure, system or component. Furthermore, the proposed change does not affect the initiators of any previously analyzed event or the assumed mitigation of accident or transient events since the plant will be operated in the same manner and within the same operating limits that are currently in place. The proposed TS change is administrative in nature given that inspection of containment sump components for debris accumulation and structural integrity will continue to be performed. The revised TS surveillance wording is being implemented as a clarification to facilitate inspection of the containment sump in its current configuration, as well as after containment sump modifications have been implemented in response to GSI-191 and GL 2004-002. As a result, the proposed change to the Surry TS does not involve any increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities nor consequences are being affected by this proposed change. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change is administrative in nature and, as such, does not involve any changes in station operation or physical modifications to the plant. In addition, no changes are being made in the methods used to respond to plant transients that have been previously analyzed. No changes are being made to plant parameters within which the plant is normally operated or in the setpoints, that initiate protective or mitigative actions, since the plant will be operated in the same manner and within the same operating limits that are currently in place. Since plant operation will not be affected by this change, no new failure modes are being introduced. Therefore, the proposed change to the Surry TS does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated. 3. Does the change involve a significant reduction in the margin of safety? The proposed change is administrative in nature given that inspection of the containment sump components for debris accumulation and structural integrity will continue to be performed on an established frequency. The more general nature of the TS surveillance wording is being implemented as a clarification to facilitate inspection of the containment sump in its current configuration, as well as after containment sump modifications have been implemented in response to GSI-191 and GL 2004-002. The proposed change does not impact station operation or any plant structure, system or component that is relied upon for accident mitigation. Furthermore, the margin of safety assumed in the plant safety analysis is not affected in any way by the proposed change since the plant will be operated in the same manner and within the same operating limits and setpoints that are currently in place. Therefore, the proposed change to the Surry Technical Specifications does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. *Attorney for licensee:* Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385. *NRC Branch Chief:* Evangelos C. Marinos. Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing in connection with these actions was published in the **Federal Register** as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see
(1)the applications for amendment,
(2)the amendment, and
(3)the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, *http://www.nrc.gov/reading-rm/adams.html.* If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1
(800)397-4209,
(301)415-4737 or by e-mail to *pdr@nrc.gov.* Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina *Date of application for amendment:* November 30, 2005. *Brief description of amendment:* The amendment revises the surveillance requirements
(SR)for the emergency diesel generator automatic trips bypass of SR 3.8.1.11 from 18 months to 24 months. *Date of issuance:* October, 4, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days. *Amendment No.* 208. *Renewed Facility Operating License No. DPR-23.* Amendment revises the Technical Specifications. *Date of initial notice in* Federal Register: February 28, 2006 (71 FR 10072). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 4, 2006. 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 Duke Power Company LLC, et al., Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina *Date of application for amendments:* July 27, 2005, as supplemented May 4, 2006, and August 8, 2006. *Brief description of amendments:* The amendments revise the Catawba and McGuire Technical Specification 3.4.15, “RCS Leakage Detection Instrumentation.” *Date of issuance:* September 30, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days from the date of issuance. *Amendment Nos.:* 234/230 and 235/217. *Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and NPF-17:* Amendments revised the licenses and the technical specifications. *Date of initial notice in* Federal Register: August 30, 2006 (71 FR 51644). The supplement dated August 8, 2006, provided clarifying information that did not expand the scope of the July 27, 2005, application as modified May 4, 2006. The Commission's related evaluation, Final No Significant Hazards Finding, and State consultation of the amendments are contained in a Safety Evaluation dated September 30, 2006. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket No. STN 50-457, Braidwood Station, Unit No. 2, Will County, Illinois *Date of application for amendment:* November 18, 2005, as supplemented by letters dated August 18 and September 28, 2006. *Brief description of amendment:* The amendment revised TS 5.5.9, “Steam Generator
(SG)Tube Surveillance Program,” regarding the required SG inspection scope for Braidwood Station, Unit No. 2, during refueling outage 12 and the subsequent operating cycle. *Date of issuance:* October 24, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days. *Amendment No.:* 141. *Facility Operating License No. NPF-77:* The amendment revised the Technical Specifications and License. *Date of initial notice in* Federal Register: (71 FR 29676; May 23, 2006). The August 18 and September 28, 2006, supplements contained clarifying information and did not change the NRC 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 October 24, 2006. No significant hazards consideration comments received: No. FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire *Date of amendment request:* March 23, 2006. *Description of amendment request:* The amendment deleted License Condition 2.G, “Reporting to the Commission,” as described in the Notice of Availability published in the **Federal Register** on April 25, 2006 (71 FR 23955). The change was requested as part of the consolidated line item improvement process and consistent with the model safety evaluation published in the **Federal Register** on November 4, 2005 (70 FR 67202). *Date of issuance:* October 17, 2006. *Effective date:* As of its date of issuance, and shall be implemented within 90 days. *Amendment No.:* 113. *Facility Operating License No. NPF-86:* The amendment revised Facility Operating License No. NPF-86 and the Technical Specifications. *Date of initial notice in* Federal Register: April 25, 2006 (71 FR 23955). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 17, 2006. No significant hazards consideration comments received: No. Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska *Date of amendment request:* June 16, 2006. *Brief description of amendment:* The amendment revised the Technical Specification 3.10.1, “Inservice Leak and Hydrostatic Testing Operation,” to extend the scope to include provisions for temperature increases above 212 °F as a consequence of inservice leak or hydrostatic testing, and as a consequence of control rod scram time testing initiated in conjunction with the inservice leak test or hydrostatic test, when initial test conditions are below 212 °F. *Date of issuance:* October 23, 2006. *Effective date:* As of the date of issuance and shall be implemented within 30 days of issuance. *Amendment No.:* 225. *Facility Operating License No. DPR-46:* Amendment revised the Technical Specifications. *Date of initial notice in* Federal Register: August 1, 2006 (71 FR 43535) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 23, 2006. 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:* December 7, 2005, as supplemented by letters dated July 20 and September 5, 2006. *Brief description of amendments:* These amendments revised the Technical Specifications to delete Surveillance Requirement
(SR)4.9.2.b, which requires performance of a channel functional test
(CFT)of each source range neutron flux monitor within 8 hours prior to the initial start of core alterations. An associated administrative change would renumber current SR 4.9.2.c as SR 4.9.2.b. The amendments would also eliminate the restriction in SRs 4.10.3.2 and 4.10.4.2 that the CFTs of the intermediate and power range monitors be performed within 12 hours prior to initiating physics tests. *Date of issuance:* October 13, 2006. *Effective date:* As of the date of issuance, to be implemented in 60 days. *Amendment Nos.:* 275, 257. *Facility Operating License Nos. DPR-70 and DPR-75:* The amendments revised the Technical Specifications and License. *Date of initial notice in* Federal Register: August 2, 2006 (71 FR 43819). The supplements 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 amendments is contained in a Safety Evaluation dated October 13, 2006. No significant hazards consideration comments received: No. South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina *Date of application for amendment:* November 15, 2005, as supplemented May 31, August 31, and September 29, 2006. *Brief description of amendment:* The amendment revises the Virgil C. Summer Nuclear Station Technical Specifications
(TS)3/4.3 for the reactor trip instrumentation and the engineered safety feature actuation system instrumentation to implement the allowed outage time and bypass test time changes approved in WCAP-14333-P-A, Revision 1, “Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,” and makes several additional changes to TS outside of the scope of WCAP-14333. *Date of issuance:* October 24, 2006. *Effective date:* As of the date of issuance and shall be implemented within 60 days. *Amendment No. 177.* *Renewed Facility Operating License No. NPF-12:* Amendment revises the Technical Specifications. *Date of initial notice in* Federal Register: December 20, 2005 (70 FR 75496). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 24, 2006. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama *Date of application for amendment:* November 10, 2003 (TS-430), as supplemented by letter dated November 8, 2004. *Brief description of amendment:* The amendment incorporates the necessary Technical Specification
(TS)changes for the planned replacement of the power range monitoring portion of the existing Neutron Monitoring System with a digital upgrade. These changes expand the current allowable operating domain to the Maximum Extended Load Line Limit region of the power/flow chart. *Date of issuance:* September 27, 2006. *Effective date:* Date of issuance, to be implemented within 30 days. *Amendment No.:* 262. *Facility Operating License No. DPR-33:* Amendment revised the TSs. *Date of initial notice in* Federal Register: February 3, 2004 (69 FR 5208). The November 8, 2004, supplement, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 27, 2006. No significant hazards consideration comments received: No. Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee *Date of application for amendments:* February 6, 2006. *Brief description of amendments:* The amendments modify Technical Specification
(TS)requirements for inoperable snubbers by adding Limiting Condition for Operation 3.0.7. This operating license improvement was made available by the Nuclear Regulatory Commission
(NRC)on May 4, 2005 (70 FR 23252) as part of the consolidated line item improvement process and is consistent with NRC approved Technical Specification Task Force
(TSTF)standard TS change TSTF-372, Revision 4. *Date of issuance:* October 4, 2006. *Effective date:* As of the date of issuance and shall be implemented within 45 days. *Amendment Nos.* 312/301. *Facility Operating License Nos. DPR-77 and DPR-79:* Amendments revised the technical specifications. *Date of initial notice in* Federal Register: March 28, 2006 (71 FR 15487). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 4, 2006. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 30th day of October 2006. For the Nuclear Regulatory Commission. Catherine Haney, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. E6-18595 Filed 11-6-06; 8:45 am] BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION Notice of Availability of Model License Amendment Request and Safety Evaluation on Technical Specification Improvement Regarding Revision to the Completion Time in STS 3.6.6A, “Containment Spray and Cooling Systems” for Combustion Engineering Pressurized Water Reactors Using the Consolidated Line Item Improvement Process AGENCY: Nuclear Regulatory Commission. ACTION: Notice of availability. SUMMARY: Notice is hereby given that the staff of the U.S. Nuclear Regulatory Commission
(NRC)has prepared a model license amendment request (LAR), model safety evaluation (SE), and model proposed no significant hazards consideration
(NSHC)determination related to changes to the completion times
(CT)in Standard Technical Specification
(STS)3.6.6A, “Containment Spray and Cooling Systems,” contained in NUREG-1432 (Standard Technical Specifications for Combustion Engineering Plants, Rev. 3.0). The proposed changes would revise STS 3.6.6A by extending the CT for one containment spray system
(CSS)train inoperable from 72 hours to seven days, and add a Condition, Required Actions and associated CT when one CSS train and one containment cooling system
(CCS)train are inoperable. These changes are based on analyses provided in a joint applications report submitted by the Combustion Engineering Owner's Group (CEOG). The CEOG participants in the Technical Specifications Task Force
(TSTF)proposed these changes to the STS in Change Traveler No. TSTF-409, Revision 2. The purpose of these models is to permit the NRC to efficiently process amendments to incorporate these changes into plant-specific STS for Combustion Engineering pressurized water reactors (PWRs). Since TSTF-409 involves a risk-informed approach to extending the CT for one CSS inoperable, the NRC staff must verify that licensees who apply for this TS change have a valid, up-to-date probabilistic risk assessment
(PRA)model that employs PRA principles to ensure that public health and safety are maintained when the CSS CT of 7 days is implemented. Therefore, the model LAR contains several conditions requiring licensees to make specific validations of their plant PRA quality and methods. The intent of using the CLIIP to adopt TSTF-409 is to eliminate the need for additional technical review and requests for additional information
(RAIs)on plant-specific amendments. Licensees of nuclear power reactors to which the models apply can request amendments conforming to the models. In such a request, a licensee should confirm the applicability of the model SE and NSHC determination to its plant, and provide the expected supplemental information requested in the model LAR. DATES: The NRC staff issued a Federal Register Notice (71 FR 18380, April 11, 2006) which provided for public comment a model SE, model LAR, and NSHC determination related to changes to the CT for one CSS train inoperable in STS 3.6.6A. The NRC staff herein provides a revised model SE, revised model LAR, and NSHC determination. The NRC staff can most efficiently consider applications based upon the model LAR, which references the Model SE, if the application is submitted within one year of this **Federal Register** Notice. FOR FURTHER INFORMATION CONTACT: Tim Kobetz, Mail Stop: O-12H2, Division of Inspection Program Management, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone 301-415-1932. SUPPLEMENTARY INFORMATION: Background Regulatory Issue Summary 2000-06, “Consolidated Line Item Improvement Process [CLIIP] for Adopting Standard Technical Specifications Changes for Power Reactors,” was issued on March 20, 2000. The CLIIP is intended to improve the efficiency and transparency of NRC licensing processes. This is accomplished by processing proposed changes to the STS in a manner that supports subsequent license amendment applications. The CLIIP includes an opportunity for the public to comment on proposed changes to the STS following a preliminary assessment by the NRC staff and finding that the change will likely be offered for adoption by licensees. The CLIIP includes NRC staff evaluation of any comments received for a proposed change to the STS. In several instances, the staff's evaluation did result in changes to the model LAR and/or model SE. Those licensees opting to apply for the subject changes to TSs are responsible for reviewing the staff's evaluation, referencing the applicable technical justifications, and providing any necessary plant-specific information. The model LAR shows licensees the expected level of detail that needs to be included in order to adopt TSTF-409, Rev. 2, as well as guidelines for staff review. The NRC has established an internal review plan that designates the appropriate staff and approximate timelines to review plant-specific LARs that reference TSTF-409, Rev. 2. Each amendment application made in response to the notice of availability will be processed and noticed in accordance with applicable NRC rules and procedures. This notice involves an increase in the allowed CT to restore an inoperable CSS train on Combustion Engineering PWRs. By letter dated November 10, 2003, the CEOG proposed this change for incorporation into the STS as TSTF-409, Revision 2. This change is based on the NRC staff-approved analyses contained in CE NPSD-1045-A, “Joint Applications Report: Modification to the Containment Spray System, and Low Pressure Safety Injection System Technical Specifications,” dated March 2000, as approved by the NRC in a SE dated December 21, 1999, accessible electronically from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet (ADAMS Accession No. ML993620241) at the NRC Web site *http://www.nrc.gov/reading-rm/adams.html* . Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC Public Document Room Reference staff by telephone at 1-800-397-4209, 301-415-4737, or by e-mail to *pdr@nrc.gov* . This notice, along with TSTF-409, Rev. 2, will be posted on the NRC Web site at *http://www.nrc.gov/reactors/operating/licensing/techspecs/changes-issued-for-adoption.html* . Applicability This proposed change to revise the Technical Specification
(TS)CT for one inoperable CSS train is applicable to Combustion engineering PWRs. To efficiently process the incoming license amendment applications, the NRC staff requests that each licensee applying for the changes addressed by TSTF-409, Revision 2, use the CLIIP to submit a LAR that adheres to the following model. Any deviations from the model LAR should be explained in the licensee's submittal. When applying, licensees should ensure they address the eight conditions and one regulatory commitment listed in the model LAR and model SE. The CLIIP does not prevent licensees from requesting an alternative approach, proposing changes without providing the information described in the eight model LAR conditions, or making the requested commitment. Variations from the approach recommended in this notice may, however, require additional review by the NRC staff and may increase the time and resources needed for the review. Significant variations from the approach, or inclusion of additional changes in the LAR, will result in staff rejection of the submittal under the CLIIP. Instead, licensees desiring significant variations and/or additional changes should either submit a LAR that does not claim to adopt TSTF-409, or specifically state in their LAR that they are adopting TSTF-409 without using the CLIIP. Public Notices In a notice in the **Federal Register** dated April 11, 2006 (71 FR 18380), the staff requested comment on the use of the CLIIP to process requests to revise the CE PWR TS regarding Containment Spray System completion time extensions as discussed in TSTF-409. In response to this notice, the staff received one set of comments (developed by the PWR Owners Group, and submitted by the Nuclear Energy Institute in a letter dating May 10, 2006 (ADAMs Accession No. ML061570029)). Specific comments on the model LAR and model SE were offered. These comments, along with the NRC staff's responses, are summarized and discussed below. 1. *Comment:* Based on discussions with the author regarding the intent of the “Model SE,” [i.e., to allow acceptance review without RAIs while satisfying the CLIIP] it is recommended that additional explanatory information be included. * * * At the very minimum, a clear preamble to the FRN should be provided that places the scope of the FRN in perspective. *Response:* The following preamble has been inserted after the first sentence of the second paragraph of the FRN. “Since TSTF-409 involves a risk-informed approach to extending the CT for one CSS inoperable, the NRC staff must verify that licensees who apply for this TS change have a valid, up to date probabilistic risk assessment
(PRA)model that employs PRA principles to ensure that public health and safety are maintained when the CSS CT of 7 days is implemented. Therefore, the model LAR contains several conditions requiring licensees to make specific validations of their plant PRA quality and methods. The intent of using the CLIIP to adopt TSTF-409 is to eliminate the need for additional technical review and requests for additional information
(RAIs)on plant-specific amendments.” 2. *Comment:* [The FRN] should equally note that existing strategies for approval are valid and may also be used. *Response:* The second to last paragraph of the FRN discusses how a licensee should proceed if it desires to deviate from the approach outlined in the CLIIP. The NRC's position is that, if a licensee is going to submit a LAR that adopts TSTF-409 using the CLIIP, then the plant-specific LAR should provide all the information requested in the model LAR. Any variations/deviations should be explained, and may require additional review by the staff (including issuance of RAIs). Significant variations from the CLIIP methodology should be submitted as normal license amendment requests. The staff has changed the last sentence of second to last paragraph of the FRN to read: “Instead, licensees desiring significant variations and/or additional changes should either submit a LAR that does not claim to adopt TSTF-409, or specifically state in their LAR that they are adopting TSTF-409 without using the CLIIP.” This will correctly define the scope of the review for the staff when processing an incoming LAR that does not conform to the CLIIP. 3. *Comment:* The essence of the proposed CSS TS change focuses on a single CSS train. Thus, the mention of ACTION G (regarding two CSS trains out-of-service) seems unnecessary. *Response:* The staff agrees with this comment. Mention of ACTION G has been removed from Section 4.1 of the model LAR, and Section 3.1 of the model SE. 4. *Comment:* The last paragraph of section 4.2.1 item 1 notes that “If a zero maintenance PRA model is used * * * in performing these calculations, then the licensee must commit to performing no other maintenance during the extended CSS CT * * *”. This restriction has no technical merit. The risk of maintenance is generated as incremental risks from the baseline. The initial submittal noted that for plants with emergency grade fan coolers (most of the applicants), the actual risk increases as a result of removing a CSS out of service is very low. Furthermore, CSS have very little (if any) overlap with other systems. Because the risk important function of CSSs is to maintain the containment pressure within acceptable limits (and control sump temperature to ensure adequate NPSH for ECCS equipment—a function left out of FRN Section 3), those functions can be accommodated by the redundant CS train or the fan coolers. Furthermore, by using RG 1.177 to support low risk, the risk impact of removal of the CSS for the duration of the 7 day AOT is small. Because plants perform maintenance on a frequent basis, not allowing repair or maintenance on another system (which is likely to be of greater risk importance than the CSS) is unnecessary and likely to have worse risk. Another unusual aspect of the restriction implies that the incremental risk calculated using zero maintenance conditions is significantly different from that calculated using annualized plant-wide system out-of-service values. While the baseline PRA for zero maintenance is less than the baseline PRA value for nominal maintenance, its impact on incremental risk will be small. *Response:* The staff accepts NEI's comment in that it creates a regulatory condition that is overly restrictive to plants using a zero maintenance PRA model. The staff has inserted alternate wording (from RG 1.177 Section 2.3.4. #2) to the last sentence of condition 1 in Section 4.2.1 of the model LAR as follows: If the licensee utilizes a “zero maintenance” PRA model for the assessment, they should state they are using a “zero maintenance” model in the evaluation, and provide a discussion as to the ability of that model to produce comparable results to the “average maintenance” assessment. 5. *Comment:* It is understood that documented quantitative external event information for the plants may be limited. However, reference to plant individual plant examination
(IPE)and individual plant examination for external events (IPEEE) and the requirements to explain the evolution of the PRA since 1988 as identified in Section in item 4.2.1 part 2.b is unnecessary. Item 2.c requires that the peer review results be discussed along with the overall disposition of relevant facts and observations (F&Os) and item e (which includes an overall determination of the adequacy of the plant specific PRA with respect to this application). These assessment[s] are current and of more importance to the application. Where external events rely on IPEEE vintage information, a discussion/statement of the risk significance of the spray system in mitigating external events should be performed. *Response:* The staff agrees that peer reviews of plant-specific PRA are important. However, it is equally important to have an understanding of PRA updates and upgrades since the IPE, IPEEE, and peer reviews were conducted, especially if plant improvements and/or commitments are cited and credited in the analyses as being implemented. Licensees who have given this information in prior submittals may incorporate the information by reference. 6. *Comment:* Section 4.2.1 item 3 requirements on consideration of fire and external events and the associated EXPECTATIONS are too restrictive and do not correspond to safety benefits. The CSS has limited risk overlap with fires or external initiating events. Challenges to power induced by tornadoes, high winds or seismic events have limited importance to the spray system and [are] more appropriate with AOTs associated with AC-power related components. It was our understanding that the intent of this restriction was to assure the regulator that the overall combined plant risk remains below a CDF of 10 −4 per year (per requirements of RG 1.174). The intent of this section should be clarified. This requirement should be reduced to providing information regarding the reasons underlying low risk associated with this system. *Response:* The staff acknowledges that, for many plants, the impact of the CT extension on external event risk will be minimal. If this is the case, the licensee needs to confirm this in its submittal and explain why there is limited overlap. 7. *Comment:* Section 4.2.1 item 3 ACCEPTANCE CRITERIA requires “combining internal events, internal flooding, external events and shutdown PRA results.” The requirements for the combination of events should be modified to have the utility provide a technical basis for demonstrating the plant CDF to be less than 10 −4 per year or has no plant specific vulnerabilities (per SECY-88-20). Requirements for a fully quantified external events (including fire) PRA and shutdown PRA [are] beyond the state of the art. Few plants have all the above. The Fire PRA standard is just undergoing peer review and no shutdown PRA standard has been written. Methods for combining these PRA results [are] also not defined (particularly merging shutdown and “at power” PRA results). Instead, it should be noted that the utility may use existing external event evaluations including IPEEE results and qualitative external event assessments, where appropriate, to provide confidence that the overall plant CDF is not within RG 1.174 risk region 1. *Response:* The staff is requesting that licensees provide ΔCDF and ΔLERF calculations for those external events for which the licensee has a PRA. For external events for which the licensee does not have a PRA, the licensee will need to confirm there are no vulnerabilities that would indicate that the total CDF is >10 −4 or the total LERF is >10 − 5 yr. this stipulation allows the staff to ensure that plans whose ΔCDF or ΔLERF calculation puts them in Region II of either Figure 3 or Figure 4 of RG 1.174 are still within the RG 1.174 Section 2.2.4 acceptance guidelines for total plant risk (CDF and LERF). With regard to NEI's comments on a fully-quantified external events (including fire) PRA and shutdown PRA being beyond state-of-the-art, the staff believes the wording in the EXPECTATIONS for Section 4.2.1 condition 3 was misinterpreted. The wording has been revised to read ``(quantitatively and/or qualitatively, as appropriate).'' However, the staff notes that while fire and shutdown PRA standards have not yet been endorsed, there are available methods to quantify fire and shutdown PRA. Therefore, the staff does not believe such evaluations are beyond the state of the art. Rather, they are areas where some evaluation is still ongoing. 8. *Comment:* EXPECTATIONS supporting 4.2.1 item 4. The TS is structured to have a revised CT. Once the new CT is adopted the old CT will disappear as a regulatory item. Thus, there is no entry into an extended CSS CT. It is simply an entry into the CT. There are no significant external event interactions and the outage is limited to a single spray train. Therefore, The Tier 2 requirement should be limited to one CSS out of service, which is already governed in the TS with a cautionary note that Maintenance rule or tier 3 guidance to not simultaneously disable both the emergency grade fan coolers and the sprays. *Response:* The staff agrees that ``extended CT'' should not be used in the model LAR. Appropriate changes will be made here and in other sections of the FRN where appropriate. The staff believes that a tier 2 justification by the licensee is warranted with regard to removing one CSS train from service due to scheduled ``preventive'' maintenance for the 7-day period. If there are no risk-significant configurations or risk-significant external event conditions identified in the tier 2 evaluation, then the licensee should include a statement that there are no risk-significant configurations or external event conditions that would preclude them from using the 7-day CT. 9. *Comment:* End of [Section 4.2.1 item 7]. Note that the RGs provide guidelines. Risk values are not rigid thresholds. Thus small deviations to the guidance can be and are somewhat fuzzy to allow for the mathematical uncertainties inherent in these studies. *Response:* The staff agrees that RG 1.174 and 1.177 guidelines are not rigid standards, and has revised condition 7 to delete the second paragraph of the EXPECTATIONS section. Note that Condition 5 of the model LAR requires licensees to confirm that their CRMP or associated (a)(4) program meets all aspects of Section 2.3.7.2 or RG 1.177. Dated at Rockville, Maryland; this 19th day of October 2006. For the Nuclear Regulatory Commission. Timothy J. Kobetz, Branch Chief, Technical Specifications Branch, Division of Inspection and Regional Support, Office of Nuclear Reactor Regulation. FOR INCLUSION ON THE TECHNICAL SPECIFICATION WEB PAGE THE FOLLOWING EXAMPLE OF A LICENSE AMENDMENT REQUEST
(LAR)WAS PREPARED BY THE NRC STAFF TO FACILITATE THE ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE
(TSTF)TRAVELER TSTF-409, REVISION 2 ``CONTAINMENT SPRAY SYSTEM COMPLETION TIME EXTENSION (CE NPSD-1045-A).'' THE MODEL PROVIDES THE EXPECTED LEVEL OF DETAIL AND CONTENT FOR A LAR TO ADOPT TSTF-409, REVISION 2. LICENSEES REMAIN RESPONSIBLE FOR ENSURING THAT THEIR PLANT-SPECIFIC LAR FULFILLS THEIR ADMINISTRATIVE REQUIREMENTS AS WELL AS NRC REGULATIONS. U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 SUBJECT: [PLANT NAME] APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE COMPLETION TIME FOR CONTAINMENT SPRAY SYSTEM INOPERABILITY IN ACCORDANCE WITH TSTF-409, REVISION 2 Dear Sir or Madam: In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR 50.90), [LICENSEE] is submitting a request for an amendment to the technical specifications
(TS)for [PLANT NAME, UNIT NOS.]. The proposed changes would revise TS 3.6.6A, ``Containment Spray and Cooling Systems,'' by extending from 72 hours to seven days the completion time
(CT)to restore an inoperable containment spray system
(CSS)train. In addition, a Condition would be added to the TS to allow one CSS train and one containment cooling system
(CCS)train to be inoperable for a period of 72 hours. The changes are consistent with NRC-approved Industry Technical Specification Task Force
(TSTF)Standard Technical Specification Change Traveler, TSTF-409, Revision 2, ``Containment Spray System Completion Time Extension (CE NPSD-1045-A).'' Enclosure 1 provides a description and assessment of the proposed changes and confirmation of applicability. Enclosure 2 provides the existing TS pages marked-up to show the proposed changes. Enclosure 3 provides the existing TS Bases marked-up to reflect the proposed changes (for information only). Final TS Bases will be provided in a future update to the Updated Final Safety Analysis Report (UFSAR) in accordance with the Bases Control Program. Attachments 1 through 8 provide the discussions of [LICENSEE'S] evaluations and supporting information with regard to the conditions stipulated in Section 4.2.1 of Enclosure 1. [LICENSEE] requests approval of the proposed license amendment by [DATE], with the amendment being implemented [BY DATE OR WITHIN X DAYS]. in accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated [STATE] Official. I declare under penalty of perjury under the laws of the United States of America that I am authorized by [LICENSEE] to make this request and that the foregoing is true and correct. [Note that request may be notarized in lieu of using this oath or affirmation statement]. If you should have any questions regarding this submittal, please contact [ ]. Sincerely, Name, Title Enclosures: 1. Description and Assessment of Proposed Changes 2. Proposed Technical Specification Changes 3. Proposed Technical Specification Bases Changes (if applicable) Attachments: 1. Licensee's supporting information for condition 1 2. Licensee's supporting information for condition 2 3. Licensee's supporting information for condition 3 4. Licensee's supporting information for condition 4 5. Licensee's supporting information for condition 5 6. Licensee's supporting information for condition 6 7. Licensee's supporting information for condition 7 8. Licensee's supporting information for condition 8 cc: NRR Project Manager Regional Office Resident Inspector State Contact ITSB Branch Chief 1.0 Description The letter is a request to amend Operating License(s) [LICENSE NUMBER(S)] for [PLANT/UNIT NAME(S)]. The proposed changes would revise Technical Specification
(TS)3.6.6A, “Containment Spray and Cooling Systems,” by extending from 72 hours to seven days the completion time
(CT)to restore an inoperable containment spray system
(CSS)train to operable status, and would add a Condition describing the required action and CT when one CSS train and one containment cooling system
(CCS)train are inoperable. The changes are consistent with NRC approved Industry Owner's Group Technical Specification Task Force
(TSTF)Standard Technical Specification Change Traveler TSTF-409, Revision 2 (Rev. 2), “Containment Spray System Completion Time Extension (CE NPSD-1045-A).” TSTF-409, Rev. 2 was approved by the NRC on [DATE]. 2.0 Proposed Change Specifically, the proposed revision extends the CT (or allowed outage time) that one CSS train is permitted to remain inoperable from 72 hours to seven days based on Reference 1, as accepted by, and subject to the limitations specified in, Reference 2. TSTF-409, Rev. 2 states that the longer CT will enhance overall plant safety by avoiding potential unscheduled plant shutdowns and allowing greater availability of safety significant components during shutdown. In addition, TSTF-409, Rev. 2 states that this extension provides for increased flexibility in scheduling and performing maintenance and surveillance activities in order to enhance plant safety and operational flexibility during lower modes of operation. The revision also adds a Condition to allow one CSS train and one CCS train to be inoperable for up to 72 hours. Since Reference 1 did not evaluate the concurrent inoperabilities of one CSS train and one CCS train, the CT for this Condition was limited to 72 hours. [LICENSEE] also proposes to make changes to the supporting TS Bases in accordance with TSTF-409, Rev. 2. Changes to the Bases include supporting information justifying the addition of the Condition for one CSS train and one CCS train inoperable. The Bases changes also include a reviewer's note that requires [LICENSEE] to adopt Reference 1 and meet the requirements of References 1 and 2 prior to utilizing the 7-day CT for one inoperable CSS. Finally, a reference to Reference 1 is added to the Bases. Markups of the TS Bases are provided in enclosure 3. Changes to the Bases will be implemented in accordance with [LICENSEE's] bases control program. In summary, [LICENSEE] proposes to extend the CT for one inoperable CSS train from 72 hours to 7 days based on Reference 1, and add a Condition to allow one CSS train and one CCS train to be inoperable for up to 72 hours. 3.0 Background The function of the containment heat removal systems under accident conditions is to remove heat from the containment atmosphere, thus maintaining the containment pressure and temperature at acceptably low levels. The systems also serve to limit offsite radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby decreasing the driving force for fission product leakage across the containment. The two containment heat removal systems are the CCS and the CSS. The CCS fan coolers are designed to operate during both normal plant operations and under loss-of-coolant accident [LOCA] or main steam line break
(MSLB)conditions. The CSS is designed to operate during accident conditions only. The heat removal capacity of the CCS and CSS is sufficient to keep the containment temperature and pressure below design conditions for any size break, up to and including a double-ended break of the largest reactor coolant pipe. The systems are also designed to mitigate the consequences of any size break, up to and including a double-ended break of a main stream line. The CCS and CSS continue to reduce containment pressure and temperature and maintain them at acceptable levels post-accident. The CCS and CSS at [PLANT NAME] each consist of [Substitute plant-specific configuration if it differs from the following description] two redundant loops and are designed such that a single failure does not degrade their ability to provide the required heat removal capability. Two of four containment fan coolers and one CSS loop are powered from one safety-related bus. The other two containment fan coolers and CSS loop are powered from another independent safety-related bus. The loss of one bus does not affect the ability of the containment heat removal systems to maintain containment temperature and pressure below the design values in a post-accident mode. The [PLANT NAME] CSS consists of [Substitute plant-specific configuration if it differs from the following description] two independent and redundant loops each containing a spray pump, shutdown heat exchanger, piping, valves, spray headers, and spray nozzles. It has two modes of operation, which are: 1. The injection mode, during which the system sprays borated water from the refueling water tank
(RWT)into the containment, and 2. The recirculation mode, which is automatically initiated by the recirculation actuation signal
(RAS)after low level is reached in the RWT. During this mode of operation, the safety injection system
(SIS)sump provides suction for the spray pumps. Containment spray is automatically initiated by the containment spray actuation signal coincident with the safety injection actuation signal and high containment pressure signal. If required, the operator can manually activate the system from the main control room. Each CSS pump, together with a CCS loop, provides the flow necessary to remove the heat generated inside the containment following a LOCA or MSLB. Upon system activation, the pumps are started and the borated water flows into the containment spray headers. When low level is reached in the RWT, sufficient water has been transferred to the containment to allow for the recirculation mode of operation. Spray pump suction is automatically realigned to the SIS sump upon a RAS. During the recirculation mode, the spray water is cooled by the shutdown heat exchangers prior to discharge into the containment. The shutdown heat exchangers are cooled by the component cooling water system. Post-LOCA pH control is provided by [Substitute plant-specific configuration if it differs from the following description] trisodium phosphate dodecahydrate, which is stored in stainless steel baskets located in the containment near the SIS sump intake. The longer CT for an inoperable CSS train will enhance overall plant safety by avoiding potential unscheduled plant shutdowns and allowing greater availability of safety significant components during shutdown. In addition, this extension provides for increased flexibility in scheduling and performing maintenance and surveillance activities in order to enhance plant safety and operational flexibility during lower modes of operation. 4.0 Technical analysis [LICENSEE] has reviewed References 1 and 2, as well as TSTF-409, Rev. 2, and the model SE published on [DATE] ([] FR []) as part of the CLIIP Notice of Availability. [LICENSEE] has applied the methodology in Reference 1 to develop the proposed TS changes. [LICENSEE] has also concluded that the justifications presented in TSTF-409, Rev. 2 and the model SE prepared by the NRC staff are applicable to [PLANT NAME], and justify this amendment for the incorporation of changes to the [PLANT NAME] TS. In determining the suitability and safety impact of its adoption of TSTF-409, Rev. 2, [LICENSEE] analyzed the effect of increasing the CT for one CSS train to remain out of service using both traditional engineering considerations and probabilistic risk assessment
(PRA)methods. 4.1 Traditional (Deterministic) Engineering Analysis The functions and operation of the CSS and CCS were described in Section 3.0 of this application. Based on a review of the design-basis requirements for the CSS, [LICENSEE] concluded that the loss of one CSS train is well within the design-basis analyses. This conclusion is based on the fact that each CSS pump, together with a CCS loop, provides the flow necessary to remove the heat generated inside the containment following a LOCA or MSLB. Therefore, the combination of one CSS pump and one CCS loop can carry out the design functions of maintaining the containment pressure and temperature at acceptably low levels following a design-basis accident (DBA), and limiting offsite radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby decreasing the driving force for fission product leakage across the containment. The plant status with one CSS train and one CCS train inoperable is covered by TS 3.6.6A, ACTION [D], which states: “[With] one containment spray and one containment cooling train inoperable, restore containment spray train to OPERABLE status within 72 hours, or restore containment cooling train to OPERABLE status within 72 hours.” ACTION [D] ensures that the iodine removal capabilities of the CSS are available, along with 100 percent of the heat removal needs after an accident. The supporting analyses performed in Reference 1 did not evaluate the concurrent inoperabilities of one CSS train and one CCS train, therefore, the current CT of 72 hours is retained in Condition [D]. The 72 hour Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the CSS and CCS, the iodine removal function of the CSS, and the low probability of a DBA occurring during this period. 4.2 Probabilistic Risk Assessment Evaluation [LICENSEE] evaluated the proposed CT extension for the CSS using Reference 3 and Reference 4. This is the same methodology that the NRC staff used in Reference 2. The Key Principles of A Risk-Informed Integrated Decisionmaking Process listed in Reference 3 are as follows: Principle I: The proposed change meets the current regulations. Principle II: The proposed change is consistent with the defense-in-depth philosophy. Principle III: The proposed change maintains sufficient safety margin. Principle IV: When the proposed change results in an increase in core damage frequency or risk, the increase should be small and consistent with the Commission's Safety Goal Policy Statement. Principle V: The impact of the proposed change should be monitored using performance measurement strategies. In Reference 2, the NRC staff found, and [LICENSEE] agrees, that in risk-informed TS CT applications, Principle I is met, since regulations do not require specific CTs, but, rather, require “remedial actions” when an LCO cannot be met. Additionally, in its analysis of Principle III, the NRC staff found, and [LICENSEE] agrees, that the proposed CT extension maintains sufficient safety margins, For [PLANT NAME], the loss of one CSS train is well within the plant's design basis. In Reference 2, the NRC staff determined that the intent of Principles II, IV, and V would be met by a three-tiered approach to evaluate the plant-specific risk impact associated with the proposed TS changes, consistent with the requirements of Reference 4. The first tier evaluates the plant-specific PRA model and the impact of the proposed CT extension on plant operational risk. The second tier addresses the need to preclude potentially high risk configurations by identifying the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration during the time when one CSS train is out of service. The third tier evaluates [LICENSEE'S] proposed Configuration Risk Management Program
(CRMP)to ensure that the applicable plant configuration will be appropriately assessed from a risk perspective before entering into or during the proposed CT. In addition, the NRC staff determined in Reference 2, that the risk analysis methodology and approach used by the CEOG to estimate the risk impact of increasing the CT were reasonable. For most plants that participated in the joint application report, the NRC staff found that the risk impact was shown to be consistent with the acceptance guidelines for change in core damage frequency (ΔCDF), change in large early release frequency (ΔLERF), incremental conditional core damage probability (ICCDP), and incremental conditional large early release probability (ICLERP) specified in References 3 and 4 and Chapters 19.0 and 16.1 of Reference 5. However, not all Combustion Engineering
(CE)plants participated in the joint application report, and the estimated risk impacts for some plans exceeded the Reference 3 and/or Reference 4 acceptance guidelines, which would require additional justifications and/or compensatory measures to be provided for these plants to be determined to have acceptable risk impacts. In addition, the NRC staff found that the Tier 2 and Tier 3 evaluations, as described in Reference 4, could not be approved generically since they were not complete, which would require that each individual plant-specific license amendment seeking adoption of TSTF-409, Rev. 2 would need to include an assessment with respect to the Tier 2 and Tier 3 principles of Reference 4. 4.2.1 Conditions and Supporting Information The following conditions are provided to support adoption of TSTF-409, Rev. 2 by [PLANT NAME]. Responses to the conditions are contained in Attachments 1 through 8 to this application: [NOTE: Licensees who cannot meet the Expectation and Acceptance Criteria listed in these conditions, or choose not to submit the associated information, should not submit an application to adopt TSTF-409, Rev. 2 under the CLIIP.] 1. As shown in Attachment 1, the plant-specific Tier 1 information associated with extending the CSS CT meets the acceptance guidelines of References 3 and 4 associated with ΔCDF, ΔLERF, ICCDP, and ICLERP. [ **EXPECTATIONS/ACCEPTANCE CRITERIA:** the licensee's submittal must provide the ΔCDF, ΔLERF, ICCDP, and ICLERP values related to the CSS 7-day CT and confirm that these values meet the associated acceptance guidelines of References 3 and 4 as no more than a small risk increase (i.e., are in Region II or III of the acceptance guidelines figures). The licensee should utilize an “average maintenance” PRA model for this assessment. If the licensee utilizes a “zero maintenance” PRA model for the assessment, they should state they are using a “zero maintenance” model in the evaluation, and provide a discussion as to the ability of that model to produce comparable results to the “average maintenance” assessment.] 2. As shown in Attachment 2, the technical adequacy (quality) of [PLANT NAME'S] plant-specific PRA is acceptable for this application in accordance with the guidance provided in Reference 3. Specifically, the supporting information addresses the following areas: a. Justification that the plant-specific PRA reflects the as-built, as-operated plant. b. Discussion of plant-specific PRA updates and upgrades since the individual plant examination (IPE), individual plant examination of external events (IPEEE), and subsequent peer reviews and self-assessment. Reference to past submittals discussing this information is acceptable. c. Discussion of plant-specific PRA peer reviews and/or self-assessments performed, their overall conclusions, any facts and observations (F&Os) applicable to this application, and the licensee evaluation and resolution (e.g., by implementing model changes and/or sensitivity studies) of these F&Os to demonstrate the conclusions of the plant-specific analyses for this application are not adversely impacted (i.e., continued acceptability of the proposed extension of the CSS CT). d. Description of the licensee's plant-specific PRA configuration control (quality assurance) program and associated procedures. e. Overall determination of the adequacy of the plant-specific PRA with respect to this application. [ **EXPECTATION:** The licensee's submittal must describe the scope of the plant-specific PRA and must justify its technical adequacy (quality) for this application in accordance with the guidance provided in Reference 3. Specifically, the supporting information must address each area in sufficient detail as shown in the following **ACCEPTANCE CRITERIA:** a. The licensee must provide a justification that confirms that the plant-specific PRA reflects the as-built, as-operated plant. This should include a description of the licensee's data and model update process, and the frequency of these activities. The licensee should also describe how the plant/corporate PRA staff are involved in (and/or made aware of) plant and operational/procedural modifications. b. The licensee must provide a summary description of the plant-specific PRA updates and upgrades since the IPE and peer review of their plant and confirm that the changes identified during the IPEEE have been implemented or otherwise dispositioned. c. The licensee must discuss their plant-specific PRA peer reviews and/or any self-assessments performed (especially noting those conducted per the Nuclear Energy Institute
(NEI)industry peer review guidelines and American Society of Mechanical Engineers
(ASME)PRA Standard), their overall conclusions, any A&B level F&Os applicable to this application, and the licensee's evaluation and resolution (e.g., by implementing model changes and/or sensitivity studies) of these A&B level F&Os to demonstrate the conclusions of the plant-specific analyses for this application are not adversely impacted (i.e., continued acceptability of the proposed extension of the CSS CT). d. The licensee must describe their plant-specific PRA configuration control (quality assurance) program and associated procedures. e. The licensee must make an overall determination of the adequacy of their plant-specific PRA, confirming it is adequate with respect to this application.] 3. Attachment 3 provides supporting information verifying that the plant risk impact associated with external events (e.g., fires, seismic, tornados, high winds, etc.) does not adversely impact or has no impact on the conclusions of the plant-specific analyses for this application and that the overall combined plant CDF and LERF are expected to be within the acceptance guidelines as identified in References 3 and Reference 4 (i.e., total CDF <1E-4/year and total LERF <1E-5/year) [ **EXPECTATIONS:** The licensee's submittal must discuss the plant risks associated with external events and specifically identify (quantitatively and/or qualitatively, as appropriate) that the impact of CSS train CT extension on the risks associated with external events is small. The NRC staff acknowledges that any increase in the external event risk associated with the CSS train CT extension should be minimal. The licensee must address this impact and discuss why the risk overlap with external events (including internal fires) is negligible. Key insights from the IPEEE screening or quantitative approaches may be used to support qualitative arguments. If the licensee has performed updated analyses of an external event since the staff review and acceptance of their IPEEE, and a quantitative PRA demonstration is used to support the submittal, the licensee must describe the significant changes involved in their updated analyses and the impact of these changes on plant risk associated with this external event and with respect to this application. **ACCEPTANCE CRITERIA:** For the NRC staff to conclude the quantified risk associated with the extension request is acceptable, the total CDF and LERF values must meet Reference 3 and Reference 4 acceptance guidelines. For external events for which the licensee has a PRA, and the licensee provides those risk values (i.e., CDF and LERF) and risk metrics (i.e., ΔCDF, ΔLERF, ICCDP, and ICLERP) associated with the specifically analyzed external events, the licensee must also provide the total “at-power” plant risk and total “at-power” change in risk due to all PRA-analyzed contributors (combining internal events, internal flooding, internal fires, and external events. Results may be provided as a summation of values from separate PRA analyses or as a result of an integrated analysis (using a common PRA model for all contributors) or a combination of the above. For external events for which the licensee does not have a PRA (and it is not screened out as above), but rather relies on a non-PRA method (e.g., seismic margins analysis
(SMA)or fire-induced vulnerability evaluation (FIVE)), to determine if the plant risk is acceptable, the licensee must confirm for this application that there were and still are either no vulnerabilities or outliers associated with these external events, or confirm that any vulnerabilities or outliers that were identified in their documented analyses (most likely in their IPEEE) have been resolved and, as needed, the appropriate plant/procedural modifications have been implemented as described in their documented analyses.] 4. Supporting information is provided in Attachment 4, consistent with the evaluation summary and conclusions (Sections 7 and 8) provided in Reference 2, in which licensees discuss implementation of procedures that prohibited entry into a 7-day CSS CT for scheduled maintenance purposes if external event conditions or warnings (e.g., severe weather warnings for ice, tornados, high winds, etc.) are in effect or confirm that these external events do not impact the submittal. [LICENSEE'S] discussion confirms that [PLANT NAME'S] procedures include compensatory measures and normal plant practices that help avoid potentially high risk configurations during the proposed extension of the CSS CT. This supporting information must also address the Tier 2 aspects of Reference 4. [ **EXPECTATIONS:** The licensee's submittal must discuss (including licensee commitments related to) implementation of procedures that prohibit entry into a 7-day CSS CT for scheduled maintenance purposes if external event conditions or warnings are in effect. If the licensee does not want to implement this prohibition for specific severe weather conditions or warnings, the licensee must explicitly identify these event conditions/warnings and provide a justification for not including them. If there are no risk significant configurations or risk significant external event conditions identified in the Tier 2 evaluation, then the licensee should include a statement that there are no risk significant configurations that would preclude them from using a 7-day CT. The licensee must also confirm that its procedures include compensatory measures and normal plant practices that help avoid potentially high risk configurations during the proposed extension of the CSS train CT. This supporting information must also address the Tier 2 aspects of Reference 4. The Tier 2 evaluation is meant to be an early evaluation (at the license submittal stage) to identify and preclude potentially high-risk plan configurations that could result if equipment, in addition to that associated with the proposed license amendment, is taken out of service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved. **ACCEPTANCE CRITERIA:** The Tier 2 evaluation needs to identify, as part of the licensee's submittal, potentially high-risk plant configurations associated with the CSS train CT extension that need to be precluded, if any, and identify how this is implemented (i.e., typically these aspects result in licensees establishing compensatory measures/commitments to ensure these configurations are precluded). If, in conducting the evaluation, the licensee identifies no high-risk plant configurations, then the licensee needs to explicitly state this fact.] 5. Attachment 5 provides supporting information, consistent with the evaluation summary and conclusions (Sections 7 and 8) provided in Reference 2, that describes the plant-specific risk-informed CRMP to assess the risk associated with the removal of equipment from service during the 7-day CSS CT. If the licensee utilizes the Maintenance Rule (a)(4) program to evaluate the risk significance of configurations, it should state so in its submittal. In this description, [LICENSEE] confirms that the program provides the necessary assurances that appropriate assessments of plant risk configurations are sufficient to support the proposed CSS CT extension request. This supporting information also addresses the Tier 3 aspects of Reference 4. [ **EXPECTATIONS/ACCEPTANCE CRITERIA:** The licensee's submittal must describe its CRMP or associated (a)(4) program (as appropriate), including how it reflects the current plant PRA model (specifically identifying any deviations and simplifications in the CRMP model from the plant-specific PRA model) and how the CRMP is updated to remain consistent with the plant-specific PRA. The licensee's submittal must also describe how the CRMP or associated (a)(4) program provides the necessary assurances that appropriate assessments of plant risk configurations are sufficient to support the proposed CT extension request for the CSS. Finally, the licensee's submittal must address the Tier 3 aspects of Reference 4, including he description of the CRMP, and must confirm that its CRMP or associated (a)(4) program meets all aspects of Section 2.3.7.2 of Reference 4, specifically describing how its program meets each of the four Key Components identified in this Section. The Tier 3 evaluation ensures that the CRMP or associated (a)(4) program is adequate when maintenance is about to commence, as opposed to the early (submittal stage) evaluation performed for Tier 2.] 6. Attachment 6 provides supporting information, consistent with the evaluation summary (Section 7) provided in Reference 2, describing the relationship between components of the CSS and the shutdown cooling system (SDCS). For plants where components of the two systems may be used as backup to the other, the licensee must either confirm that Tier 2 conditions exist in the licensee's CRMP or associated (a)(4) program that will not allow “at power” maintenance of the CSS and SDCS at the same time or that the risk significance of such maintenance configurations is low. If the CSS and SDCS have backup components, the plant should also describe how this backup capability is considered as part of the plant's shutdown operations program (SOP). If this backup feature is not considered when one train of the SDCS is in maintenance or otherwise unavailable, it should be stated in the licensee's application. [ **EXPECTATION:** The licensee's submittal must describe the relationship/interfaces between the CSS and SDCS. **ACCEPTANCE CRITERIA:** If the SDCS can be used as a backup to the CSS, then the licensee must confirm that “at power” maintenance of the CSS and SDCS will not be allowed at the same time and describe how this is controlled (e.g., specifically identified in the CRMP as a configuration that is not allowed) or provide justification that the risks associated with a simultaneous “at-power” outage of one SDCS train and one CSS train is small. If the SDCS cannot be used (and is not credited) as a backup to CSS, then the licensee needs to explicitly state this fact. If CSS pumps can be used as a backup to the SDCS pumps, then the licensee must confirm that at least one CSS pump is required to be operable when maintenance of the CSS is performed in lower modes of operation (consistent with the plant's Technical Specifications) and must describe how this is controlled or demonstrate that the SOP provides adequate risk management for that configuration. If CSS pumps cannot be used (and are not credited) as a backup to SDCS pumps in lower modes of operation, then the licensee needs to explicitly state this fact.] 7. Attachment 7 provides supporting information confirming that the licensee's Maintenance Rule program includes the ability to compute ICDP (incremental core damage probability), and ILERP (incremental large early release probability). [ **EXPECTATIONS/ACCEPTANCE CRITERIA:** The licensee must confirm that their CRMP quantitative model (e.g., model used to provide quantitative assessments in support of 10 CFR 50.65 (a)(4)) calculates ICDP and ILERP, and that their CRMP quantitative model (e.g., model used to provide quantitative assessment in support of 10 CFR 50.65 (a)(4)) explicitly models the CSS or has been modified to include the CSS, which will be used whenever CSS components are made unavailable. 8. Attachment 8 provides information addressing how plant-specific systems, structures and components
(SSC)are monitored and assessed at the plant under the Maintenance Rule (i.e. 10 CFR 50.65). Maintenance Rule unavailability and unreliability targets for CSS are also provided. These targets will be monitored in accordance with provisions of the Maintenance Rule. [ **EXPECTATIONS/ACCEPTANCE CRITERIA:** The licensee must describe how plant-specific SSC reliability and availability are monitored and assessed at the plant under the Maintenance Rule (i.e., 10 CFR 50.65) to confirm that performance continues to be consistent with the analyses used to justify the 7-day CT. In providing this description, the licensee should also indicate how it periodically assesses previous risk-informed licensing action decisions to ensure that these decisions remain valid (i.e., continue to meet the Reference 3 and Reference 4 acceptance guidelines) for the current plant operations and plant-specific PRA and what actions it takes if a previously-approved risk-informed licensing action decision is determined to no longer meet these acceptance guidelines.] 4.2.2 Regulatory Commitment The Reference 4 Tier 3 program ensures that, while the plant is following the TS ACTIONS associated with a 7-day CT for restoring an inoperable CSS to operable status, additional activities will not be performed that could further degrade the capabilities of the plant to respond to a condition that the inoperable CSS is designed to mitigate and, as a result, increase plant risk beyond that determined by the Reference 1 analyses. [LICENSEE's] implementation of Reference 4 Tier 3 guidelines generally implies the assessment of risk with respect to CDF. However, the proposed CSS 7-day CT impacts accident sequences that can be mitigated following core damage and, consequently, impacts LERF as well as CDF. Therefore, [LICENSEE] has enhanced its CRMP, [OPTIONAL: as implemented under 10 CFR 50.65(a)(4), the Maintenance Rule,] to include a LERF assessment to support this application. 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration [LICENSEE] has reviewed the proposed no significant hazards consideration determination published in the **Federal Register** on [DATE] ([ ] FR [ ]) as part of the CLIP. [LICENSEE] has concluded that the proposed determination presented in the notice is applicable to [PLANT NAME] and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a). 5.2 Applicable Regulatory Requirements/Criteria Based on its answers to the Section 4.2.1 questions provided in Attachments 1 through 8 to this application [LICENSEE] determines is based on the following: 1. The traditional engineering evaluation reveals that the loss of one CSS train is well within [PLANT NAME's] design basis analyses. Key principles 1,2,3, and 5 in Section 2 of Reference 3 are met. 2. By meeting the conditions identified in Section 4.2.1, [LICENSEE] believes that its PRA model is acceptable for this application and also concludes that there is minimal impact of the CT extensions for the CSS system on plant operational risk (Tier 1 evaluation). 3. By meeting the conditions identified in Section 4.2.1, [LICENSEE] will ensure that its implementation will identify potentially high risk configurations and the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration (Tier 2 evaluation), or state that no Tier 2 limitations have been identified. 4. By meeting the conditions identified in Section 4.2.1, [PLANT NAME] will ensure that its risk-informed CRMP will satisfactorily assess the risk associated with the removal of equipment from service during the proposed CSS CT (Tier 3 evaluation) and the CRMP and plant risk will be managed by plant procedures, including implementation and monitoring of SSCs (CSS). In conclusion, based on the consideration discussed above,
(1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner,
(2)such activities will be conducted in compliance with the Commission's regulations, and
(3)the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 6.0 Environmental Consideration [LICENSEE] has reviewed the environmental evaluation included in the model safety evaluation as pat of the CLIIP. [LICENSEE] concluded that the staff's findings presented in that the evaluation are applicable to [PLANT NAME] and the evaluation is hereby incorporated by reference for this application. 7.0 References [Licensee should include an applicable list of references, including but not limited to] 1. Joint Applications Report: Modification to the Containment Spray System, and Low Pressure Safety Injection System Technical, CE Owners Group, CE NPSD-1045, March 2000. 2. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to CE Owners Group CE-NPSD-1045, “Joint Application Report, Modification to the Containment Spray System, and the Low Pressure Safety Injection System Technical Specifications, December 21, 1999.” 3. USNRC Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 1, November 2002. 4. USNRC Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” August 1998. 5. NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” June 1996. PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) Enclosure 2 CHANGES TO TS BASES Enclosure 3 CONDITION
(1)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 1 CONDITION
(2)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 2 CONDITION
(3)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 3 CONDITION
(4)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 4 CONDITION
(5)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 5 CONDITION
(6)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 6 CONDITION
(7)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 7 CONDITION
(8)[LICENSEE'S] EVALUATION AND SUPPORTING INFORMATION Attachemnt 8 MODEL SAFETY EVALUATION U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Consolidated Line Item Improvement Technical Specification Task Force TSTF-409, Revision 2 “Containment Spray System Completion Time Extension” 1.0 Introduction By letter to the Nuclear Regulatory Commission (NRC, Commission) dated [DATE] (Agencywide Documents Access and Management System (ADAMS) Accession Number MLXXXXXXXXX), [LICENSEE] (the licensee) requested changes to the Technical Specifications
(TSs)for [PLANT NAME]. The proposed changes would revise TS 3.6.6A, “Containment Spray and Cooling Systems,” by extending from 72 hours to seven days the completion time
(CT)to restore an inoperable containment spray system
(CSS)train to operable status, and would add a Condition describing the required action and CT when one CSS train and one containment cooling system
(CCS)train are inoperable. The changes are based on Technical Specification Task Force
(TSTF)Change Traveler, TSTF-409, Revision 2 (Rev.), “Containment Spray System Completion Time Extension (CE NPSD-2045-A)” and associated TS Bases. TSTF-409, Rev. 2, submitted to the NRC by the TSTF in a letter dated November 10, 2003 (ADMS Accession Number MLO33280006), was approved by the NRC on [DATE]. TSTF-409, Rev. 2 is based on Combustion Engineering Owner's Group
(CEOG)Joint Application Report CE NPSD-1045-A, “Joint Applications Report for Modifications to the Containment Spray System Technical Specifications,” dated March 2000 (Reference 1), as accepted by, and subject to the limitations specified in, the associated NRC safety evaluation (SE), dated December 212, 1999 (ADMS Accession Number ML993620241) (Reference 2). In TSTF-409, Rev. 2, the CEOG states that the longer CT for restoring an inoperable CSS train to operable status will enhance overall plant safety by avoiding potential unscheduled plant shutdowns and allowing greater availability of safety significant components during shutdown. In addition the CEOG states that this extension provides for increased flixibility in scheduling and performing maintenance and surveillance activities in order to enhance plant safety and operational flexibility during lower modes of operation. 2.0 Regulatory Evaluation Since the mid-1980's, the NRC has been reviewing and granting improvements to TS that are based, at least in part, on probabilistic risk assessment
(PRA)insights. In its final policy statement on TX improvements dated July 22, 1993 (58 FR 39132), the NRC stated that it: * * * expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PSA [probabilistic safety assessment] 1 or risk survey and any available literature on risk insights and PSAs * * * Similarly, the NRC staff will also employ risk insights an PSAs in evaluating Technical Specifications related submittals. Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements. 1 PSA and PRA are used interchangeably herein. The NRC reiterated this point when it issued the revision to 10 CFR 50.36, “Technical Specifications,” in July 1995. In August 1995, the NRC adopted a final policy statement on the use of PRA methods in nuclear regulatory activities that encouraged greater use of PRA to improve safety decision-making and regulatory efficiency. The PRA policy statement included the following points: 1. The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data, and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy. 2. PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters; where practical within the bounds of the state- of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements. 3. PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review. In March 1998, the CEOG submitted a joint applications report for the NRC staff's review entitled, “Joint Applications Report for Modifications to the Containment Spray System and Low Pressure Safety System Technical Specifications.” The NRC review accepting this joint applications report for referencing in license applications for Combustion Engineering
(CE)plants, including appropriate exclusions, conditions, and limitations, is documented in Reference 2. The final, NRC-approved joint applications report, (Reference 1) is dated March 2000. 3.0 Technical Evaluation The NRC staff evaluated the licensee's proposed amendment to extend the TS CT for one CSS train out of service from 72 hours to seven days using insights derived from traditional engineering considerations and the use of PRA methods to determine the safety impact of extending the CT. 3.1 Traditional Engineering Evaluation The function of the containment heat removal systems under accident conditions is to remove heat from the containment atmosphere, thus maintaining the containment pressure and temperature at acceptably low levels. The systems also serve to limit offsite radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby decreasing the driving force for fission product leakage across the containment. The two containment heat removal systems are the CCS and CSS. The CCS fan coolers are designed to operate during both normal plant operations and under loss-of-coolant accident
(LOCA)or main stream line break
(MSLB)conditions. The CSS is designed to operate during accident conditions only. The heat removal capacity of the CCS and CSS is sufficient to keep the containment temperature and pressure below design conditions for any size break, up to and including a double-ended break of the largest reactor coolant pipe. The systems are also designed to mitigate the consequences of any size break, up to and including a double-ended break of a main stream line. The CCS and CSS continue to reduce containment pressure and temperature and maintain them at acceptable levels post-accident. The CCS and CSS at [PLANT NAME] each consist of [Substitute plant-specific configuration if it differs from the following description] two redundant loops and are designed such that a single failure does not degrade their ability to provide the required heat removal capability. Two of four containment fan coolers and one CSS loop are powered from one safety-related bus. The other two containment fan coolers and one CSS loop are powered from another independent safety related bus. The loss of one bus does not affect the ability of the containment heat removal systems to maintain containment temperature and pressure below the design values in a post-accident mode. The [PLANT NAME] CSS consists of [Substitute plant-specific configuration if it differs from the following description] two independent and redundant loops each containing a spray pump, shutdown heat exchanger, piping, valves, spray headers, and spray nozzles. It has two modes of operation, which are: 1. The injection mode, during which the system sprays borated water from the refueling water tank
(RWT)into the containment, and 2. The recirculation mode, which is automatically initiated by the recirculation actuation signal
(RAS)after low level is reached in the RWT. During this mode of operation, the safety injection system
(SIS)sump provides suction for the spray pumps. Containment spray is automatically initiated by the containment spray actuation signal coincident with the safety injection actuation signal and high containment pressure signal. If required, the operator can manually activate the system from the main control room. Each CSS pump, together with a CCS loop, provides the flow necessary to remove the heat generated inside the containment following a LOCA or MSLB. Upon system activation, the pumps are started, and borated water flows into the containment spray headers. When low level is reached in the RWT, sufficient water has been transferred to the containment to allow for the recirculation mode of operation. Spray pump suction is automatically realigned to the SIS sump upon a RAS. During a recirculation mode, the spray water is cooled by the shutdown heat exchangers prior to discharge into the containment. The shutdown heat exchangers are cooled by the component cooling water system. Post-LOCA pH control is provided by [Substitute plant-specific configuration if it differs from the following description] trisodium phosphate dodecahydrate, which is stored in stainless steel baskets located in the containment near the SIS sump intake. Based on a review of the design-basis requirements for the CSS, the NRC staff concluded that the loss of one CSS train is well within the design-basis analyses. The plant status with one CSS train and one CCS train inoperable is covered by TS3.6.6A, ACTION D, which states: “[With] one containment spray and one containment cooling train inoperable, restore containment spray train to OPERABLE status within 72 hours, or restore containment cooling train to OPERABLE status within 72 hours.” ACTION D ensures that the iodine removal capabilities of the CSS are available, along with 100 percent of the heat removal needs after an accident. The supporting analyses performed in Reference 1 did not evaluate the concurrent inoperabilities of one CSS train and one CCS train. Therefore, the current CT of 72 hours is retained in Condition D. The 72-hour CT was development taking into account the redundant heat removal capabilities afforded by combinations of the CSS and CCS, the iodine removal function of the CSS, and the low probabilities of a DBA occurring during this period. 3.2 Probabilistic Risk Assessment Evaluation The proposed extension of the CSS CT for one inoperable train from 72 hours to seven days affects plant risk by impacting: 1. Accident sequences that can be prevented from leading to core damage. 2. Accident sequences that can be mitigated following core damage. The CSS therefore affects both core damage frequency
(CDF)and large early release frequency (LERF). This is because the CSS performs the critical function of controlling containment temperature and pressure to cool the reactor coolant system
(RCS)inventory that is spilled in the sump as a result of a LOCA (core damage prevention role) and preventing the release of radionuclides subsequent to a core damage event (core damage and radionuclide release mitigation role). [The following paragraph will contain plant-specific information based on the plant's ability to use the shutdown cooling system
(SDCS)as a backup to the CSS. The licensee should provide a plant-specific system configuration description based on whether its SDCS can be used a backup to the CSS pump.] The proposed CT extension also impacts the long-term cooling function that can be provided by the SDCS following a small-break LOCA, steam generator tube rupture (SGTR), or MSLB. If entry into the 7-day CT is caused by a CSS pump outage, the plants with the ability to use the SDCS as a backup to the CSS pump can still preserve the spray function of the affected train. If, however, a SDCS heat exchanger is removed from service, then both the CSS and SDCS capability of the affected train would be lost unless cross-connect capability with another unaffected system (e.g., service water) is possible. However, this cross-connect capability should not be credited unless it is proceduralized. The NRC staff used a three-tiered approach to evaluate the plant-specific risk impact associated with the proposed TS changes. The first tier evaluates the plant-specific PRA model and the impact of the proposed CT extension on plant operational risk. The second tier addresses the need to preclude potentially high risk configurations by identifying the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration during the time when on CSS train is out of service. The third tier evaluates the licensee's proposed Configuration Risk Management Program
(CRMP)to ensure that the applicable plant configuration will be appropriately assessed from a risk perspective before entering into, or during, the proposed CT. In Reference 2, the NRC staff found that the risk analysis methodology and approach used by the CEOG to estimate the risk impact were reasonable. In its SE, the NRC staff also stated that, for most plants that participated in the joint application report, the risk impact can be shown to be consistent with the acceptance guidelines for change in CDF (ΔCDF), change in LERF (ΔLERF), incremental conditional core damage probability (ICCDP), and incremental large early release frequency (ICLERP) specified in Regulatory Guide
(RG)1.174 (Reference 3) and RG 1.177 (Reference 4) and the associated Standard Review Plan
(SRP)Chapters 19.0 and 16.1 of NUREG-0800 (Reference 5). However, not all CE plants participated in the joint application report, and the estimated risk impacts for some plants exceeded the Reference 3 and/or Reference 4 acceptance guidelines, which would require additional justifications and/or compensatory measures to be provided for these plants to be determined to have acceptable risk impacts. In Reference 2, the NRC staff also found that the Tier 2 and Tier 3 evaluations, as described in Reference 4, could not be approved generically since they were not complete from the perspective of addressing plant-specific Tier 2 and Tier 3 issues which would require that each individual plant-specific license amendment seeking approval through TSTF-409, Rev. 2 would need to include an assessment with respect to the Tier 2 and Tier 3 principles of Reference 4. Based on the above discussion, the NRC staff identified conditions that must be addressed in the licensee's plant-specific application requesting adoption of TSTF-409, Revision 2. In its application dated [DATE], the licensee provided supporting information for each of the conditions which met the NRC staff's expectations and acceptance criteria [with the following exceptions: list any exceptions to the conditions stated in the model LAR]. [Provide a discussion of any significant plant-specific exceptions to or modifications of the conditions described in the model LAR]. 3.2.1 Commitment The Reference 4 Tier 3 program ensures that, while the plant is following the TS ACTIONS associated with a 7-day CT for restoring an inoperable CSS to operable status, additional activities will not be performed that could further degrade the capabilities of the plant to respond to a condition that the inoperable CSS is designed to mitigate and, as a result, increase plant risk beyond that determined by the Reference 1 analyses. A licensee's implementation of Reference 4 Tier 3 guidelines indicates that it has assessed risk with respect to CDF. However, the proposed CSS 7-day CT impacts accident sequences that can be mitigated following core damage and, consequently, LERF as well as CDF. Therefore, the licensee enhnaced its CRMP [ *optional:* as implemented under 10 CFR 50.65(a)(4), the Maintenance Rule,] to include a LERF assessment. [ *The licensee should confirm that performance of LERF assessments is included in the plant's Maintenance Rule program.* ] 3.3 Summary On [DATE], ([ ] FR [ ]), the NRC announced the availability of TSTF-409, Rev. 2 for adoption by licensees using the consolidated line item improvement program (CLIIP). In its model license amendment request (LAR), the NRC staff asked each licensee to verify several aspects of its plant-specific PRA program including: 1) verification of PRA quality, 2) plant-specific analyses of the impact of this TS change on overall risk, 3) Maintenance Rule and CRMP considerations associated with the proposed changes, and, 4) system interdependencies. In its [DATE] submittal, the licensee provided satisfactory information related to the eight conditions and one licensee commitments set forth in the model LAR. Having met the conditions identified in the model LAR, the NRC staff finds that the licensee's plant-specific LAR is consistent with the previous NRC staff approval of Reference 1, as documented in Reference 2 and TSTF-409, Rev. 2, and thus is acceptable. This determination is based on the following: 1. The traditional engineering evaluation reveals that the loss of one CSS train is well within the design-basis analyses. 2. Since the licensee meets the conditions identified in the model LAR, the NRC staff finds that there is minimal impact of the CT extensions for the CSS system on plant operational risk (Tier 1 evaluation). 3. Meeting the conditions identified in the model LAR will ensure that the licensee's implementation will identify potentially high risk configurations and the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration (Tier 2 evaluation). 4. Meeting the conditions identified in the model LAR will ensure that the risk-informed CRMP proposed by the licensee will satisfactorily assess the risk associated with the removal of equipment from service during the proposed CSS CT (Tier 3 evaluation) and the CRMP and plant risk will be managed by plant procedures. 4.0 Regulatory Commitment The licensee's letter dated [DATE], contained the following regulatory commitment: [STATE THE LICENSEE'S COMMITMENT AND ENSURE THAT IT SATISFIES THE COMMITMENT IN SECTION 3.2.1 OF THIS SE]. The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment are best provided by the licensee's administrative controls process, including its commitment management program. The above regulatory commitment does not warrant the creation of a license condition (item requiring prior NRC approval of subsequent changes). 5.0 State Consultation In accordance with the Commission's regulations, the [STATE] State official was notified of the proposed issuance of the amendment[s]. The State official had [CHOOSE ONE:
(1)No comments, OR
(2)the following comments—with subsequent disposition by the staff]. 6.0 Environmental Consideration The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding [(XX FR XXXXX, dated Monthly DD, YYYY)]. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment. 7.0 Conclusion The Commission has concluded, based on the considerations discussed above, that
(1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner,
(2)such activities will be conducted in compliance with the Commission's regulations, and
(3)the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 8.0 References 1. *Joint Applications Report:* Modification to the Containment Spray System, and Low Pressure Safety Injection System Technical, CE Owners Group, CE NPSD-1045, March 2000. 2. SE by the Office of Nuclear Reactor Regulation Related to CE Owners Group CE-NPSD-1045, “Joint Application Report, Modification to the Containment Spray System, and the Low Pressure Safety Injection System Technical Specifications,” December 21, 1999. 3. U.S. NRC RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 1, November 2002. 4. U.S. NRC RG 1.177, “An Approach for Plant-Specific, Risk-Informed *Decisionmaking:* Technical Specifications,” August 1998. 5. NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” June 1996. Model No Significant Hazards Consideration *Description of Amendment Request:* The proposed amendment would revise the technical specifications to extend the completion time
(CT)from 72 hours to seven days to restore an inoperable containment spray system
(CSS)train to operable status, and add a Condition describing the required Actions and CT when one CSS and one containment cooling system
(CCS)are inoperable. *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 an accident previously evaluated? *Response:* No. The proposed change extends from 72 hours to 7 days the CT for restoring an inoperable CSS train to operable status. Being in an ACTION 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 ACTIONS during the 7-day CT are no different than the consequences of an accident while relying on the ACTION during the existing 72-hour CT. 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 change create the possibility of a new or different kind of accident from any accident previously evaluated? *Response:* No. The proposed change extends from 72 hours to 7 days the CT for restoring an inoperable CSS train to operable status. 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 operation. 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 extends from 72 hours to 7 days the CT for restoring an inoperable CSS train to operable status. The licensee performed risk-based evaluations using its plant-specific probabilistic risk assessment
(PRA)model in order to determine the effect of this change on plant risk. The PRA evaluations were based on the conditions stipulated in NRC staff safety evaluations approving both Joint Applications Report CE NPSD-1045-A, “Joint Applications Report, Modifications to the Containment Spray System and The Low Pressure Safety Injection System Technical Specifications,” and Technical Specification Task Force Change Traveler, TSTF-409, Revision 2, “Containment Spray System Completion Time Extension (CE NPSD-1045-A).” The results of these plant-specific evaluations determined that the effect of the proposed change on plant risk is very small. Therefore, this change does not involve a significant reduction in a margin of safety. Based on the above, the proposed change involves 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. Dated at Rockville, Maryland, this XX day of XXXXXXXX, 2006. FOR THE NUCLEAR REGULATORY COMMISSION Project Manager Plant Licensing Branch [ ] Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation [FR Doc. 06-9094 Filed 11-6-06; 8:45 am]
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CFR
- Issuance of amendment.§ 50.92
- Hearing requests, petitions to intervene, requirements for standing, and contentions.§ 2.309
- Notice for public comment; State consultation.§ 50.91
- Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.§ 50.46
- Changes, tests, and experiments.§ 50.59
- Technical specifications.§ 50.36
- Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review.§ 51.22
- Application of subpart to ongoing environmental work.§ 51.12
- Application for amendment of license, construction permit, or early site permit.§ 50.90
- Requirements for monitoring the effectiveness of maintenance at nuclear power plants.§ 50.65
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- 10 CFR 2
- 10 CFR 20
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