Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement for B&W Reactor Plants To Risk-Inform Requirements Regarding Selected Required Action End-States Using the Consolidated Line Item Improvement Process, 65615-65629 [E7-22738]

Download as PDF Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2: The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. No new or different accidents result from utilizing the proposed change. The proposed change permits physical alteration of the plant involving removal of the CAD system. The CAD system is not an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building from any design basis event. The changes to the TS do not alter assumptions made in the safety analysis, but reflect changes to the design requirements allowed under the revised 10 CFR 50.44. The proposed change is consistent with the revised safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3: The proposed change does not involve a significant reduction in a margin of safety. The Commission has determined that the DBA LOCA hydrogen release is not risk significant, therefore is not required to be analyzed in a facility accident analysis. The proposed change reflects this new position and, due to remaining plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, including postulated beyond design basis events, does not result in a significant reduction in a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, the NRC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of ‘‘no significant hazards consideration’’ is justified. pwalker on PROD1PC71 with NOTICES [FR Doc. E7–22740 Filed 11–20–07; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement for B&W Reactor Plants To Risk-Inform Requirements Regarding Selected Required Action End-States Using the Consolidated Line Item Improvement Process Nuclear Regulatory Commission. ACTION: Request for comment. AGENCY: SUMMARY: Notice is hereby given that the staff of the Nuclear Regulatory Commission (NRC) has prepared a model safety evaluation (SE) and model license amendment request (LAR) relating to changes to the end-state requirements for required actions in B&W reactor plants’ technical specifications (TS). Current technical specification action requirements frequently require that the unit be brought to cold shutdown when the technical specification limiting condition for operation for a system has not been met. Depending on the system, and the affected safety function, the requirement to go to cold shutdown may not represent the most risk effective course of action. In accordance with a qualitative risk analysis that provides a basis for changes to the action requirement to shutdown, where appropriate the shutdown end-state is changed from cold shutdown to hot shutdown. The affected TS are: 3.3.5 Engineered Safety Feature Actuation System (ESFAS) Instrumentation. 3.3.6 ESFAS Manual Initiation. 3.4.6 Reactor Coolant System (RCS) Loops—MODE 4. 3.4.15 RCS Leakage Detection Instrumentation. 3.5.4 Borated Water Storage Tank (BWST). 3.6.2 Containment Air Locks. 3.6.3 Containment Isolation Valves. 3.6.4 Containment Pressure. 3.6.5 Containment Air Temperature. 3.6.6 Containment Spray and Cooling Systems. 3.7.7 Component Cooling Water System. 3.7.8 Service Water System. 3.7.9 Ultimate Heat Sink. 3.7.10 Control Room Emergency Ventilation System (CREVS). 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS). 3.8.1 AC Sources—Operating. 3.8.4 DC Sources—Operating. 3.8.7 Inverters—Operating. 3.8.9 Distribution Systems—Operating. The NRC staff has also prepared a model no significant hazards consideration (NSHC) determination relating to this matter. The purpose of VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 PO 00000 Frm 00059 Fmt 4703 Sfmt 4703 65615 these models is to permit the NRC to efficiently process amendments that propose to adopt technical specification changes, designated as TSTF–431, Revision 2, related to Topical Report BAW–2441, Revision 2, ‘‘Risk Informed Justification for LCO End-State Changes,’’ September 2006. Licensees of B&W nuclear power reactors to which the models apply could then request amendments utilizing the models and justifying the applicability of the SE and NSHC determination to their reactors. The NRC staff is requesting comments on the model SE, model LAR, and model NSHC determination prior to announcing their availability for referencing in license amendment applications. DATES: The comment period expires December 21, 2007. Comments received after this date will be considered if it is practical to do so, but the Commission is able to ensure consideration only for comments received on or before this date. ADDRESSES: Comments may be submitted either electronically or via U.S. mail. Submit written comments to Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, Mail Stop: T–6 D59, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001. Hand deliver comments to: 11545 Rockville Pike, Rockville, Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays. Copies of comments received may be examined at the NRC’s Public Document Room, 11555 Rockville Pike (Room O– 1F21), Rockville, Maryland. Comments may be submitted by electronic mail to CLIIP@nrc.gov. FOR FURTHER INFORMATION CONTACT: Timothy Kobetz, Mail Stop: O–12H2, Technical Specifications Branch, Division of Inspection & Regional Support, 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 for Adopting Standard Technical Specification Changes for Power Reactors,’’ was issued on March 20, 2000. The consolidated line item improvement process (CLIIP) is intended to improve the efficiency of NRC licensing processes, by processing proposed changes to the standard technical specifications (STS) in a manner that supports subsequent license amendment applications. The E:\FR\FM\21NON1.SGM 21NON1 65616 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pwalker on PROD1PC71 with NOTICES CLIIP includes an opportunity for the public to comment on proposed changes to the STS after a preliminary assessment by the NRC staff and finding that the change will likely be offered for adoption by licensees. The CLIIP directs the NRC staff to evaluate any comments received for a proposed change to the STS and to either reconsider the change or announce the availability of the change for adoption by licensees. Licensees opting to apply for this TS change are responsible for reviewing the staff’s evaluation, referencing the applicable technical justifications, and providing any necessary plant-specific information. 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 solicits comment on changes to the end-state requirements for required actions, if risk is assessed and managed, for the primary purpose of accomplishing short-duration repairs which necessitated exiting the original Mode of operation. The change was proposed in Topical Report BAW–2441, Revision 2, ‘‘Risk Informed Justification for LCO End-State Changes,’’ September 2006. This change was proposed for incorporation into the standard technical specifications by the owners groups participants in the Technical Specification Task Force (TSTF) and is designated TSTF–431, Revision 2. TSTF–431, Revision 2, can be viewed on the NRC’s web page at http:// www.nrc.gov/reactors/operating/ licensing/techspecs.html. Applicability This proposal to modify technical specification requirements by the adoption of TSTF–431, Revision 2, is applicable to all licensees of B&W plants. To efficiently process the incoming license amendment applications, the staff requests that each licensee applying for the changes proposed in TSTF–431, Revision 2, include Bases for the proposed TS consistent with the Bases proposed in TSTF–431, Revision 2. To efficiently process the incoming license amendment applications, the staff requests that each licensee applying for the changes proposed in TSTF–431, Revision 2, use the CLIIP. Licensees are not prevented from requesting an alternative approach or proposing the changes without the requested Bases and Bases control program. Variations from the approach recommended in this notice may require additional review by the NRC staff, and may increase the time and resources needed for the review. Significant variations from the VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 approach, or inclusion of additional changes to the license, will result in staff rejection of the submittal. Instead, licensees desiring significant variations and/or additional changes should submit a LAR that does not claim to adopt TSTF–431, Revision 2. Public Notices This notice requests comments from interested members of the public within 30 days of the date of publication in the Federal Register. After evaluating the comments received as a result of this notice, the staff will either reconsider the proposed change or announce the availability of the change in a subsequent notice (perhaps with some changes to the SE, LAR, or the proposed NSHC determination as a result of public comments). If the staff announces the availability of the change, licensees wishing to adopt the change must submit an application in accordance with applicable rules and other regulatory requirements. For each application, the staff will publish a notice of consideration of issuance of amendment to facility operating licenses, a proposed NSHC determination, and a notice of opportunity for a hearing. The staff will also publish a notice of issuance of an amendment to operating license to announce the modification of end-state requirements for required actions in plant technical specifications. Proposed Model Plant Specific Safety Evaluation for Technical Specification Task Force (TSTF) Change TSTF–431, Revision 2, Change in Technical Specifications End-States (BAW–2441), a Consolidated Line Item Improvement U.S. NUCLEAR REGULATORY COMMISSION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. [lll] TO FACILITY OPERATING LICENSE NFP[lll] [UTILITY NAME] [PLANT NAME], [UNIT lll] DOCKET NO. -[lll] 1.0 Introduction By letter dated llllll, 20ll, [Utility Name] (the licensee) proposed changes to the technical specifications (TS) for [plant name]. The requested changes are the adoption of TSTF–431, Revision 2, to the B&W Reactor Standard Technical Specifications (STS) (NUREG–1430), which was proposed by the Technical Specifications Task Force (TSTF) on July 13, 2007, on behalf of the industry. TSTF–431, Revision 2, incorporates the B&W Owners Group (B&WOG) approved Topical Report BAW–2441, Revision 2, ‘‘Risk Informed PO 00000 Frm 00060 Fmt 4703 Sfmt 4703 Justification for LCO End-State Changes,’’ September 2006, (Reference 1), into the B&W STS (Note: The changes are made with respect to Revision 3 of the STS NUREGs). TSTF–431, Revision 2, is one of the industry’s initiatives developed under the Risk Management Technical Specifications (RMTS) program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and management techniques in TS, while reducing unnecessary burden and making TS requirements consistent with the Commission’s other risk-informed regulatory requirements, in particular the maintenance rule. The Code of Federal Regulations, 10 CFR 50.36, ‘‘Technical Specifications,’’ states: ‘‘When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow the remedial action permitted by the technical specification until the condition can be met.’’ The STS and many plant TS provide a completion time (CT) for the plant to meet the limiting condition for operation (LCO). If the LCO or the remedial action cannot be met, then the reactor is required to be shut down. When the STS and individual plant technical specifications were written, the shutdown condition or end-state specified was usually cold shutdown. Topical Report BAW–2441, Revision 2, provides the technical basis to change certain required end-states when the TS Actions for remaining in power operation cannot be met within the CTs. Most of the requested TS changes permit an end-state of hot shutdown (Mode 4), if risk is assessed and managed, rather than an end-state of cold shutdown (Mode 5) contained in the current TS. The request was limited to those end-states where: (1) Entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. The STS for B&W plants defines six operational modes. In general, they are: • Mode 1—Power Operation: Keff ≥ 0.99 and power >5% RTP. • Mode 2—Startup: Keff ≥ 0.99 and power ≤ 5% RTP. • Mode 3—Hot Standby: Keff < 0.99 and Tav ≥ [330]°F. • Mode 4—Hot Shutdown: Keff < 0.99 and [330]°F ≥ Tav ≥ [200]°F. • Mode 5—Cold Shutdown: Keff < 0.99 and Tav ≤ [200]°F. E:\FR\FM\21NON1.SGM 21NON1 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices • Mode 6—Refueling: One or more reactor vessel head closure bolts are less than fully tensioned. TSTF–431, Revision 2, generally allows a Mode 4 end-state rather than a Mode 5end-state for selected initiating conditions in order to perform shortduration repairs which necessitate exiting the original Mode of operation. The affected TS are: pwalker on PROD1PC71 with NOTICES 3.3.5 Engineered Safety Feature Actuation System (ESFAS) Instrumentation. 3.3.6 ESFAS Manual Initiation. 3.4.6 Reactor Coolant System (RCS) Loops—MODE 4. 3.4.15 RCS Leakage Detection Instrumentation. 3.5.4 Borated Water Storage Tank (BWST). 3.6.2 Containment Air Locks. 3.6.3 Containment Isolation Valves. 3.6.4 Containment Pressure. 3.6.5 Containment Air Temperature. 3.6.6 Containment Spray and Cooling Systems. 3.7.7 Component Cooling Water System. 3.7.8 Service Water System. 3.7.9 Ultimate Heat Sink. 3.7.10 Control Room Emergency Ventilation System (CREVS). 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS). 3.8.1 AC Sources—Operating. 3.8.4 DC Sources—Operating. 3.8.7 Inverters—Operating. 3.8.9 Distribution Systems—Operating. 2.0 Regulatory Evaluation In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. Pursuant to 10 CFR 50.36(c), TS are required to include items in the following five specific categories related to plant operation: (1) Safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant’s TS. As stated in 10 CFR 50.36(c)(2)(i), the ‘‘Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications * * * .’’ BAW–2441–A, Revision 2, ‘‘RiskInformed Justification for LCO End-State Changes,’’ September 2006 (Reference 1), provides justification for changes to the end-states of selected LCO from Mode 5, cold shutdown, to Mode 4, hot shutdown, in order to (1) reduce risk associated with unnecessary shutdown VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 cooling (SDC) operations, and (2) reduce plant unavailability associated with reduced plant downtime caused by unnecessary cooldown to Mode 5 and subsequent reheat to Mode 3 or 4. Reference 1 provides both a qualitative assessment and a quantitative analysis to confirm that Mode 4 is the preferred end-state from a risk and operational perspective. The qualitative assessment describes the risk associated with operation in Mode 4 compared to operation in Mode 5, in order to justify that the end-state of Mode 4, versus Mode 5, for the proposed LCO conditions invoked, is acceptable. The qualitative assessment concludes that the risk advantages associated with Mode 4 operation versus Mode 5 operation are that: More initiating event mitigating resources are available; human error during SDC initiation and subsequent operation cannot occur; SDC vulnerabilities are avoided; and inadvertent RCS draining via SDC system related misalignments cannot occur. Most of today’s TS and the design basis analyses were developed based on the perception that putting a plant in cold shutdown would result in the safest condition and that the design basis analyses would bound credible shutdown accidents. In the late 1980s and early 1990s, the NRC and licensees recognized that this perception was incorrect and took corrective actions to improve shutdown operation. At the same time, standard TS were developed and many licensees improved their TS. Since enactment of a shutdown rule was expected, almost all TS changes involving power operation, including a revised end-state requirement, were postponed (see, e.g., the Final Policy Statement on TS Improvements (Reference 2)). However, in the mid 1990s, the Commission decided a shutdown rule was not necessary in light of industry improvements. Controlling shutdown risk encompasses control of conditions that can cause potential initiating events and responses to those initiating events that may occur. Initiating events are a function of equipment malfunctions and human error. Responses to events are a function of plant sensitivity, ongoing activities, human error, defense-indepth, and additional equipment malfunctions. In practice, the risk during shutdown operations is often addressed via voluntary actions and application of 10 CFR 50.65 (Reference 3), the maintenance rule. Section 50.65(a)(4) states: ‘‘Before performing maintenance activities * * * the licensee shall assess and manage the increase in risk that PO 00000 Frm 00061 Fmt 4703 Sfmt 4703 65617 may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a riskinformed evaluation process has shown to be significant to public health and safety.’’ Regulatory Guide (RG) 1.182 (Reference 4) provides guidance on implementing the provisions of 10 CFR 50.65(a)(4) by endorsing the revised Section 11 (published separately) to NUMARC 93–01, Revision 2. That section was subsequently incorporated into Revision 3 of NUMARC 93–01 (Reference 5). However, Revision 3 has not yet been formally endorsed by the NRC. The changes in TSTF–431 are consistent with the rules, regulations and associated regulatory guidance, as noted above. 3.0 Technical Evaluation The changes proposed in TSTF–431, Revision 2, are consistent with the changes proposed and justified in Topical Report BAW–2441, Revision 2, as approved by the associated NRC SE (Reference 6). The evaluation included in Reference 6, as appropriate and applicable to the changes of TSTF–431, Revision 2 (Reference 7), is reiterated herein. In its application, the licensee shall commit to TSTF–IG–07–01, Implementation Guidance for TSTF– 431, Revision 1, ‘‘Change in Technical Specifications End-States (BAW–2441),’’ (Reference 8), which addresses a variety of issues. An overview of the generic evaluation and associated risk assessment is provided below, along with a summary of the associated TS changes justified by Reference 1. 3.1 Risk Assessment The objective of the BAW–2441, Revision 2, (Reference 1) risk assessment was to show that any risk increases associated with the proposed changes in TS end-states are either negligible or negative (i.e., a net decrease in risk). BAW–2441, Revision 2, documents a risk-informed analysis of the proposed TS change. Probabilistic Risk Assessment (PRA) results and insights were used, in combination with results of deterministic assessments, to identify and propose changes in ‘‘end-states’’ for B&W plants. This is in accordance with guidance provided in RG 1.174 (Reference 9) and RG 1.177 (Reference 10). The three-tiered approach documented in RG 1.177, ‘‘An Approach for Plant-Specific, RiskInformed Decision Making: Technical Specifications,’’ was followed. The first tier of the three-tiered approach E:\FR\FM\21NON1.SGM 21NON1 pwalker on PROD1PC71 with NOTICES 65618 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices includes the assessment of the risk impact of the proposed change for comparison to acceptance guidelines consistent with the Commission’s Safety Goal Policy Statement, as documented in RG 1.174 ‘‘An Approach for Using Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific Changes to the Licensing Basis.’’ In addition, the first tier aims at ensuring that there are no unacceptable temporary risk increases during the implementation of the proposed TS change, such as when equipment is taken out of service. The second tier addresses the need to preclude potentially high-risk configurations which could result if equipment is taken out of service concurrently with the implementation of the proposed TS change. The third tier addresses the application of a configuration risk management program (CRMP), implemented to comply with 10 CFR 50.65(a)(4) of the Maintenance Rule, for identifying risk-significant configurations resulting from maintenance-related activities and taking appropriate compensatory measures to avoid such configurations. Unless invoked, such as by this or another TS application, 50.65(a)(4) is applicable to maintenance-related activities and does not cover other operational activities beyond the effect they may have on existing maintenance related risk. The risk assessment approach of BAW–2441, Revision 2, was found acceptable in the SE for the topical report. In addition, the analyses show that the the three-tiered approach criteria for allowing TS changes are met as follows: • Risk Impact of the Proposed Change (Tier 1). The risk changes associated with the TS changes in TSTF–431, in terms of mean yearly increases in core damage frequency (CDF) and large early release frequency (LERF), are risk neutral or risk beneficial. In addition, there are no significant temporary risk increases, as defined by RG 1.177 criteria, associated with the implementation of the TS end-state changes. • Avoidance of Risk-Significant Configurations (Tier 2). The performed risk analyses, which are based on single LCOs, show that there are no high-risk configurations associated with the TS end-state changes. The reliability of redundant trains is normally covered by a single LCO. To provide assurance that risk-significant plant equipment outage configurations will not occur when specific equipment is out of service, as part of the implementation of TSTF– 431, the licensee will commit to follow VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 Section 11 of NUMARC 93–01, Revision 3, and to include guidance in appropriate plant procedures and/or administrative controls to preclude high-risk plant configurations when the plant is at the proposed end-state. The staff finds that such guidance is adequate for preventing risk-significant plant configurations. • Configuration Risk Management (Tier 3). The licensee shall have a program, the CRMP, in place to comply with 10 CFR 50.65(a)(4) to assess and manage the risk from proposed maintenance activities. This program can be used to support a licensee decision in selecting the appropriate actions to control risk for most cases in which a risk-informed TS is entered. When multiple LCOs occur, which affect trains in several systems, the plant’s risk-informed CRMP, implemented in response to the Maintenance Rule 10 CFR 50.65(a)(4), shall ensure that high-risk configurations are avoided. In addition, to the extent that the plant PRA is utilized in the CRMP, the plant PRA quality will be assessed in accordance with NRC Regulatory Issue Summary 2007–06, ‘‘Regulatory Guide 1.200 Implementation,’’ (Reference 11). The generic risk impact of the proposed end-state mode change was evaluated subject to the following assumptions: 1. The entry into the proposed endstate is initiated by the inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification. 2. The primary purpose of entering the end-state is to correct the initiating condition and return to power as soon as practical. 3. Plant implementation guidance for the proposed end-state changes is developed to ensure that insights and assumptions made in the risk assessment are properly reflected in the plant-specific CRMP. These assumptions are consistent with typical entries into Mode 4 for short duration repairs, which is the intended use of the TS end-state changes. The staff concludes that, in general, going to Mode 4 (hot shutdown) instead of going to Mode 5 (cold shutdown) to carry out equipment repairs does not have any adverse effect on plant risk. 3.2 Assessment of TS Changes The changes proposed by the licensee and in TSTF–431, Revision 2, are consistent with the changes proposed in topical report BAW–2441, Revision 2, and approved by the NRC SE of August PO 00000 Frm 00062 Fmt 4703 Sfmt 4703 25, 2006. [NOTE: Only those changes proposed in TSTF–431, Revision 2, are addressed in this SE. The SE and associated topical report address the entire fleet of B&W plants, and the plants adopting TSTF–431, Revision 2, must confirm the applicability of the changes to their plant.] Following are the proposed changes, including a synopsis of the STS LCO, the change, and a brief conclusion of acceptability. 3.2.1 TS 3.3.5 Engineering Safety Features Actuation System (ESFAS) Instruments ESFAS instruments initiate high pressure injection (HPI), low pressure injection (LPI), containment spray and cooling, containment isolation, and onsite standby power source start. ESFAS also provides a signal to the Emergency Feedwater Isolation and Control (EFIC) System. This signal initiates emergency feed water (EFW) when HPI is initiated. All functions associated with these systems, structures and components (SSCs) can be initiated via operator action. This may be accomplished at the channel level or the individual component level. LCO: Three channels of ESFAS instrumentation for the applicable parameters shall be operable in each ESFAS train. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.3.5 Condition B, Required Action B.2.3 and addresses only the reactor building (RB) High Pressure and RB High-High Pressure setpoints. Specifically, if two or more channels are inoperable or one channel is inoperable and the required action is not met, then the Mode 5 end-state is prescribed within 36 hours subsequent to an initial cooldown to Mode 3 within 6 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2.3 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: When operating in Mode 4, the reactor system thermal-hydraulic conditions are very different from those associated with a design basis accident (DBA) (at-power). That is, the energy in the RCS is only that associated with decay heat in the core and the stored energy in the reactor coolant system (RCS) components and RCS pressure is reduced (especially toward the lower end of Mode 4). This means that the likelihood of an initiating event (IE) occurring, for which ESFAS would provide mitigating functions, is greatly reduced when operating in Mode 4. Nonetheless, all E:\FR\FM\21NON1.SGM 21NON1 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pwalker on PROD1PC71 with NOTICES redundant functions initiated by ESFAS can be manually initiated to mitigate transients that will proceed more slowly and with reduced challenge to the reactor and containment systems than those associated with at-power operations. Also, when operating toward the lower end of Mode 4, with the steam generators (SGs) in operation and SDC not in operation, risk is reduced; risk associated with shutdown cooling (SDC) operation is avoided. When operating in Mode 4 there are more mitigation systems (e.g., HPI and EFW/auxiliary feed water (AFW)) available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. Based on the above analysis, the staff finds that the above requested change is acceptable. 3.2.2 TS 3.3.6 ESFAS Manual Initiation The ESFAS manual initiation capability allows the operator to actuate ESFAS functions from the main control room in the absence of any other initiation condition. Manually actuated functions include HPI, LPI, containment spray and cooling, containment isolation, and control room isolation. The ESFAS manual initiation ensures that the control room operator can rapidly initiate Engineered Safety Features (ESF) functions at any time. In the absence of manual ESFAS initiation capability, the operator can initiate any and all ESF functions individually at a lower level. LCO: Two manual initiation channels of each one of the following ESFAS functions shall be operable: HPI, LPI, RB Cooling, RB Spray, RB Isolation, and Control Room Isolation. Conditions Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.3.6 Condition B, Required Action B.2. Specifically, if one or more ESFAS functions with one channel are inoperable and the required action and associated completion time are not met, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: When operating in Mode 4, the thermalhydraulic conditions are very different than those associated with a DBA (atpower). That is, the energy in the RCS is only that associated with decay heat in the core and the stored energy in the VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 RCS components and RCS pressure is reduced (especially toward the lower end of Mode 4). This means that the likelihood of an IE occurring, for which ESFAS manual initiation would provide mitigating functions, is greatly reduced when operating in Mode 4. Nonetheless, all redundant functions initiated by ESFAS manual initiation can be manually initiated via individual component controls. In this way, transients, that will proceed more slowly and with reduced challenge to the reactor and containment systems than those associated with at-power operations, will be mitigated. Also, when operating toward the lower end of Mode 4, with the SGs in operation and SDC not in operation, risk is reduced (i.e., the risk associated with SDC avoided). When operating in Mode 4 there are more mitigation systems (e.g. HPI and EFW/AFW) available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. Based on the above assessment, the staff finds that the above requested change is acceptable. 3.2.3 TS 3.4.6 RCS Loops—MODE 4 The purpose of this LCO is to provide forced flow from at least one RCP or one decay heat removal (DHR) pump for core decay heat removal and transport. This LCO allows the two loops that are required to be operable to consist of any combination of RCS or DHR system loops. Any one loop in operation provides enough flow to remove the decay heat from the core. The second loop that is required to be operable provides redundant paths for heat removal. An ancillary function of the RCS and/or DHR loops is to provide mixing of boron in the RCS. When operating in Mode 4 if both RCS loops and one DHR loop is inoperable, the existing LCO requires cooldown to Mode 5. In this situation, SGs are available for core heat removal and transport via natural circulation (NC) in Mode 4 without a need for significant RCS heatup. Proceeding to Mode 5 makes few if any additional systems available for decay heat removal (assuming a failure of the remaining DHR/LPI system). The one system that can be made available in Mode 5 to provide backup to the DHR system is the Borated Water Storage Tank (BWST). It can provide gravity draining to the RCS after cooldown to Mode 5 and subsequent RCS drain down and removal of SG primary side manway covers. This would require a considerable time delay, during which RC temperature would be increasing. PO 00000 Frm 00063 Fmt 4703 Sfmt 4703 65619 LCO: Two loops consisting of any combination of RCS loops and DHR loops shall be operable and one loop shall be in operation. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.4.6 Condition A, Required Action A.2. Specifically, if one required loop is inoperable, then action is taken immediately to restore a second loop to operable status. Further, if the remaining operable loop is a DHR loop, then entry into Mode 5 is required within 24 hours. Proposed Modification for End-State Required Actions: It is proposed that Required Action A.2 be deleted, thus allowing continued operations in Mode 4. Assessment and Finding: When operating in Mode 4, if both RCS loops and one DHR loop are inoperable, the existing LCO requires cooldown to Mode 5. In this situation, SGs are available for core heat removal and transport via NC in Mode 4 without the need for significant RCS heatup. Proceeding to Mode 5 makes few if any additional systems available for decay heat removal (assuming a failure of the remaining DHR system). The one system that can be made available in Mode 5 to provide backup to the DHR system is the BWST. It can provide gravity draining to the RCS after cooldown to Mode 5 and subsequent RCS drain down and removal of SG primary side manway covers. This would require a considerable time delay, during which RC temperature would be increasing. Given these considerations and magnitude of feedwater systems available to feed the SGs, continued use of SGs for this situation will adequately cool the core while avoiding the additional risk associated with SDC. RC boron concentration will have been adjusted prior to cooldown to Mode 4 to provide 1% shutdown margin (SDM) at the target cooldown temperature. Thus, boron concentration adjustments would not be necessary; RC boron would be sufficiently mixed to an equilibrium concentration by this time. When operating in Mode 4 there are more mitigation systems available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. Based upon the above assessment, the staff finds that the above requested change is acceptable. 3.2.4 TS 3.4.15 RCS Leakage Detection Instrumentation One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect E:\FR\FM\21NON1.SGM 21NON1 pwalker on PROD1PC71 with NOTICES 65620 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices extremely small leaks. This LCO requires instruments of diverse monitoring principles to be operable to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition when RCS leakage indicates possible RC pressure boundary (RCPB) degradation. The LCO requirements are satisfied when monitors of diverse measurement means are available. LCO: The following RCS leakage detection instrumentation shall be operable: a. One containment sump monitor and b. One containment atmosphere radioactivity monitor (gaseous or particulate). Conditions Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.4.15 Condition C, Required Action C.2. Specifically, if either the sump monitor or containment atmosphere radioactivity monitor are inoperable and cannot be restored to operability within 30 days, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action C.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: Due to reduced RCS pressures when operating in Mode 4, especially toward the lower end of Mode 4, the likelihood of occurrence of a LOCA is very small; LOCA IE frequencies are reduced compared to at-power operation. Because of this and because the reactor is shutdown with significant radionuclide decay having occurred, the probability of occurrence of a LOCA is decreased while the consequence of such an event is not increased. Additional instruments are available to provide secondary indication of a LOCA, e.g., additional containment radioactivity monitors, grab samples of containment atmosphere, humidity, temperature and pressure. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. When operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to lEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. Based upon the above assessment, the staff finds that the above requested change is acceptable. VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 3.2.5 TS 3.5.4 Tank (BWST) Borated Water Storage The BWST supports the emergency core cooling system (ECCS) and the RB spray (RBS) system by providing a source of borated water for ECCS and containment spray pump operation. The BWST supplies two ECCS trains, each by a separate, redundant supply header. Each header also supplies one train of RBS . A normally open, motor operated isolation valve is provided in each header to allow the operator to isolate the BWST from the ECCS after the ECCS pump suction has been transferred to the containment sump following depletion of the BWST during a LOCA. The ECCS and RBS are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at shutoff head conditions. This LCO ensures that: the BWST contains sufficient borated water to support the ECCS during the injection phase, sufficient water volume exists in the containment sump to support continued operation of the ECCS and containment spray pumps at the time of transfer to the recirculation mode of cooling, and the reactor remains subcritical following a LOCA. Insufficient water inventory in the BWST could result in insufficient cooling capacity of the ECCS when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following a LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside containment. LCO: The BWST shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.5.4 Condition C, Required Action C.2. Specifically, if boron concentration is not within limits for 8 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification: The end-state associated with Required Action C.2, as it relates to the boron concentration requirement of this LCO, is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. No change is being proposed for the water temperature requirement of the LCO. The end-state associated with existing C.2 is proposed to be changed as follows: 4. Split existing Condition A into two conditions (A and C) such that boron concentration and water temperature are addressed separately, i.e., Condition A would address boron concentration and Condition C would address water PO 00000 Frm 00064 Fmt 4703 Sfmt 4703 temperature. In either case the Required Action, i.e., A.1 and C.1, would be to restore the BWST to operable status within 8 hours. 5. A new Condition B would address boron concentration not within limits and the Required Action and associated Completion Time not met. Required Action B.1 would be to be in Mode 3 within 6 hours and B.2 would be to be in Mode 4 within 12 hours. 6. Existing Condition B would be renamed Condition D and would address BWST inoperable for reasons other than Conditions A or C with a Required Action D.1 to restore the BWST to operable status within I hour. Existing Condition C would be renamed Condition E and would address Required Action and associated Completion Time for Conditions other than Condition C or D not met. It would have the Required Action to be in Mode 3 within 6 hours and Mode 5 within 36 hours. Assessment and Finding: The limit for minimum boron concentration in the BWST was established to ensure that, following a DBA large break loss of coolant accident (LBLOCA), with a minimum BWST level, the reactor will remain shut down in the cold condition following mixing of the BWST and RCS water volumes. LBLOCA accident analyses assume that all control rods remain withdrawn from the core. When operating in Mode 4, the control rods will either be inserted or the regulating rod groups will be inserted with one or more of the safety rod groups cocked and armed for automatic RPS insertion. Hence, all rods will not be out should an IE occur. Also, given the highly unlikely possibility of a LBLOCA occurring, it can be assumed all control rods will be inserted should an IE occur while in Mode 4. This provides for the reactor shutdown margin to be very conservative, i.e., in excess of approximately ¥9.0% Dk/k. For these reasons, and the design basis assumptions that (a) deviations in boron concentration will be relatively slow and small and that (b) boric acid addition systems would normally be available (can be powered by [onsite standby power sources]), the staff finds that the above requested change is acceptable. 3.2.6 TS 3.6.2 Containment Air Locks Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all modes of operation. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limits in the event of a DBA. Each air lock is E:\FR\FM\21NON1.SGM 21NON1 pwalker on PROD1PC71 with NOTICES Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices fitted with redundant seals and doors as a design feature for mitigating the DBA. When operating in Mode 4 the energy that can be released to the RB is a fraction of that which would be released for a DBA. Also, the redundant containment spray and cooling systems, required to be operable in Mode 4 but not in Mode 5, will be available to ensure that containment pressure remains low should a LOCA occur. LCO: Two containment air locks shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.6.2 Condition D, Required Action D.2. Specifically, if one or more containment air locks are inoperable for reasons other than condition A or B, then restore the air lock to operable within 24 hours or Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action D.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The energy that can be released to the RB when operating in Mode 4 is only a fraction of that associated with a DBA, thus RB pressure will be only slightly higher should a LOCA occur when operating in Mode 4 as compared to operating in Mode 5. Required Action C.2 requires at least one air lock door to be closed, which combined with reduced RB pressure should result in small containment air lock leakage. Also, significant radionuclide decay will have occurred, i.e., due to plant shutdown. For these reasons, no increase in large early release frequency (LERF) is expected. In the unlikely event that at least one door cannot be closed, evaluation of the effect on plant risk and implementation of any required compensatory measures will be accomplished in accordance with 10 CFR 50.65, i.e., the ‘‘Maintenance Rule.’’ Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5 because there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to IEs that could challenge RCS inventory or decay heat removal. Also, the likelihood of occurrence of a LOCA is very remote, thus the probability of occurrence of a LOCA is decreased while the consequence of such and event is not increased, and the staff finds that the above requested change is acceptable. VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 3.2.7 TS 3.6.3 Containment Isolation Valves (CIVs) The CIVs form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on an automatic isolation signal. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. One of these barriers may be a closed system. These barriers (typically CIVs) make up the Containment Isolation System. Containment isolation occurs upon receipt of a high containment pressure or diverse containment isolation signal. The containment isolation signal closes automatic containment isolation valves in fluid penetrations not required for operation of ESF to prevent leakage of radioactive material. Upon actuation of HPI, automatic containment valves also isolate systems not required for containment or RCS heat removal. Other penetrations are isolated by the use of valves in the closed position or blind flanges. As a result, the CIVs (and blind flanges) help ensure that the containment atmosphere will be isolated in the event of a release of radioactive material to containment atmosphere from the RCS following a DBA. Operability of the containment isolation valves (and blind flanges) supports containment operability during accident conditions. The operability requirements for containment isolation valves help ensure that containment is isolated within the time limits assumed in the safety analyses. Therefore, the operability requirements provide assurance that the containment function assumed in the safety analyses will be maintained. When operating in Mode 4, there is decreased potential for challenges to the containment than assumed in the licensing basis; thus, containment pressures associated with lEs that transfer energy to the containment will be only slightly higher when operating in Mode 4 versus operating in Mode 5. When operating in Mode 4, versus Mode 5, there are more systems available to mitigate precursor events, e.g., loss of feedwater and LOCA, that could cause potential challenges to containment; also, potential fission product release is reduced due to radionuclide decay. LCO: Each containment isolation valve shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is PO 00000 Frm 00065 Fmt 4703 Sfmt 4703 65621 associated with LCO 3.6.3 Condition E, Required Action E.2. Specifically, if the required action and associated completion time cannot be met for penetration flow paths with inoperable isolation valves or RB purge valve leakage limits (Conditions A, B, C and Required Actions A.1, A.2, B.1, C.1 and C.2), then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action E.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: When in Mode 4 (not on SDC) there are more mitigation systems available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. The redundant RBS and RB cooling systems will be available to ensure that containment pressure remains low should a LOCA occur. Because the energy that can be released to the RB when operating in Mode 4 is only a fraction of that associated with a DBA, RB pressure will be only slightly higher should a LOCA occur when operating in Mode 4 as compared to when operating in Mode 5. For these reasons, containment leakage associated with CIVs is small, and with the plant shutdown significant radionuclide decay will have occurred, therefore no increase in LERF is expected. Due to reduced RCS pressures when operating in Mode 4, especially toward the lower end of Mode 4, the likelihood of occurrence of a LOCA is very small, i.e., LOCA IE frequencies are reduced compared to at-power operation. The probability of occurrence of a LOCA is decreased while the consequence of such an event is not increased. Thus, plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Therefore, the staff finds that the above requested change is acceptable. 3.2.8 TS 3.6.4 Containment Pressure The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA or steam line break (SLB). The containment air pressure limit also prevents the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of inadvertent actuation of the containment spray system. Maintaining containment pressure less than or equal to the LCO upper pressure limit (in E:\FR\FM\21NON1.SGM 21NON1 pwalker on PROD1PC71 with NOTICES 65622 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices conjunction with maintaining the containment temperature limit) ensures that: in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure; the containment environmental qualification operating envelope is maintained; and, the ability of containment to perform its design function is ensured. The containment high pressure limit is an initial condition used in the DBA analyses to establish the maximum peak containment internal pressure. Because only a small percentage of the energy assumed for the DBA could be released to the containment, this limit is overly conservative during operations in Mode 4. The low containment pressure limit is based on inadvertent full (both trains) actuation of the RB spray system. Invoking any condition associated with the LCOs being proposed for an endstate change cannot initiate this event; however, should it occur, there is ample time for operator response to mitigate it. LCO: Containment pressure shall be ≥[-2.0] PSIG and ≤ [+3.0] PSIG. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.6.4 Condition B, Required Action B.2. Specifically, if containment pressure exceeds the limit and cannot be restored within one hour, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The redundant RBS and RB cooling systems will be available to ensure that containment pressure remains low should a LOCA occur. Because the energy that can be released to the RB when operating in Mode 4 is only a fraction of that associated with a DBA, RB pressure will be only slightly higher should a LOCA occur when operating in Mode 4 as compared to when operating in Mode 5. In such a situation, the margin to the RB design pressure will be large, i.e., on the order of several tens of PSI. Also, the occurrence of a LOCA of any kind during operation in Mode 4 is considered highly unlikely. Because of this and the occurrence of significant radionuclide decay (i.e., the plant has been shutdown), no increase in LERF is expected should the LCO for high containment pressure be invoked while in Mode 4. This is especially germane considering that operations personnel will commence actions to restore RB pressure to within the limit immediately upon notification that it has exceeded VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 the limit. RB vacuum conditions will not compromise containment integrity of large dry containment of either prestressed or reinforced concrete designs. One plant has a steel containment configuration fitted with a vacuum breaker to mitigate vacuum conditions. The risk associated with Mode 4 operation and RB pressure below the LCO low pressure limit coincident with inadvertent RB spray actuation is considered to be so low as to be inconsequential (a search of available data bases found no record of this situation having occurred to date at any B&W design plants). Also, operations personnel will commence actions to restore RB pressure to within the limit on notification that it has exceeded the limit. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the RB when operating in Mode 4 versus Mode 5, and therefore the staff finds that the above requested change is acceptable. 3.2.9 TS 3.6.5 Containment Air Temperature The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA or SLB. The containment average air temperature limit is derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. This LCO ensures that initial conditions assumed in the analysis of a DBA are not violated during unit operations. The total amount of energy to be removed from the RB Cooling system during post accident conditions is dependent upon the energy released to the containment due to the event as well as the initial containment temperature and pressure. The higher the initial temperature, the higher the resultant peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis. Operation with containment temperature in excess of the LCO limit violates an initial condition assumed in the accident analysis. The limit for containment average air temperature ensures that operation is maintained within the PO 00000 Frm 00066 Fmt 4703 Sfmt 4703 assumptions used in the DBA analysis for containment; LOCA results in the greatest sustained increase in containment temperature. By maintaining containment air temperature at less than the initial temperature assumed in the LOCA analysis, the reactor building design condition will not be exceeded. As a result, the ability of containment to perform its design function is ensured. LCO: Containment average air temperature shall be < [130]°F. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.6.5 Condition B, Required Action B.2. Specifically, if containment air temperature exceeds the limit and cannot be restored within 8 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The redundant RBS and RB cooling systems will be available to ensure that containment temperature remains low should a LOCA occur. Because the energy that can be released to the RB when operating in Mode 4 is only a fraction of that associated with a DBA, the attendant RB temperature (and associated pressure) rise will be well below that associated with a DBA. Also, the occurrence of a LOCA of any kind during operation in Mode 4 is considered highly unlikely. For these reasons and because of the occurrence of significant radionuclide decay (i.e., the plant has been shut down), no increase in LERF is expected. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFV/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the RB when operating in Mode 4 versus Mode 5. Therefore, the staff finds that the above requested change is acceptable. 3.2.10 TS 3.6.6 Containment Spray and Cooling Systems The containment spray and cooling systems provide containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values. Reduction of containment E:\FR\FM\21NON1.SGM 21NON1 pwalker on PROD1PC71 with NOTICES Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pressure and the iodine removal capability of the spray reduces the release of fission product radioactivity from containment to the environment, in the event of a DBA. When operating in Mode 4, the release of stored energy to the RB can be only a small fraction of the energy associated with a DBA. This, along with the fact there are redundant trains of containment spray and cooling, assures this engineered safety feature (ESF) will be supported during operation in Mode 4. Also, the function associated with containment spray iodine removal capability will be less challenged when operating in Mode 4 due to radionuclide decay. LCO: Two containment spray trains and two containment cooling trains shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.6.6 Condition B, Required Action B.2 (containment spray system) and Condition F, Required Action F.2 (containment cooling system). Specifically: if one containment spray train is inoperable and cannot be restored within 72 hours or within 10 days of discovery of failure to meet the LCO, then Mode 3 is prescribed within 6 hours and Mode 5 within 84 hours; and, if two containment cooling trains are inoperable and cannot be restored within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 84 hours to Mode 4 within 60 hours, and the endstate associated with Required Action F.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: In Mode 4 the release of stored energy to the RB would be only that associated with decay heat energy and energy stored in the RCS components. That is, over 95% of the energy assumed to be released to the RB during the DBA LOCA is associated with the core thermal power resulting from 100% full power. Since the reactor is already shut down, such a thermal release to the RB is not possible; only a small fraction of this energy could be released. Occurrence of the DBA, a 28 inch cold leg guillotine break at a RCP discharge, is considered to be very unlikely to occur at any time, much less while operating in Mode 4. Indeed, the occurrence of a LOCA of any kind during operation in this Mode is considered highly unlikely. Due to the redundancy of the containment spray VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 and cooling systems, both their functions are available to control and maintain RB pressure well below the design limit; the function to remove radioactive iodine from the containment atmosphere will also be available. Because the energy that can be released to the RB when operating in Mode 4 is only a fraction of that associated with a DBA, RB pressure will be only slightly higher should a LOCA occur when operating in Mode 4 as compared to when operating in Mode 5. For these reasons containment leakage is small and because significant radionuclide decay will have occurred, (i.e., because the plant has been shut down), no increase in LERF is expected. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the containment spray and cooling systems when operating in Mode 4 versus Mode 5. Therefore, the staff finds that the above requested change is acceptable. 3.2.11 LCO 3.7.7 Component Cooling Water (CCW) System This system provides cooling for ECCS equipment including EFW pumps that function to mitigate loss of feedwater IEs, and containment control equipment. LCO: Two CCW trains shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.7.7 Condition B, Required Action B.2. Specifically, if a CCW train becomes inoperable and cannot be restored within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: In Mode 4 the stored energy of the reactor system would be only that associated with reduced decay heat energy and energy stored in the RCS components. Because of this, heat loads on the CCW system will be greatly reduced from those associated with the DBA, i.e., a LOCA. Also, occurrence of a design bases LOCA is considered to be very unlikely to occur at anytime much less while PO 00000 Frm 00067 Fmt 4703 Sfmt 4703 65623 operating in Mode 4. Indeed, the occurrence of a LOCA of any kind during operation in this Mode is considered highly unlikely. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the CCW system when operating in Mode 4 versus Mode 5. Therefore, the staff finds that the above requested change is acceptable. 3.2.12 TS 3.7.8 Service Water System (SWS) This system provides cooling for equipment that supplies boron to the RCS, i.e., HPI and emergency boration system. LCO: Two SWS trains shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.7.8 Condition B, Required Action B.2. Specifically, if an SWS train becomes inoperable and cannot be restored within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: In Mode 4 the stored energy of the reactor system would be only that associated with reduced decay heat energy and energy stored in the RCS components. Because of this, heat loads on the SWS will be greatly reduced from those associated with the DBA, i.e., a LOCA. Also, occurrence of a design bases LOCA is considered to be very unlikely to occur at anytime much less while operating in Mode 4. Indeed, the occurrence of a LOCA of any kind during operation in this Mode is considered highly unlikely. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the SWS when operating in Mode 4 versus Mode 5, and E:\FR\FM\21NON1.SGM 21NON1 65624 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pwalker on PROD1PC71 with NOTICES therefore, the staff finds that the above requested change is acceptable. 3.2.13 TS 3.7.9 Ultimate Heat Sink (UHS) The UHS provides a heat sink for process and operating heat from safety related components during a transient or accident as well as during normal operation. The UHS has been defined as that complex of water sources, including necessary retaining structures (e.g., a pond with its dam, or a river with its dam), and the canals or conduits connecting the sources with, but not including, the cooling water system intake structures. The two principal functions of the UHS are the dissipation of residual heat after a reactor shutdown, and dissipation of residual heat after an accident. The UHS is the sink for heat removal from the reactor core following all accidents and anticipated occurrences (AOs) in which the unit is cooled down and placed on DHR. Its maximum post accident heat load occurs approximately 20 minutes after a design basis LOCA. Near this time, the unit switches from injection to recirculation and the containment cooling systems are required to remove the core decay heat. LCO: The UHS shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.7.9 Condition C, Required Action C.2. Specifically, if the UHS complex becomes inoperable due to condition A and cannot be restored within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action C.2, as it relates to Condition A only, of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. It is proposed that a new Action B be added, that addresses Condition A only. The Required Action of the new Condition B if Required Action and associated Completion Time of Condition A is not met is proposed to be Mode 3 within 6 hours and Mode 4 within 12 hours. Existing Condition B would be re-lettered to Condition C and existing Condition C would be relettered to Condition D. The first Boolean statement of Condition D would refer only to Condition C. Assessment and Finding: In Mode 4 the stored energy of the reactor system would be only that associated with reduced decay heat energy and energy stored in the RCS components. Because of this, heat loads on the UHS will be greatly reduced from those associated with the DBA, i.e., a LOCA. Also, VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 occurrence of a design basis LOCA is considered to be very unlikely to occur at anytime much less while operating in Mode 4. The occurrence of a LOCA of any kind during operation in this Mode is considered highly unlikely. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) available to respond to an IE that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These considerations ultimately lead to reduced challenges to the UHS when operating in Mode 4 versus Mode 5, and therefore the staff finds that the above requested change is acceptable. 3.2.14 TS 3.7.10 Control Room Emergency Ventilation System (CREVS) The CREVS provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, [chemicals, or toxic gas]. The CREVS consists of two independent, redundant, fan filter assemblies. Upon receipt of the activating signal(s), the normal control room ventilation system is automatically shut down and the CREVS can be manually started. The CREVS is designed to maintain the control room for 30 days of continuous occupancy after a DBA without exceeding a 5 rem whole body dose or its equivalent to any part of the body. LCO: Two CREVS trains shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.7.10 Condition C, Required Action C.2. Specifically, if one train of CREVS becomes inoperable and cannot be restored within 7 days or two CREVS trains become inoperable (due to inoperable control room boundary) and cannot be restored within 24 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action C.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: This system would be required in the event the main control room (MCR) was isolated. Such an isolation would be directly due to an uncontrolled release of radioactivity, [chemicals, or toxic gas]. Uncontrolled release of radioactivity would be associated with a LOCA. A LOCA is considered highly unlikely to occur during Mode 4 operations. This is especially true of operations toward the PO 00000 Frm 00068 Fmt 4703 Sfmt 4703 lower end of Mode 4 while operating on SGs (SDC not in operation). Regardless of the CREVS status, the risks associated with Mode 4 are lower than the Mode 5 operating state. Relative to the uncontrolled release of [chemicals, or toxic gas], this situation is the same as when operating in Mode 5, i.e., frequencies for occurrence of these IEs are the same in Mode 5 as Mode 4. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 there are more mitigation systems available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. These considerations should ultimately lead to reduced challenges to CREVS when operating in Mode 4 versus Mode 5, and therefore, the staff finds that the above requested change is acceptable. 3.2.15 TS 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS) The CREATCS provides temperature control for the control room following isolation of the control room. The CREATCS consists of two independent and redundant trains that provide cooling of recirculated control room air. A cooling coil and a water cooled condensing unit are provided for each system to provide suitable temperature conditions in the control room for operating personnel and safety related control equipment. Ductwork, valves or dampers, and instrumentation also form part of the system. Two redundant air cooled condensing units are provided as a backup to the water cooled condensing unit. Both the water cooled and air cooled condensing units must be operable for the CREATCS to be operable. During emergency operation, the CREATCS maintains the temperature between 70°F and 85°F. The CREATCS is a subsystem of CREVS providing air temperature control for the control room. LCO: Two CREATCS trains shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.7.11 Condition B, Required Action B.2. Specifically, if a CREATCS train becomes inoperable and cannot be restored within 30 days, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action B.2 of this LCO is being proposed to be E:\FR\FM\21NON1.SGM 21NON1 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pwalker on PROD1PC71 with NOTICES changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: This system is a subsystem of CREVS and would be required in the event the MCR was isolated. Such an isolation would be directly due to an uncontrolled release of radioactivity, [chemicals, or toxic gas]. Uncontrolled release of radioactivity would be associated with a LOCA. A LOCA is considered highly unlikely to occur during Mode 4 operations. This is especially true of operations toward the lower end of Mode 4 while operating on SGs (SDC not in operation). Relative to the uncontrolled release of [chemicals, or toxic gas], this situation is the same as when operating in Mode 5, i.e., frequencies for occurrence of these IEs are the same in Mode 5 as in Mode 4. When operating in Mode 4 there are more mitigation systems available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. This should ultimately lead to reduced challenges to CREACTS when operating in Mode 4 versus Mode 5. Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Therefore, the staff finds that the above requested change is acceptable. 3.2.16 TS 3.8.1 AC Source— Operating The unit Class IE AC Electrical Power Distribution System alternating current (AC) sources consist of the offsite power sources (preferred power sources, normal and alternate(s)) and the [onsite standby power sources]. The AC electrical power system provides independence and redundancy to ensure an available source of power to the ESF systems. The onsite Class 1E AC Distribution System is divided into redundant load groups (trains) so that the loss of any one group does not prevent the minimum safety functions from being performed. Each train has connections to two preferred offsite power sources and a single [onsite standby power source]. Offsite power is supplied to the unit switchyard(s) from the transmission network by [two] transmission lines. From the switchyard(s), two electrically and physically separated circuits provide AC power, through [step down station auxiliary transformers] to the 4.16 kV ESF buses. The initial conditions of DBA and transient analyses in the safety analysis report (SAR) assume ESF systems are VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 operable. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded. During operations in Mode 4 there is always a need to assure power is available to SSCs that support the critical safety functions. To this end, AC power sources are assured during occurrence of a loss of offsite power (LOOP) by operation of one of two redundant [onsite standby power sources]. This situation is no different than when operating in Mode 4 or 5. LCO: The following AC electrical power sources shall be operable: a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System, b. Two diesel generators (DG) each capable of supplying one train of the onsite Class 1E AC Electrical Power Distribution System, and [c. Automatic load sequencers for Train A and Train B.] Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.8.1 Condition G, Required Action G.2. Specifically, if the required actions and associated completion times of Condition A, B, C, D, E or F cannot be met, then Mode 3 is prescribed within 12 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action G.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The operability requirements of the AC electrical power sources is predicated on initial assumptions of the accident analyses most notably design basis LOCAs. A design basis LOCA is considered highly unlikely to occur during at-power operations, much less during Mode 4; indeed, the occurrence of a LOCA of any kind during operation in Mode 4 is considered highly unlikely. This is especially true of operations toward the lower end of Mode 4 while operating on SGs (SDC not in operation). Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 there are more mitigation systems (e.g., HPI and EFW/AFWV) available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and PO 00000 Frm 00069 Fmt 4703 Sfmt 4703 65625 EFWV/AFW systems. This consideration is particularly germane as it relates to loss of AC power sources because with the SGs operating in Mode 4, turbine driven EFW pumps (TDEFWPs) are immediately available with SG pressure of [50 PSIG (–2981F RCS temperature)]. These considerations ultimately lead to reduced challenges to CDF and LERF when operating in Mode 4 versus operations in Mode 5. The redundant nature of the AC power sources, including [onsite standby power sources], provides for availability of AC power even if one source becomes inoperable. Therefore, the staff finds that the above requested change is acceptable. 3.2.17 TS 3.8.4 DC Sources— Operating The station direct current (DC) electrical power system provides the alternating current (AC) emergency power system with control power. It also provides both motive and control power to selected safety related equipment and preferred AC vital bus power (via inverters). The DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The [125/250] voltage DC (VDC) electrical power system consists of two independent and redundant safety related Class IE DC electrical power subsystems ([Train A and Train B]). The need for DC power to support the ESFs is assured during a LOOP by operation of one redundant train of station DC power as backed from the [onsite standby power sources] via the associated battery charger. This situation is no different for Mode 4 or Mode 5. LCO: The Train A and Train B DC electrical subsystems shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.8.4 Condition D, Required Action D.2. Specifically, if one DC electrical power subsystem becomes inoperable and cannot be restored within 2 hours, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification: The end-state associated with Required Action D.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The operability requirements of the DC electrical power sources is predicated on initial assumptions of the accident analyses most notably design basis LOCAs. A design basis LOCA is E:\FR\FM\21NON1.SGM 21NON1 65626 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices pwalker on PROD1PC71 with NOTICES considered highly unlikely to occur during at-power operations, much less during Mode 4; indeed, the occurrence of a LOCA of any kind during operation in Mode 4 is considered highly unlikely. This is especially true of operations toward the lower end of Mode 4 while operating on SGs (SDC not in operation). Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 there are more mitigation systems available to respond to IEs that could challenge decay heat removal, than when operating in Mode 5. These systems include the HPI and EFW/AFW systems. This consideration is particularly germane as it relates to loss of DC power sources (control and circuit breaker closure power for plant equipment) because with the SGs operating in Mode 4, TDEFWPs are immediately available with SG pressure of [50 PSIG (–298°F RCS temperature)]. These considerations should ultimately lead to reduced challenges to CDF and LERF when operating in Mode 4 versus operations in Mode 5. The redundant nature of the DC power sources, provides for availability of DC power even if one source becomes in inoperable. Therefore, the staff finds that the above requested change is acceptable. 3.2.18 TS 3.8.9 Distribution Systems—Operating The onsite Class IE AC, DC, and AC vital bus electrical power distribution systems are divided by train into [two] redundant and independent AC, DC, and AC vital bus electrical power distribution subsystems. The required power distribution systems ensure the availability of AC, DC, and AC vital bus electrical power for the systems required to shut down the reactor and maintain it in a safe condition after an AOO or a postulated DBA. Maintaining the train A and B, AC, DC, and AC vital bus electrical power distribution subsystems operable ensures that the redundancy incorporated into the design of ESF is not defeated. Therefore, a single failure within any system or within the electrical power distribution subsystems will not prevent safe shutdown of the reactor. Providing for reactor shutdown is not a concern while operating in Mode 4. However, maintaining safe plant conditions is always a concern and requires that at least one redundant electrical distribution system be operable. This is assured by the redundant electrical distribution system design and the ability to power one of these systems via VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 batteries backed by [onsite standby power sources] for DC distribution and AC vital buses, and [onsite standby power sources] for AC distribution. There is no difference in this situation whether the plant is operating in Mode 4 or 5. LCO: The Train A and Train B AC, DC and AC vital bus electrical power distribution subsystems shall be operable. Condition Requiring Entry into EndState: This proposed end-state change is associated with LCO 3.8.9 Condition D, Required Action D.2. Specifically, if the required actions and associated completion times of Condition A, B or C cannot be met, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours. Proposed Modification for End-State Required Actions: The end-state associated with Required Action D.2 of this LCO is being proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 hours. Assessment and Finding: The operability requirements of the AC, DC, and AC vital bus electrical power distribution systems are predicated on providing the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded in the event of a design basis LOCA. A design basis LOCA is considered highly unlikely to occur during at-power operations, much less during Mode 4; indeed, the occurrence of a LOCA of any kind during operation in Mode 4 is considered highly unlikely. This is especially true of operations at the lower end of Mode 4 while operating on SGs (SDC not in operation). Plant risk is lower when operating in Mode 4 (not on SDC) than when operating in Mode 5; risk associated with SDC operation is avoided. Also, when operating in Mode 4 there are more mitigation systems available to respond to IEs that could challenge RCS inventory or decay heat removal, than when operating in Mode 5. These systems include the HPI system and EFW/AFW systems. This consideration is particularly germane as it relates to loss of electrical power distribution systems because with the SGs operating in Mode 4, TDEFWPs are immediately available with SG pressure of [50 PSIG (-2980F RCS temperature)]. This consideration should ultimately lead to reduced challenges to CDF and LERF when operating in Mode 4 versus operations in Mode 5. The redundant nature of the AC, DC, and AC vital bus electrical power distribution systems, including [onsite standby power sources], provides for availability of electrical power even if one power PO 00000 Frm 00070 Fmt 4703 Sfmt 4703 distribution system becomes inoperable. Therefore, the staff finds that the above requested change is acceptable. 4.0 State Consultation In accordance with the Commission’s regulations, the [____] State official was notified of the proposed issuance of the amendment. The State official had [(1) no comments or (2) the following comments—with subsequent disposition by the staff]. 5.0 Environmental Consideration The amendment changes requirements 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.20. [The NRC staff has determined that the amendment involves a change in surety, insurance, and/or indemnity requirements, or recordkeeping, reporting, or administrative procedures or requirements.] The Commission has previously issued a proposed finding that the amendment involves no significant hazards considerations, and there has been no public comment on the finding [FR ]. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) [and (c)(10)]. 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. 6.0 Conclusion The Commission has concluded, on the basis of 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 amendments will not be inimical to the common defense and security or to the health and safety of the public. 7.0 References 1. BAW–2441–A, Revision 2, ‘‘Risk-Informed Justification for LCO End-State Changes,’’ September 2006. 2. Federal Register, Vol. 58, No. 139, p. 39136, ‘‘Final Policy Statement on Technical Specifications Improvements for Nuclear Power Plants,’’ July 22, 1993. 3. 10 CFR 50.65, Requirements for E:\FR\FM\21NON1.SGM 21NON1 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices ‘‘Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.’’ 4. Regulatory Guide 1.182, ‘‘Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants,’’ May 2000. (ML003699426). 5. NUMARC 93–01, ‘‘Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,’’ Nuclear Management and Resource Council, Revision 3, July 2000. 6. NRC Safety Evaluation for Topical Report BAW–2441, Revision 2, August 25, 2006. (ML062130286). 7. TSTF–431, Revision 2, ‘‘Change in Technical Specifications End-States, BAW–2441–A.’’ 8. TSTF–IG–07–01, Implementation Guidance for TSTF–431, Revision 1, ‘‘Change in Technical Specifications End-States, BAW–2441–A,’’ April 2007. 9. Regulatory Guide 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis,’’ USNRC, August 1998. (ML003740133). 10. Regulatory Guide 1.177, ‘‘An Approach for Pant Specific Risk-Informed Decision Making: Technical Specifications,’’ USNRC, August 1998. (ML003740176). 11. Regulatory Issue Summary 2007–06, ‘‘Regulatory Guide 1.200 Implementation,’’ USNRC, March 22, 2007. pwalker on PROD1PC71 with NOTICES The Following Example of an Application Was Prepared by the NRC Staff To Facilitate Use of the Consolidated Line Item Improvement Process (CLIIP). The Model Provides the Expected Level of Detail and Content for an Application To Change Technical Specifications End-States for B&W Plants Using CLIIP. Licensees Remain Responsible for Ensuring That Their Actual Application Fulfills Their Administrative Requirements as Well as Nuclear Regulatory Commission Regulations U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555. SUBJECT: PLANT NAME DOCKET NO. 50—APPLICATION FOR ADOPTING TECHNICAL SPECIFICATION CHANGE TO REQUIRED ACTION End-States FOR B&W PLANTS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Gentleman: In accordance with th provisions of 10 CFR 50.90 [LICENSEE] is submitting a request for an amendment to the technical specifications (TS) for [PLANT NAME, UNIT NOS.]. The proposed amendment would modify TS requirements for end-states associated with implementation of BAW–2441–A, Revision 2, ‘‘Risk-Informed Justification for LCO End-State Changes.’’ Attachment 1 provides a description of the proposed change, the requested confirmation of applicability, and plant-specific VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 verifications. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides a summary of the regulatory commitments made in this submittal. [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 attachments, is being provided to the designated [STATE] Official. I declare under penalty of perjury under the laws of the United Stats 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 [NAME, TELEPHONE NUMBER] Sincerely, [Name, Title] Attachments: 1. Description and Assessment 2. Proposed Technical Specification Changes 3. Revised Technical Specification Pages 4. Regulatory Commitments 5. Proposed Technical Specification Bases Changes cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Attachment 1—Description and Assessment 1.0 Description The proposed amendment would modify TS end-state requirements associated with implementation of BAW–2441–A, Revision 2, ‘‘RiskInformed Justification for LCO End-State Changes.’’ Current technical specification action requirements frequently require that the unit be brought to cold shutdown when the TS limiting condition for operation for a system has not been met. Depending on the system, and the affected safety function, the requirement to go to cold shutdown may not represent the most risk effective course of action. In accordance with the qualitative risk analysis in BAW–2441–A, Revision 2, and the license amendment request, that provide a basis for changing the TS shutdown action requirement, where appropriate the shutdown end-state is changed from cold shutdown to hot shutdown. The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) STS change TSTF–431, Revision 2. The Federal Register notice published on PO 00000 Frm 00071 Fmt 4703 Sfmt 4703 65627 [DATE] announced the availability of this TS improvement through the consolidated line item improvement process (CLIIP). 2.0 Assessment 2.1 Applicability of Published Safety Evaluation [LICENSEE] has reviewed the safety evaluation dated [DATE] as part of the CLIIP. This review included a review of the NRC staff’s evaluation, as well as the supporting information provided to support TSTF–431, Revision 2. [LICENSEE] has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and the justifications apply to this amendment for the incorporation of the changes to the [PLANT] TS. 2.2 Optional Changes and Variations [LICENSEE] is not proposing any variations or deviations from the TS changes described in TSTF–431, Revision 2, and the NRC staff’s model safety evaluation dated [DATE]. 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Determination [LICENSEE] has reviewed the proposed no significant hazards consideration determination (NSHCD) published in the Federal Register as part of the CLIIP. [LICENSEE] has concluded that the proposed NSHCD presented in the Federal Register notice is applicable to [PLANT] and is [attached, or incorporated herein/ following] satisfying the requirements of 10 CFR 50.91(a). 3.2 Verification and Commitments As discussed in the notice of availability published in the Federal Register on [DATE] for this TS improvement, the [LICENSEE] verifies the applicability of TSTF–431, Revision 2, to [PLANT], and commits to following the guidance set forth in TSTF–IG–07–01, Implementation Guidance for TSTF–431, Revision 1, Change in Technical Specifications EndStates (BAW–2441).’’ The proposed TSTF–431, revision 2, change revises selected required action end-states for B&W STS (NUREG–1430) by allowing plants to go to hot shutdown versus cold shutdown for short durations to effect equipment repairs, after the performance of a plant configuration risk assessment. This application implements TS changes approved in BAW–2441–A, Revision 2, E:\FR\FM\21NON1.SGM 21NON1 65628 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices ‘‘Risk-Informed Justification for LCO End-State Changes.’’ 4.0 ATTACHMENT 2—Proposed Technical Specification Changes (Mark-Up) ATTACHMENT 4—List of Regulatory Commitments ATTACHMENT 3—Proposed Technical Specification Pages The following table identifies those actions committed to by [LICENSEE] in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to [CONTACT NAME]. Environmental Evaluation [LICENSEE] has reviewed the environmental evaluation included in the model safety evaluation dated [DATE] as part of the CLIIP. [LICENSEE] has concluded that the staff’s findings presented in that evaluation are applicable to [PLANT] and the evaluation is [attached, or incorporated herein/following] for this application. Regulatory commitments Due date/event [LICENSEE] will follow the guidance established in Section 11 of NUMARC 93–01, ‘‘Industry Guid- [Ongoing, or implement with amendance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,’’ Nuclear Management] ment and Resource Council, Revision 3, July 2000. [LICENSEE] will follow the guidance established in TSTF–IG–07–01, Implementation Guidance for [Implement with amendment, when TS TSTF–431, Revision 1, ‘‘Change in Technical Specifications End-States, BAW–2441–A,’’ April 2007. Required Action End State remains within the APPLICABILITY of TS] ATTACHMENT 5—Proposed Changes to Technical Specification Bases Pages pwalker on PROD1PC71 with NOTICES Proposed No Significant Hazards Consideration Determination Description of Amendment Request: A change is proposed to the technical specifications (TS) of [plant name], consistent with Technical Specifications Task Force (TSTF) change TSTF–431, Revision 2, to the standard technical specifications (STS) for B&W Plants (NUREG 1430) to allow, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Changes proposed will be made to the [plant name] TS for selected Required Action end-states providing this allowance. Basis for proposed no-significanthazards-consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no-significanthazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change allows a change to certain required end-states when the TS Completion Times for remaining in power operation will be exceeded. Most of the requested technical specification (TS) changes are to permit an end-state of hot shutdown (Mode 4) rather than an VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 end-state of cold shutdown (Mode 5) contained in the current TS. The request was limited to: (1) those end-states where entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Sections 4 and 5 of BAW–2441–A, Revision 2, ‘‘Risk Informed Justification for LCO End-State Changes,’’ for B&W Plants. They provide an integrated discussion of deterministic and probabilistic issues, focusing on specific technical specifications, which are used to support the proposed TS end-state and associated restrictions. The staff finds that the risk insights support the conclusions of the specific TS assessments. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting proposed TSTF–431, Revision 2, are no different than the consequences of an accident prior to its adoption. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. PO 00000 Frm 00072 Fmt 4703 Sfmt 4703 Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). If risk is assessed and managed, allowing a change to certain required end-states when the TS Completion Times for remaining in power operation are exceeded, i.e., entry into hot shutdown rather than cold shutdown to repair equipment, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in TSTF–IG–07–01, Implementation Guidance for TSTF– 431, Revision 1, ‘‘Changes in Technical Specifications End-States, BAW–2441– A,’’ will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The B&WOG’s risk assessment approach is comprehensive and follows staff guidance as documented in RGs 1.174 and 1.177. In addition, the analyses show that the criteria of the E:\FR\FM\21NON1.SGM 21NON1 Federal Register / Vol. 72, No. 224 / Wednesday, November 21, 2007 / Notices three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety. Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. Dated at Rockville, Maryland, this 14th day of November, 2007. For the Nuclear Regulatory Commission. Timothy J. Kobetz, Section Chief, Technical Specifications Branch, Division of Inspection & Regional Support, Office of Nuclear Reactor Regulation. [FR Doc. E7–22738 Filed 11–20–07; 8:45 am] BILLING CODE 7590–01–P FY 2008 Cost of Outpatient Medical, Dental, and Pharmacy Services Furnished by Department of Defense Medical Treatment Facilities; Certain Rates Regarding Recovery From Tortiously Liable Third Persons Office of Management and Budget, Executive Office of the President. ACTION: Notice. AGENCY: By virtue of the authority vested in the President by section 2(a) of Pub. L. 87–603 (76 Stat. 593; 42 U.S.C. 2652), and delegated to the Director of the Office of Management and Budget by the President through Executive Order No. 11541 of July 1, 1970, the rates referenced below are hereby established. These rates are for use in connection with the recovery from tortiously liable third persons for the cost of outpatient medical, dental and pharmacy services furnished by military treatment facilities through the Department of Defense (DoD). The rates have been established in accordance with the requirements of OMB Circular A–25, requiring reimbursement of the full cost of all services provided. The outpatient medical and dental rates referenced are effective upon publication of this notice in the Federal Register and will remain in effect until further notice. Pharmacy rates are updated periodically. The inpatient rates, published on December 9, 2002, pwalker on PROD1PC71 with NOTICES VerDate Aug<31>2005 16:56 Nov 20, 2007 Jkt 214001 Jim Nussle, Director. [FR Doc. E7–22701 Filed 11–20–07; 8:45 am] BILLING CODE 3110–01–P PENSION BENEFIT GUARANTY CORPORATION Proposed Submission of Information Collection for OMB Review; Comment Request; Liability for Termination of Single-Employer Plans Pension Benefit Guaranty Corporation. ACTION: Notice of intention to request extension of OMB approval. AGENCY: SUMMARY: The Pension Benefit Guaranty Corporation (‘‘PBGC’’) intends to request that the Office of Management and Budget (‘‘OMB’’) extend approval, under the Paperwork Reduction Act, of a collection of information contained in its regulation on Liability for Termination of Single-Employer Plans, 29 CFR Part 4062 (OMB Control Number 1212–0017; expires February 29, 2008). This notice informs the public of PBGC’s intent and solicits public comment on the collection of information. OFFICE OF MANAGEMENT AND BUDGET SUMMARY: remain in effect until further notice. A full analysis of the rates is posted at the DoD’s Uniform Business Office Web Site: http://www.tricare.mil/ocfo/_docs/ CY07%20Reimbursement%20Rates11. pdf. The rates can be found at: http:// www.tricare.mil/ocfo/mcfs/ubo/mhs_ rates.cfm. Comments should be submitted by January 22, 2008. ADDRESSES: Comments may be submitted by any of the following methods: Federal eRulemaking Portal: http:// www.regulations.gov. Follow the Web site instructions for submitting comments. E-mail: paperwork.comments@ pbgc.gov. Fax: 202–326–4224. Mail or Hand Delivery: Legislative and Regulatory Department, Pension Benefit Guaranty Corporation, 1200 K Street, NW., Washington, DC 20005–4026. Comments received will be posted to http://www.pbgc.gov. Copies of the collection of information may be obtained without charge by writing to PBGC’s Communications and Public Affairs Department at Suite 240 at the above address or by visiting that office or calling 202–326–4040 during normal business hours. (TTY and TDD users may call the Federal relay service tollDATES: PO 00000 Frm 00073 Fmt 4703 Sfmt 4703 65629 free at 1–800–877–8339 and ask to be connected to 202–326–4040.) The regulation on Liability for Termination of Single-Employer Plans can be accessed on PBGC’s Web site at http:// www.pbgc.gov. FOR FURTHER INFORMATION CONTACT: Thomas H. Gabriel, Attorney, or Catherine B. Klion, Manager, Regulatory and Policy Division, Legislative and Regulatory Department, Pension Benefit Guaranty Corporation, 1200 K Street, NW., Washington, DC 20005–4026, 202– 326–4024. (For TTY and TDD, call 800– 877–8339 and request connection to 202–326–4024.) SUPPLEMENTARY INFORMATION: Section 4062 of the Employee Retirement Income Security Act of 1974, as amended, provides that the contributing sponsor of a single-employer pension plan and members of the sponsor’s controlled group (‘‘the employer’’) incur liability (‘‘employer liability’’) if the plan terminates with assets insufficient to pay benefit liabilities under the plan. PBGC’s statutory lien for employer liability and the payment terms for employer liability are affected by whether and to what extent employer liability exceeds 30 percent of the employer’s net worth. Section 4062.6 of PBGC’s employer liability regulation (29 CFR 4062.6) requires a contributing sponsor or member of the contributing sponsor’s controlled group who believes employer liability upon plan termination exceeds 30 percent of the employer’s net worth to so notify PBGC and to submit net worth information. This information is necessary to enable PBGC to determine whether and to what extent employer liability exceeds 30 percent of the employer’s net worth. The collection of information under the regulation has been approved by OMB under control number 1212–0017 through February 29, 2008. PBGC intends to request that OMB extend its approval for another three years. An agency may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. PBGC estimates that an average of five contributing sponsors or controlled group members per year will respond to this collection of information. PBGC further estimates that the average annual burden of this collection of information will be 12 hours and $3,636 per respondent, with an average total annual burden of 60 hours and $18,120. PBGC is soliciting public comments to— E:\FR\FM\21NON1.SGM 21NON1

Agencies

[Federal Register Volume 72, Number 224 (Wednesday, November 21, 2007)]
[Notices]
[Pages 65615-65629]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-22738]


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

NUCLEAR REGULATORY COMMISSION


Notice of Opportunity To Comment on Model Safety Evaluation on 
Technical Specification Improvement for B&W Reactor Plants To Risk-
Inform Requirements Regarding Selected Required Action End-States Using 
the Consolidated Line Item Improvement Process

AGENCY: Nuclear Regulatory Commission.

ACTION: Request for comment.

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

SUMMARY: Notice is hereby given that the staff of the Nuclear 
Regulatory Commission (NRC) has prepared a model safety evaluation (SE) 
and model license amendment request (LAR) relating to changes to the 
end-state requirements for required actions in B&W reactor plants' 
technical specifications (TS). Current technical specification action 
requirements frequently require that the unit be brought to cold 
shutdown when the technical specification limiting condition for 
operation for a system has not been met. Depending on the system, and 
the affected safety function, the requirement to go to cold shutdown 
may not represent the most risk effective course of action. In 
accordance with a qualitative risk analysis that provides a basis for 
changes to the action requirement to shutdown, where appropriate the 
shutdown end-state is changed from cold shutdown to hot shutdown. The 
affected TS are:

3.3.5 Engineered Safety Feature Actuation System (ESFAS) 
Instrumentation.
3.3.6 ESFAS Manual Initiation.
3.4.6 Reactor Coolant System (RCS) Loops--MODE 4.
3.4.15 RCS Leakage Detection Instrumentation.
3.5.4 Borated Water Storage Tank (BWST).
3.6.2 Containment Air Locks.
3.6.3 Containment Isolation Valves.
3.6.4 Containment Pressure.
3.6.5 Containment Air Temperature.
3.6.6 Containment Spray and Cooling Systems.
3.7.7 Component Cooling Water System.
3.7.8 Service Water System.
3.7.9 Ultimate Heat Sink.
3.7.10 Control Room Emergency Ventilation System (CREVS).
3.7.11 Control Room Emergency Air Temperature Control System 
(CREATCS).
3.8.1 AC Sources--Operating.
3.8.4 DC Sources--Operating.
3.8.7 Inverters--Operating.
3.8.9 Distribution Systems--Operating.

    The NRC staff has also prepared a model no significant hazards 
consideration (NSHC) determination relating to this matter. The purpose 
of these models is to permit the NRC to efficiently process amendments 
that propose to adopt technical specification changes, designated as 
TSTF-431, Revision 2, related to Topical Report BAW-2441, Revision 2, 
``Risk Informed Justification for LCO End-State Changes,'' September 
2006. Licensees of B&W nuclear power reactors to which the models apply 
could then request amendments utilizing the models and justifying the 
applicability of the SE and NSHC determination to their reactors. The 
NRC staff is requesting comments on the model SE, model LAR, and model 
NSHC determination prior to announcing their availability for 
referencing in license amendment applications.

DATES: The comment period expires December 21, 2007. Comments received 
after this date will be considered if it is practical to do so, but the 
Commission is able to ensure consideration only for comments received 
on or before this date.

ADDRESSES: Comments may be submitted either electronically or via U.S. 
mail.
    Submit written comments to Chief, Rules and Directives Branch, 
Division of Administrative Services, Office of Administration, Mail 
Stop: T-6 D59, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001. Hand deliver comments to: 11545 Rockville Pike, Rockville, 
Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays. Copies 
of comments received may be examined at the NRC's Public Document Room, 
11555 Rockville Pike (Room O-1F21), Rockville, Maryland. Comments may 
be submitted by electronic mail to CLIIP@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Timothy Kobetz, Mail Stop: O-12H2, 
Technical Specifications Branch, Division of Inspection & Regional 
Support, 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 for Adopting Standard Technical Specification 
Changes for Power Reactors,'' was issued on March 20, 2000. The 
consolidated line item improvement process (CLIIP) is intended to 
improve the efficiency of NRC licensing processes, by processing 
proposed changes to the standard technical specifications (STS) in a 
manner that supports subsequent license amendment applications. The

[[Page 65616]]

CLIIP includes an opportunity for the public to comment on proposed 
changes to the STS after a preliminary assessment by the NRC staff and 
finding that the change will likely be offered for adoption by 
licensees. The CLIIP directs the NRC staff to evaluate any comments 
received for a proposed change to the STS and to either reconsider the 
change or announce the availability of the change for adoption by 
licensees. Licensees opting to apply for this TS change are responsible 
for reviewing the staff's evaluation, referencing the applicable 
technical justifications, and providing any necessary plant-specific 
information. 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 solicits comment on changes to the end-state 
requirements for required actions, if risk is assessed and managed, for 
the primary purpose of accomplishing short-duration repairs which 
necessitated exiting the original Mode of operation. The change was 
proposed in Topical Report BAW-2441, Revision 2, ``Risk Informed 
Justification for LCO End-State Changes,'' September 2006. This change 
was proposed for incorporation into the standard technical 
specifications by the owners groups participants in the Technical 
Specification Task Force (TSTF) and is designated TSTF-431, Revision 2. 
TSTF-431, Revision 2, can be viewed on the NRC's web page at http://
www.nrc.gov/reactors/operating/licensing/techspecs.html.

Applicability

    This proposal to modify technical specification requirements by the 
adoption of TSTF-431, Revision 2, is applicable to all licensees of B&W 
plants. To efficiently process the incoming license amendment 
applications, the staff requests that each licensee applying for the 
changes proposed in TSTF-431, Revision 2, include Bases for the 
proposed TS consistent with the Bases proposed in TSTF-431, Revision 2. 
To efficiently process the incoming license amendment applications, the 
staff requests that each licensee applying for the changes proposed in 
TSTF-431, Revision 2, use the CLIIP. Licensees are not prevented from 
requesting an alternative approach or proposing the changes without the 
requested Bases and Bases control program. Variations from the approach 
recommended in this notice may 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 to the license, will result in staff rejection of the 
submittal. Instead, licensees desiring significant variations and/or 
additional changes should submit a LAR that does not claim to adopt 
TSTF-431, Revision 2.

Public Notices

    This notice requests comments from interested members of the public 
within 30 days of the date of publication in the Federal Register. 
After evaluating the comments received as a result of this notice, the 
staff will either reconsider the proposed change or announce the 
availability of the change in a subsequent notice (perhaps with some 
changes to the SE, LAR, or the proposed NSHC determination as a result 
of public comments). If the staff announces the availability of the 
change, licensees wishing to adopt the change must submit an 
application in accordance with applicable rules and other regulatory 
requirements. For each application, the staff will publish a notice of 
consideration of issuance of amendment to facility operating licenses, 
a proposed NSHC determination, and a notice of opportunity for a 
hearing. The staff will also publish a notice of issuance of an 
amendment to operating license to announce the modification of end-
state requirements for required actions in plant technical 
specifications.

Proposed Model Plant Specific Safety Evaluation for Technical 
Specification Task Force (TSTF) Change TSTF-431, Revision 2, Change in 
Technical Specifications End-States (BAW-2441), a Consolidated Line 
Item Improvement

U.S. NUCLEAR REGULATORY COMMISSION SAFETY EVALUATION BY THE OFFICE OF 
NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. [------] TO 
FACILITY OPERATING LICENSE NFP-[------] [UTILITY NAME] [PLANT NAME], 
[UNIT ------] DOCKET NO. -[------]
1.0 Introduction
    By letter dated ------------, 20----, [Utility Name] (the licensee) 
proposed changes to the technical specifications (TS) for [plant name]. 
The requested changes are the adoption of TSTF-431, Revision 2, to the 
B&W Reactor Standard Technical Specifications (STS) (NUREG-1430), which 
was proposed by the Technical Specifications Task Force (TSTF) on July 
13, 2007, on behalf of the industry. TSTF-431, Revision 2, incorporates 
the B&W Owners Group (B&WOG) approved Topical Report BAW-2441, Revision 
2, ``Risk Informed Justification for LCO End-State Changes,'' September 
2006, (Reference 1), into the B&W STS (Note: The changes are made with 
respect to Revision 3 of the STS NUREGs).
    TSTF-431, Revision 2, is one of the industry's initiatives 
developed under the Risk Management Technical Specifications (RMTS) 
program. These initiatives are intended to maintain or improve safety 
through the incorporation of risk assessment and management techniques 
in TS, while reducing unnecessary burden and making TS requirements 
consistent with the Commission's other risk-informed regulatory 
requirements, in particular the maintenance rule.
    The Code of Federal Regulations, 10 CFR 50.36, ``Technical 
Specifications,'' states: ``When a limiting condition for operation of 
a nuclear reactor is not met, the licensee shall shut down the reactor 
or follow the remedial action permitted by the technical specification 
until the condition can be met.'' The STS and many plant TS provide a 
completion time (CT) for the plant to meet the limiting condition for 
operation (LCO). If the LCO or the remedial action cannot be met, then 
the reactor is required to be shut down. When the STS and individual 
plant technical specifications were written, the shutdown condition or 
end-state specified was usually cold shutdown.
    Topical Report BAW-2441, Revision 2, provides the technical basis 
to change certain required end-states when the TS Actions for remaining 
in power operation cannot be met within the CTs. Most of the requested 
TS changes permit an end-state of hot shutdown (Mode 4), if risk is 
assessed and managed, rather than an end-state of cold shutdown (Mode 
5) contained in the current TS. The request was limited to those end-
states where: (1) Entry into the shutdown mode is for a short interval, 
(2) entry is initiated by inoperability of a single train of equipment 
or a restriction on a plant operational parameter, unless otherwise 
stated in the applicable TS, and (3) the primary purpose is to correct 
the initiating condition and return to power operation as soon as is 
practical.
    The STS for B&W plants defines six operational modes. In general, 
they are:
     Mode 1--Power Operation: Keff >= 0.99 and power 
>5% RTP.
     Mode 2--Startup: Keff >= 0.99 and power <= 5% 
RTP.
     Mode 3--Hot Standby: Keff < 0.99 and Tav 
>= [330][deg]F.
     Mode 4--Hot Shutdown: Keff < 0.99 and 
[330][deg]F >= Tav >= [200][deg]F.
     Mode 5--Cold Shutdown: Keff < 0.99 and Tav 
<= [200][deg]F.

[[Page 65617]]

     Mode 6--Refueling: One or more reactor vessel head closure 
bolts are less than fully tensioned.
    TSTF-431, Revision 2, generally allows a Mode 4 end-state rather 
than a Mode 5end-state for selected initiating conditions in order to 
perform short-duration repairs which necessitate exiting the original 
Mode of operation. The affected TS are:

3.3.5 Engineered Safety Feature Actuation System (ESFAS) 
Instrumentation.
3.3.6 ESFAS Manual Initiation.
3.4.6 Reactor Coolant System (RCS) Loops--MODE 4.
3.4.15 RCS Leakage Detection Instrumentation.
3.5.4 Borated Water Storage Tank (BWST).
3.6.2 Containment Air Locks.
3.6.3 Containment Isolation Valves.
3.6.4 Containment Pressure.
3.6.5 Containment Air Temperature.
3.6.6 Containment Spray and Cooling Systems.
3.7.7 Component Cooling Water System.
3.7.8 Service Water System.
3.7.9 Ultimate Heat Sink.
3.7.10 Control Room Emergency Ventilation System (CREVS).
3.7.11 Control Room Emergency Air Temperature Control System 
(CREATCS).
3.8.1 AC Sources--Operating.
3.8.4 DC Sources--Operating.
3.8.7 Inverters--Operating.
3.8.9 Distribution Systems--Operating.
2.0 Regulatory Evaluation
    In 10 CFR 50.36, the Commission established its regulatory 
requirements related to the content of TS. Pursuant to 10 CFR 50.36(c), 
TS are required to include items in the following five specific 
categories related to plant operation: (1) Safety limits, limiting 
safety system settings, and limiting control settings; (2) limiting 
conditions for operation (LCOs); (3) surveillance requirements (SRs); 
(4) design features; and (5) administrative controls. The rule does not 
specify the particular requirements to be included in a plant's TS.
    As stated in 10 CFR 50.36(c)(2)(i), the ``Limiting conditions for 
operation are the lowest functional capability or performance levels of 
equipment required for safe operation of the facility. When a limiting 
condition for operation of a nuclear reactor is not met, the licensee 
shall shut down the reactor or follow any remedial action permitted by 
the technical specifications * * * .''
    BAW-2441-A, Revision 2, ``Risk-Informed Justification for LCO End-
State Changes,'' September 2006 (Reference 1), provides justification 
for changes to the end-states of selected LCO from Mode 5, cold 
shutdown, to Mode 4, hot shutdown, in order to (1) reduce risk 
associated with unnecessary shutdown cooling (SDC) operations, and (2) 
reduce plant unavailability associated with reduced plant downtime 
caused by unnecessary cooldown to Mode 5 and subsequent reheat to Mode 
3 or 4. Reference 1 provides both a qualitative assessment and a 
quantitative analysis to confirm that Mode 4 is the preferred end-state 
from a risk and operational perspective. The qualitative assessment 
describes the risk associated with operation in Mode 4 compared to 
operation in Mode 5, in order to justify that the end-state of Mode 4, 
versus Mode 5, for the proposed LCO conditions invoked, is acceptable. 
The qualitative assessment concludes that the risk advantages 
associated with Mode 4 operation versus Mode 5 operation are that: More 
initiating event mitigating resources are available; human error during 
SDC initiation and subsequent operation cannot occur; SDC 
vulnerabilities are avoided; and inadvertent RCS draining via SDC 
system related misalignments cannot occur.
    Most of today's TS and the design basis analyses were developed 
based on the perception that putting a plant in cold shutdown would 
result in the safest condition and that the design basis analyses would 
bound credible shutdown accidents. In the late 1980s and early 1990s, 
the NRC and licensees recognized that this perception was incorrect and 
took corrective actions to improve shutdown operation. At the same 
time, standard TS were developed and many licensees improved their TS. 
Since enactment of a shutdown rule was expected, almost all TS changes 
involving power operation, including a revised end-state requirement, 
were postponed (see, e.g., the Final Policy Statement on TS 
Improvements (Reference 2)). However, in the mid 1990s, the Commission 
decided a shutdown rule was not necessary in light of industry 
improvements.
    Controlling shutdown risk encompasses control of conditions that 
can cause potential initiating events and responses to those initiating 
events that may occur. Initiating events are a function of equipment 
malfunctions and human error. Responses to events are a function of 
plant sensitivity, ongoing activities, human error, defense-in-depth, 
and additional equipment malfunctions.
    In practice, the risk during shutdown operations is often addressed 
via voluntary actions and application of 10 CFR 50.65 (Reference 3), 
the maintenance rule. Section 50.65(a)(4) states: ``Before performing 
maintenance activities * * * the licensee shall assess and manage the 
increase in risk that may result from the proposed maintenance 
activities. The scope of the assessment may be limited to structures, 
systems, and components that a risk-informed evaluation process has 
shown to be significant to public health and safety.'' Regulatory Guide 
(RG) 1.182 (Reference 4) provides guidance on implementing the 
provisions of 10 CFR 50.65(a)(4) by endorsing the revised Section 11 
(published separately) to NUMARC 93-01, Revision 2. That section was 
subsequently incorporated into Revision 3 of NUMARC 93-01 (Reference 
5). However, Revision 3 has not yet been formally endorsed by the NRC.
    The changes in TSTF-431 are consistent with the rules, regulations 
and associated regulatory guidance, as noted above.

3.0 Technical Evaluation

    The changes proposed in TSTF-431, Revision 2, are consistent with 
the changes proposed and justified in Topical Report BAW-2441, Revision 
2, as approved by the associated NRC SE (Reference 6). The evaluation 
included in Reference 6, as appropriate and applicable to the changes 
of TSTF-431, Revision 2 (Reference 7), is reiterated herein.
    In its application, the licensee shall commit to TSTF-IG-07-01, 
Implementation Guidance for TSTF-431, Revision 1, ``Change in Technical 
Specifications End-States (BAW-2441),'' (Reference 8), which addresses 
a variety of issues. An overview of the generic evaluation and 
associated risk assessment is provided below, along with a summary of 
the associated TS changes justified by Reference 1.
3.1 Risk Assessment
    The objective of the BAW-2441, Revision 2, (Reference 1) risk 
assessment was to show that any risk increases associated with the 
proposed changes in TS end-states are either negligible or negative 
(i.e., a net decrease in risk).
    BAW-2441, Revision 2, documents a risk-informed analysis of the 
proposed TS change. Probabilistic Risk Assessment (PRA) results and 
insights were used, in combination with results of deterministic 
assessments, to identify and propose changes in ``end-states'' for B&W 
plants. This is in accordance with guidance provided in RG 1.174 
(Reference 9) and RG 1.177 (Reference 10). The three-tiered approach 
documented in RG 1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decision Making: Technical Specifications,'' was followed. The first 
tier of the three-tiered approach

[[Page 65618]]

includes the assessment of the risk impact of the proposed change for 
comparison to acceptance guidelines consistent with the Commission's 
Safety Goal Policy Statement, as documented in RG 1.174 ``An Approach 
for Using Probabilistic Risk Assessment in Risk-Informed Decisions on 
Plant-Specific Changes to the Licensing Basis.'' In addition, the first 
tier aims at ensuring that there are no unacceptable temporary risk 
increases during the implementation of the proposed TS change, such as 
when equipment is taken out of service. The second tier addresses the 
need to preclude potentially high-risk configurations which could 
result if equipment is taken out of service concurrently with the 
implementation of the proposed TS change. The third tier addresses the 
application of a configuration risk management program (CRMP), 
implemented to comply with 10 CFR 50.65(a)(4) of the Maintenance Rule, 
for identifying risk-significant configurations resulting from 
maintenance-related activities and taking appropriate compensatory 
measures to avoid such configurations. Unless invoked, such as by this 
or another TS application, 50.65(a)(4) is applicable to maintenance-
related activities and does not cover other operational activities 
beyond the effect they may have on existing maintenance related risk.
    The risk assessment approach of BAW-2441, Revision 2, was found 
acceptable in the SE for the topical report. In addition, the analyses 
show that the the three-tiered approach criteria for allowing TS 
changes are met as follows:
     Risk Impact of the Proposed Change (Tier 1). The risk 
changes associated with the TS changes in TSTF-431, in terms of mean 
yearly increases in core damage frequency (CDF) and large early release 
frequency (LERF), are risk neutral or risk beneficial. In addition, 
there are no significant temporary risk increases, as defined by RG 
1.177 criteria, associated with the implementation of the TS end-state 
changes.
     Avoidance of Risk-Significant Configurations (Tier 2). The 
performed risk analyses, which are based on single LCOs, show that 
there are no high-risk configurations associated with the TS end-state 
changes. The reliability of redundant trains is normally covered by a 
single LCO. To provide assurance that risk-significant plant equipment 
outage configurations will not occur when specific equipment is out of 
service, as part of the implementation of TSTF-431, the licensee will 
commit to follow Section 11 of NUMARC 93-01, Revision 3, and to include 
guidance in appropriate plant procedures and/or administrative controls 
to preclude high-risk plant configurations when the plant is at the 
proposed end-state. The staff finds that such guidance is adequate for 
preventing risk-significant plant configurations.
     Configuration Risk Management (Tier 3). The licensee shall 
have a program, the CRMP, in place to comply with 10 CFR 50.65(a)(4) to 
assess and manage the risk from proposed maintenance activities. This 
program can be used to support a licensee decision in selecting the 
appropriate actions to control risk for most cases in which a risk-
informed TS is entered. When multiple LCOs occur, which affect trains 
in several systems, the plant's risk-informed CRMP, implemented in 
response to the Maintenance Rule 10 CFR 50.65(a)(4), shall ensure that 
high-risk configurations are avoided. In addition, to the extent that 
the plant PRA is utilized in the CRMP, the plant PRA quality will be 
assessed in accordance with NRC Regulatory Issue Summary 2007-06, 
``Regulatory Guide 1.200 Implementation,'' (Reference 11).
    The generic risk impact of the proposed end-state mode change was 
evaluated subject to the following assumptions:
    1. The entry into the proposed end-state is initiated by the 
inoperability of a single train of equipment or a restriction on a 
plant operational parameter, unless otherwise stated in the applicable 
technical specification.
    2. The primary purpose of entering the end-state is to correct the 
initiating condition and return to power as soon as practical.
    3. Plant implementation guidance for the proposed end-state changes 
is developed to ensure that insights and assumptions made in the risk 
assessment are properly reflected in the plant-specific CRMP.
    These assumptions are consistent with typical entries into Mode 4 
for short duration repairs, which is the intended use of the TS end-
state changes.
    The staff concludes that, in general, going to Mode 4 (hot 
shutdown) instead of going to Mode 5 (cold shutdown) to carry out 
equipment repairs does not have any adverse effect on plant risk.
3.2 Assessment of TS Changes
    The changes proposed by the licensee and in TSTF-431, Revision 2, 
are consistent with the changes proposed in topical report BAW-2441, 
Revision 2, and approved by the NRC SE of August 25, 2006. [NOTE: Only 
those changes proposed in TSTF-431, Revision 2, are addressed in this 
SE. The SE and associated topical report address the entire fleet of 
B&W plants, and the plants adopting TSTF-431, Revision 2, must confirm 
the applicability of the changes to their plant.] Following are the 
proposed changes, including a synopsis of the STS LCO, the change, and 
a brief conclusion of acceptability.
3.2.1 TS 3.3.5 Engineering Safety Features Actuation System (ESFAS) 
Instruments
    ESFAS instruments initiate high pressure injection (HPI), low 
pressure injection (LPI), containment spray and cooling, containment 
isolation, and onsite standby power source start. ESFAS also provides a 
signal to the Emergency Feedwater Isolation and Control (EFIC) System. 
This signal initiates emergency feed water (EFW) when HPI is initiated. 
All functions associated with these systems, structures and components 
(SSCs) can be initiated via operator action. This may be accomplished 
at the channel level or the individual component level.
    LCO: Three channels of ESFAS instrumentation for the applicable 
parameters shall be operable in each ESFAS train.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.3.5 Condition B, Required Action B.2.3 
and addresses only the reactor building (RB) High Pressure and RB High-
High Pressure setpoints. Specifically, if two or more channels are 
inoperable or one channel is inoperable and the required action is not 
met, then the Mode 5 end-state is prescribed within 36 hours subsequent 
to an initial cooldown to Mode 3 within 6 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2.3 of this LCO is being proposed to 
be changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: When operating in Mode 4, the reactor 
system thermal-hydraulic conditions are very different from those 
associated with a design basis accident (DBA) (at-power). That is, the 
energy in the RCS is only that associated with decay heat in the core 
and the stored energy in the reactor coolant system (RCS) components 
and RCS pressure is reduced (especially toward the lower end of Mode 
4). This means that the likelihood of an initiating event (IE) 
occurring, for which ESFAS would provide mitigating functions, is 
greatly reduced when operating in Mode 4. Nonetheless, all

[[Page 65619]]

redundant functions initiated by ESFAS can be manually initiated to 
mitigate transients that will proceed more slowly and with reduced 
challenge to the reactor and containment systems than those associated 
with at-power operations. Also, when operating toward the lower end of 
Mode 4, with the steam generators (SGs) in operation and SDC not in 
operation, risk is reduced; risk associated with shutdown cooling (SDC) 
operation is avoided. When operating in Mode 4 there are more 
mitigation systems (e.g., HPI and EFW/auxiliary feed water (AFW)) 
available to respond to IEs that could challenge RCS inventory or decay 
heat removal, than when operating in Mode 5. These systems include the 
HPI system and EFW/AFW systems. Based on the above analysis, the staff 
finds that the above requested change is acceptable.
3.2.2 TS 3.3.6 ESFAS Manual Initiation
    The ESFAS manual initiation capability allows the operator to 
actuate ESFAS functions from the main control room in the absence of 
any other initiation condition. Manually actuated functions include 
HPI, LPI, containment spray and cooling, containment isolation, and 
control room isolation. The ESFAS manual initiation ensures that the 
control room operator can rapidly initiate Engineered Safety Features 
(ESF) functions at any time. In the absence of manual ESFAS initiation 
capability, the operator can initiate any and all ESF functions 
individually at a lower level.
    LCO: Two manual initiation channels of each one of the following 
ESFAS functions shall be operable: HPI, LPI, RB Cooling, RB Spray, RB 
Isolation, and Control Room Isolation.
    Conditions Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.3.6 Condition B, Required Action B.2. 
Specifically, if one or more ESFAS functions with one channel are 
inoperable and the required action and associated completion time are 
not met, then Mode 3 is prescribed within 6 hours and Mode 5 within 36 
hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: When operating in Mode 4, the thermal-
hydraulic conditions are very different than those associated with a 
DBA (at-power). That is, the energy in the RCS is only that associated 
with decay heat in the core and the stored energy in the RCS components 
and RCS pressure is reduced (especially toward the lower end of Mode 
4). This means that the likelihood of an IE occurring, for which ESFAS 
manual initiation would provide mitigating functions, is greatly 
reduced when operating in Mode 4. Nonetheless, all redundant functions 
initiated by ESFAS manual initiation can be manually initiated via 
individual component controls. In this way, transients, that will 
proceed more slowly and with reduced challenge to the reactor and 
containment systems than those associated with at-power operations, 
will be mitigated. Also, when operating toward the lower end of Mode 4, 
with the SGs in operation and SDC not in operation, risk is reduced 
(i.e., the risk associated with SDC avoided). When operating in Mode 4 
there are more mitigation systems (e.g. HPI and EFW/AFW) available to 
respond to IEs that could challenge RCS inventory or decay heat 
removal, than when operating in Mode 5. These systems include the HPI 
system and EFW/AFW systems. Based on the above assessment, the staff 
finds that the above requested change is acceptable.
3.2.3 TS 3.4.6 RCS Loops--MODE 4
    The purpose of this LCO is to provide forced flow from at least one 
RCP or one decay heat removal (DHR) pump for core decay heat removal 
and transport. This LCO allows the two loops that are required to be 
operable to consist of any combination of RCS or DHR system loops. Any 
one loop in operation provides enough flow to remove the decay heat 
from the core. The second loop that is required to be operable provides 
redundant paths for heat removal. An ancillary function of the RCS and/
or DHR loops is to provide mixing of boron in the RCS. When operating 
in Mode 4 if both RCS loops and one DHR loop is inoperable, the 
existing LCO requires cooldown to Mode 5. In this situation, SGs are 
available for core heat removal and transport via natural circulation 
(NC) in Mode 4 without a need for significant RCS heatup. Proceeding to 
Mode 5 makes few if any additional systems available for decay heat 
removal (assuming a failure of the remaining DHR/LPI system). The one 
system that can be made available in Mode 5 to provide backup to the 
DHR system is the Borated Water Storage Tank (BWST). It can provide 
gravity draining to the RCS after cooldown to Mode 5 and subsequent RCS 
drain down and removal of SG primary side manway covers. This would 
require a considerable time delay, during which RC temperature would be 
increasing.
    LCO: Two loops consisting of any combination of RCS loops and DHR 
loops shall be operable and one loop shall be in operation.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.4.6 Condition A, Required Action A.2. 
Specifically, if one required loop is inoperable, then action is taken 
immediately to restore a second loop to operable status. Further, if 
the remaining operable loop is a DHR loop, then entry into Mode 5 is 
required within 24 hours.
    Proposed Modification for End-State Required Actions: It is 
proposed that Required Action A.2 be deleted, thus allowing continued 
operations in Mode 4.
    Assessment and Finding: When operating in Mode 4, if both RCS loops 
and one DHR loop are inoperable, the existing LCO requires cooldown to 
Mode 5. In this situation, SGs are available for core heat removal and 
transport via NC in Mode 4 without the need for significant RCS heatup. 
Proceeding to Mode 5 makes few if any additional systems available for 
decay heat removal (assuming a failure of the remaining DHR system). 
The one system that can be made available in Mode 5 to provide backup 
to the DHR system is the BWST. It can provide gravity draining to the 
RCS after cooldown to Mode 5 and subsequent RCS drain down and removal 
of SG primary side manway covers. This would require a considerable 
time delay, during which RC temperature would be increasing. Given 
these considerations and magnitude of feedwater systems available to 
feed the SGs, continued use of SGs for this situation will adequately 
cool the core while avoiding the additional risk associated with SDC. 
RC boron concentration will have been adjusted prior to cooldown to 
Mode 4 to provide 1% shutdown margin (SDM) at the target cooldown 
temperature. Thus, boron concentration adjustments would not be 
necessary; RC boron would be sufficiently mixed to an equilibrium 
concentration by this time. When operating in Mode 4 there are more 
mitigation systems available to respond to IEs that could challenge RCS 
inventory or decay heat removal, than when operating in Mode 5. These 
systems include the HPI system and EFW/AFW systems. Based upon the 
above assessment, the staff finds that the above requested change is 
acceptable.
3.2.4 TS 3.4.15 RCS Leakage Detection Instrumentation
    One method of protecting against large RCS leakage derives from the 
ability of instruments to rapidly detect

[[Page 65620]]

extremely small leaks. This LCO requires instruments of diverse 
monitoring principles to be operable to provide a high degree of 
confidence that extremely small leaks are detected in time to allow 
actions to place the plant in a safe condition when RCS leakage 
indicates possible RC pressure boundary (RCPB) degradation. The LCO 
requirements are satisfied when monitors of diverse measurement means 
are available.
    LCO: The following RCS leakage detection instrumentation shall be 
operable:
    a. One containment sump monitor and
    b. One containment atmosphere radioactivity monitor (gaseous or 
particulate).
    Conditions Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.4.15 Condition C, Required Action C.2. 
Specifically, if either the sump monitor or containment atmosphere 
radioactivity monitor are inoperable and cannot be restored to 
operability within 30 days, then Mode 3 is prescribed within 6 hours 
and Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action C.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: Due to reduced RCS pressures when operating 
in Mode 4, especially toward the lower end of Mode 4, the likelihood of 
occurrence of a LOCA is very small; LOCA IE frequencies are reduced 
compared to at-power operation. Because of this and because the reactor 
is shutdown with significant radionuclide decay having occurred, the 
probability of occurrence of a LOCA is decreased while the consequence 
of such an event is not increased. Additional instruments are available 
to provide secondary indication of a LOCA, e.g., additional containment 
radioactivity monitors, grab samples of containment atmosphere, 
humidity, temperature and pressure. Plant risk is lower when operating 
in Mode 4 (not on SDC) than when operating in Mode 5; risk associated 
with SDC operation is avoided. When operating in Mode 4 (not on SDC) 
there are more mitigation systems (e.g., HPI and EFW/AFW) available to 
respond to lEs that could challenge RCS inventory or decay heat 
removal, than when operating in Mode 5. Based upon the above 
assessment, the staff finds that the above requested change is 
acceptable.
3.2.5 TS 3.5.4 Borated Water Storage Tank (BWST)
    The BWST supports the emergency core cooling system (ECCS) and the 
RB spray (RBS) system by providing a source of borated water for ECCS 
and containment spray pump operation. The BWST supplies two ECCS 
trains, each by a separate, redundant supply header. Each header also 
supplies one train of RBS . A normally open, motor operated isolation 
valve is provided in each header to allow the operator to isolate the 
BWST from the ECCS after the ECCS pump suction has been transferred to 
the containment sump following depletion of the BWST during a LOCA. The 
ECCS and RBS are provided with recirculation lines that ensure each 
pump can maintain minimum flow requirements when operating at shutoff 
head conditions. This LCO ensures that: the BWST contains sufficient 
borated water to support the ECCS during the injection phase, 
sufficient water volume exists in the containment sump to support 
continued operation of the ECCS and containment spray pumps at the time 
of transfer to the recirculation mode of cooling, and the reactor 
remains subcritical following a LOCA. Insufficient water inventory in 
the BWST could result in insufficient cooling capacity of the ECCS when 
the transfer to the recirculation mode occurs. Improper boron 
concentrations could result in a reduction of SDM or excessive boric 
acid precipitation in the core following a LOCA, as well as excessive 
caustic stress corrosion of mechanical components and systems inside 
containment.
    LCO: The BWST shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.5.4 Condition C, Required Action C.2. 
Specifically, if boron concentration is not within limits for 8 hours, 
then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours.
    Proposed Modification: The end-state associated with Required 
Action C.2, as it relates to the boron concentration requirement of 
this LCO, is being proposed to be changed from Mode 5 within 36 hours 
to Mode 4 within 12 hours. No change is being proposed for the water 
temperature requirement of the LCO. The end-state associated with 
existing C.2 is proposed to be changed as follows:
    4. Split existing Condition A into two conditions (A and C) such 
that boron concentration and water temperature are addressed 
separately, i.e., Condition A would address boron concentration and 
Condition C would address water temperature. In either case the 
Required Action, i.e., A.1 and C.1, would be to restore the BWST to 
operable status within 8 hours.
    5. A new Condition B would address boron concentration not within 
limits and the Required Action and associated Completion Time not met. 
Required Action B.1 would be to be in Mode 3 within 6 hours and B.2 
would be to be in Mode 4 within 12 hours.
    6. Existing Condition B would be renamed Condition D and would 
address BWST inoperable for reasons other than Conditions A or C with a 
Required Action D.1 to restore the BWST to operable status within I 
hour.

Existing Condition C would be renamed Condition E and would address 
Required Action and associated Completion Time for Conditions other 
than Condition C or D not met. It would have the Required Action to be 
in Mode 3 within 6 hours and Mode 5 within 36 hours.
    Assessment and Finding: The limit for minimum boron concentration 
in the BWST was established to ensure that, following a DBA large break 
loss of coolant accident (LBLOCA), with a minimum BWST level, the 
reactor will remain shut down in the cold condition following mixing of 
the BWST and RCS water volumes. LBLOCA accident analyses assume that 
all control rods remain withdrawn from the core. When operating in Mode 
4, the control rods will either be inserted or the regulating rod 
groups will be inserted with one or more of the safety rod groups 
cocked and armed for automatic RPS insertion. Hence, all rods will not 
be out should an IE occur. Also, given the highly unlikely possibility 
of a LBLOCA occurring, it can be assumed all control rods will be 
inserted should an IE occur while in Mode 4. This provides for the 
reactor shutdown margin to be very conservative, i.e., in excess of 
approximately -9.0% [Delta]k/k. For these reasons, and the design basis 
assumptions that (a) deviations in boron concentration will be 
relatively slow and small and that (b) boric acid addition systems 
would normally be available (can be powered by [onsite standby power 
sources]), the staff finds that the above requested change is 
acceptable.
3.2.6 TS 3.6.2 Containment Air Locks
    Containment air locks form part of the containment pressure 
boundary and provide a means for personnel access during all modes of 
operation. As such, air lock integrity and leak tightness is essential 
for maintaining the containment leakage rate within limits in the event 
of a DBA. Each air lock is

[[Page 65621]]

fitted with redundant seals and doors as a design feature for 
mitigating the DBA. When operating in Mode 4 the energy that can be 
released to the RB is a fraction of that which would be released for a 
DBA. Also, the redundant containment spray and cooling systems, 
required to be operable in Mode 4 but not in Mode 5, will be available 
to ensure that containment pressure remains low should a LOCA occur.
    LCO: Two containment air locks shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.6.2 Condition D, Required Action D.2. 
Specifically, if one or more containment air locks are inoperable for 
reasons other than condition A or B, then restore the air lock to 
operable within 24 hours or Mode 3 is prescribed within 6 hours and 
Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action D.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: The energy that can be released to the RB 
when operating in Mode 4 is only a fraction of that associated with a 
DBA, thus RB pressure will be only slightly higher should a LOCA occur 
when operating in Mode 4 as compared to operating in Mode 5. Required 
Action C.2 requires at least one air lock door to be closed, which 
combined with reduced RB pressure should result in small containment 
air lock leakage. Also, significant radionuclide decay will have 
occurred, i.e., due to plant shutdown. For these reasons, no increase 
in large early release frequency (LERF) is expected. In the unlikely 
event that at least one door cannot be closed, evaluation of the effect 
on plant risk and implementation of any required compensatory measures 
will be accomplished in accordance with 10 CFR 50.65, i.e., the 
``Maintenance Rule.'' Plant risk is lower when operating in Mode 4 (not 
on SDC) than when operating in Mode 5 because there are more mitigation 
systems (e.g., HPI and EFW/AFW) available to respond to IEs that could 
challenge RCS inventory or decay heat removal. Also, the likelihood of 
occurrence of a LOCA is very remote, thus the probability of occurrence 
of a LOCA is decreased while the consequence of such and event is not 
increased, and the staff finds that the above requested change is 
acceptable.
3.2.7 TS 3.6.3 Containment Isolation Valves (CIVs)
    The CIVs form part of the containment pressure boundary and provide 
a means for fluid penetrations not serving accident consequence 
limiting systems to be provided with two isolation barriers that are 
closed on an automatic isolation signal. Two barriers in series are 
provided for each penetration so that no single credible failure or 
malfunction of an active component can result in a loss of isolation or 
leakage that exceeds limits assumed in the safety analyses. One of 
these barriers may be a closed system. These barriers (typically CIVs) 
make up the Containment Isolation System. Containment isolation occurs 
upon receipt of a high containment pressure or diverse containment 
isolation signal. The containment isolation signal closes automatic 
containment isolation valves in fluid penetrations not required for 
operation of ESF to prevent leakage of radioactive material. Upon 
actuation of HPI, automatic containment valves also isolate systems not 
required for containment or RCS heat removal. Other penetrations are 
isolated by the use of valves in the closed position or blind flanges. 
As a result, the CIVs (and blind flanges) help ensure that the 
containment atmosphere will be isolated in the event of a release of 
radioactive material to containment atmosphere from the RCS following a 
DBA. Operability of the containment isolation valves (and blind 
flanges) supports containment operability during accident conditions. 
The operability requirements for containment isolation valves help 
ensure that containment is isolated within the time limits assumed in 
the safety analyses. Therefore, the operability requirements provide 
assurance that the containment function assumed in the safety analyses 
will be maintained. When operating in Mode 4, there is decreased 
potential for challenges to the containment than assumed in the 
licensing basis; thus, containment pressures associated with lEs that 
transfer energy to the containment will be only slightly higher when 
operating in Mode 4 versus operating in Mode 5. When operating in Mode 
4, versus Mode 5, there are more systems available to mitigate 
precursor events, e.g., loss of feedwater and LOCA, that could cause 
potential challenges to containment; also, potential fission product 
release is reduced due to radionuclide decay.
    LCO: Each containment isolation valve shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.6.3 Condition E, Required Action E.2. 
Specifically, if the required action and associated completion time 
cannot be met for penetration flow paths with inoperable isolation 
valves or RB purge valve leakage limits (Conditions A, B, C and 
Required Actions A.1, A.2, B.1, C.1 and C.2), then Mode 3 is prescribed 
within 6 hours and Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action E.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: When in Mode 4 (not on SDC) there are more 
mitigation systems available to respond to IEs that could challenge RCS 
inventory or decay heat removal, than when operating in Mode 5. The 
redundant RBS and RB cooling systems will be available to ensure that 
containment pressure remains low should a LOCA occur. Because the 
energy that can be released to the RB when operating in Mode 4 is only 
a fraction of that associated with a DBA, RB pressure will be only 
slightly higher should a LOCA occur when operating in Mode 4 as 
compared to when operating in Mode 5. For these reasons, containment 
leakage associated with CIVs is small, and with the plant shutdown 
significant radionuclide decay will have occurred, therefore no 
increase in LERF is expected. Due to reduced RCS pressures when 
operating in Mode 4, especially toward the lower end of Mode 4, the 
likelihood of occurrence of a LOCA is very small, i.e., LOCA IE 
frequencies are reduced compared to at-power operation. The probability 
of occurrence of a LOCA is decreased while the consequence of such an 
event is not increased. Thus, plant risk is lower when operating in 
Mode 4 (not on SDC) than when operating in Mode 5; risk associated with 
SDC operation is avoided. Therefore, the staff finds that the above 
requested change is acceptable.
3.2.8 TS 3.6.4 Containment Pressure
    The containment pressure is limited during normal operation to 
preserve the initial conditions assumed in the accident analyses for a 
LOCA or steam line break (SLB). The containment air pressure limit also 
prevents the containment pressure from exceeding the containment design 
negative pressure differential with respect to the outside atmosphere 
in the event of inadvertent actuation of the containment spray system. 
Maintaining containment pressure less than or equal to the LCO upper 
pressure limit (in

[[Page 65622]]

conjunction with maintaining the containment temperature limit) ensures 
that: in the event of a DBA, the resultant peak containment accident 
pressure will remain below the containment design pressure; the 
containment environmental qualification operating envelope is 
maintained; and, the ability of containment to perform its design 
function is ensured. The containment high pressure limit is an initial 
condition used in the DBA analyses to establish the maximum peak 
containment internal pressure. Because only a small percentage of the 
energy assumed for the DBA could be released to the containment, this 
limit is overly conservative during operations in Mode 4. The low 
containment pressure limit is based on inadvertent full (both trains) 
actuation of the RB spray system. Invoking any condition associated 
with the LCOs being proposed for an end-state change cannot initiate 
this event; however, should it occur, there is ample time for operator 
response to mitigate it.
    LCO: Containment pressure shall be >=[-2.0] PSIG and <= [+3.0] 
PSIG.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.6.4 Condition B, Required Action B.2. 
Specifically, if containment pressure exceeds the limit and cannot be 
restored within one hour, then Mode 3 is prescribed within 6 hours and 
Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: The redundant RBS and RB cooling systems 
will be available to ensure that containment pressure remains low 
should a LOCA occur. Because the energy that can be released to the RB 
when operating in Mode 4 is only a fraction of that associated with a 
DBA, RB pressure will be only slightly higher should a LOCA occur when 
operating in Mode 4 as compared to when operating in Mode 5. In such a 
situation, the margin to the RB design pressure will be large, i.e., on 
the order of several tens of PSI. Also, the occurrence of a LOCA of any 
kind during operation in Mode 4 is considered highly unlikely. Because 
of this and the occurrence of significant radionuclide decay (i.e., the 
plant has been shutdown), no increase in LERF is expected should the 
LCO for high containment pressure be invoked while in Mode 4. This is 
especially germane considering that operations personnel will commence 
actions to restore RB pressure to within the limit immediately upon 
notification that it has exceeded the limit. RB vacuum conditions will 
not compromise containment integrity of large dry containment of either 
pre-stressed or reinforced concrete designs. One plant has a steel 
containment configuration fitted with a vacuum breaker to mitigate 
vacuum conditions. The risk associated with Mode 4 operation and RB 
pressure below the LCO low pressure limit coincident with inadvertent 
RB spray actuation is considered to be so low as to be inconsequential 
(a search of available data bases found no record of this situation 
having occurred to date at any B&W design plants). Also, operations 
personnel will commence actions to restore RB pressure to within the 
limit on notification that it has exceeded the limit.
    Plant risk is lower when operating in Mode 4 (not on SDC) than when 
operating in Mode 5; risk associated with SDC operation is avoided. 
Also, when operating in Mode 4 (not on SDC) there are more mitigation 
systems (e.g., HPI and EFW/AFW) available to respond to an IE that 
could challenge RCS inventory or decay heat removal, than when 
operating in Mode 5. These considerations ultimately lead to reduced 
challenges to the RB when operating in Mode 4 versus Mode 5, and 
therefore the staff finds that the above requested change is 
acceptable.
3.2.9 TS 3.6.5 Containment Air Temperature
    The containment average air temperature is limited during normal 
operation to preserve the initial conditions assumed in the accident 
analyses for a LOCA or SLB. The containment average air temperature 
limit is derived from the input conditions used in the containment 
functional analyses and the containment structure external pressure 
analysis. This LCO ensures that initial conditions assumed in the 
analysis of a DBA are not violated during unit operations. The total 
amount of energy to be removed from the RB Cooling system during post 
accident conditions is dependent upon the energy released to the 
containment due to the event as well as the initial containment 
temperature and pressure. The higher the initial temperature, the 
higher the resultant peak containment pressure and temperature. 
Exceeding containment design pressure may result in leakage greater 
than that assumed in the accident analysis. Operation with containment 
temperature in excess of the LCO limit violates an initial condition 
assumed in the accident analysis. The limit for containment average air 
temperature ensures that operation is maintained within the assumptions 
used in the DBA analysis for containment; LOCA results in the greatest 
sustained increase in containment temperature. By maintaining 
containment air temperature at less than the initial temperature 
assumed in the LOCA analysis, the reactor building design condition 
will not be exceeded. As a result, the ability of containment to 
perform its design function is ensured.
    LCO: Containment average air temperature shall be < [130][deg]F.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.6.5 Condition B, Required Action B.2. 
Specifically, if containment air temperature exceeds the limit and 
cannot be restored within 8 hours, then Mode 3 is prescribed within 6 
hours and Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: The redundant RBS and RB cooling systems 
will be available to ensure that containment temperature remains low 
should a LOCA occur. Because the energy that can be released to the RB 
when operating in Mode 4 is only a fraction of that associated with a 
DBA, the attendant RB temperature (and associated pressure) rise will 
be well below that associated with a DBA. Also, the occurrence of a 
LOCA of any kind during operation in Mode 4 is considered highly 
unlikely. For these reasons and because of the occurrence of 
significant radionuclide decay (i.e., the plant has been shut down), no 
increase in LERF is expected. Plant risk is lower when operating in 
Mode 4 (not on SDC) than when operating in Mode 5; risk associated with 
SDC operation is avoided. Also, when operating in Mode 4 (not on SDC) 
there are more mitigation systems (e.g., HPI and EFV/AFW) available to 
respond to an IE that could challenge RCS inventory or decay heat 
removal, than when operating in Mode 5. These considerations ultimately 
lead to reduced challenges to the RB when operating in Mode 4 versus 
Mode 5. Therefore, the staff finds that the above requested change is 
acceptable.
3.2.10 TS 3.6.6 Containment Spray and Cooling Systems
    The containment spray and cooling systems provide containment 
atmosphere cooling to limit post accident pressure and temperature in 
containment to less than the design values. Reduction of containment

[[Page 65623]]

pressure and the iodine removal capability of the spray reduces the 
release of fission product radioactivity from containment to the 
environment, in the event of a DBA. When operating in Mode 4, the 
release of stored energy to the RB can be only a small fraction of the 
energy associated with a DBA. This, along with the fact there are 
redundant trains of containment spray and cooling, assures this 
engineered safety feature (ESF) will be supported during operation in 
Mode 4. Also, the function associated with containment spray iodine 
removal capability will be less challenged when operating in Mode 4 due 
to radionuclide decay.
    LCO: Two containment spray trains and two containment cooling 
trains shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.6.6 Condition B, Required Action B.2 
(containment spray system) and Condition F, Required Action F.2 
(containment cooling system). Specifically: if one containment spray 
train is inoperable and cannot be restored within 72 hours or within 10 
days of discovery of failure to meet the LCO, then Mode 3 is prescribed 
within 6 hours and Mode 5 within 84 hours; and, if two containment 
cooling trains are inoperable and cannot be restored within 72 hours, 
then Mode 3 is prescribed within 6 hours and Mode 5 within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 84 hours to Mode 4 within 60 hours, and the 
end-state associated with Required Action F.2 of this LCO is being 
proposed to be changed from Mode 5 within 36 hours to Mode 4 within 12 
hours.
    Assessment and Finding: In Mode 4 the release of stored energy to 
the RB would be only that associated with decay heat energy and energy 
stored in the RCS components. That is, over 95% of the energy assumed 
to be released to the RB during the DBA LOCA is associated with the 
core thermal power resulting from 100% full power. Since the reactor is 
already shut down, such a thermal release to the RB is not possible; 
only a small fraction of this energy could be released. Occurrence of 
the DBA, a 28 inch cold leg guillotine break at a RCP discharge, is 
considered to be very unlikely to occur at any time, much less while 
operating in Mode 4. Indeed, the occurrence of a LOCA of any kind 
during operation in this Mode is considered highly unlikely. Due to the 
redundancy of the containment spray and cooling systems, both their 
functions are available to control and maintain RB pressure well below 
the design limit; the function to remove radioactive iodine from the 
containment atmosphere will also be available.
    Because the energy that can be released to the RB when operating in 
Mode 4 is only a fraction of that associated with a DBA, RB pressure 
will be only slightly higher should a LOCA occur when operating in Mode 
4 as compared to when operating in Mode 5. For these reasons 
containment leakage is small and because significant radionuclide decay 
will have occurred, (i.e., because the plant has been shut down), no 
increase in LERF is expected.
    Plant risk is lower when operating in Mode 4 (not on SDC) than when 
operating in Mode 5; risk associated with SDC operation is avoided. 
Also, when operating in Mode 4 (not on SDC) there are more mitigation 
systems (e.g., HPI and EFW/AFW) available to respond to an IE that 
could challenge RCS inventory or decay heat removal, than when 
operating in Mode 5. These considerations ultimately lead to reduced 
challenges to the containment spray and cooling systems when operating 
in Mode 4 versus Mode 5. Therefore, the staff finds that the above 
requested change is acceptable.
3.2.11 LCO 3.7.7 Component Cooling Water (CCW) System
    This system provides cooling for ECCS equipment including EFW pumps 
that function to mitigate loss of feedwater IEs, and containment 
control equipment.
    LCO: Two CCW trains shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.7.7 Condition B, Required Action B.2. 
Specifically, if a CCW train becomes inoperable and cannot be restored 
within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 
within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: In Mode 4 the stored energy of the reactor 
system would be only that associated with reduced decay heat energy and 
energy stored in the RCS components. Because of this, heat loads on the 
CCW system will be greatly reduced from those associated with the DBA, 
i.e., a LOCA. Also, occurrence of a design bases LOCA is considered to 
be very unlikely to occur at anytime much less while operating in Mode 
4. Indeed, the occurrence of a LOCA of any kind during operation in 
this Mode is considered highly unlikely. Plant risk is lower when 
operating in Mode 4 (not on SDC) than when operating in Mode 5; risk 
associated with SDC operation is avoided. Also, when operating in Mode 
4 (not on SDC) there are more mitigation systems (e.g., HPI and EFW/
AFW) available to respond to an IE that could challenge RCS inventory 
or decay heat removal, than when operating in Mode 5. These 
considerations ultimately lead to reduced challenges to the CCW system 
when operating in Mode 4 versus Mode 5. Therefore, the staff finds that 
the above requested change is acceptable.
3.2.12 TS 3.7.8 Service Water System (SWS)
    This system provides cooling for equipment that supplies boron to 
the RCS, i.e., HPI and emergency boration system.
    LCO: Two SWS trains shall be operable.
    Condition Requiring Entry into End-State: This proposed end-state 
change is associated with LCO 3.7.8 Condition B, Required Action B.2. 
Specifically, if an SWS train becomes inoperable and cannot be restored 
within 72 hours, then Mode 3 is prescribed within 6 hours and Mode 5 
within 36 hours.
    Proposed Modification for End-State Required Actions: The end-state 
associated with Required Action B.2 of this LCO is being proposed to be 
changed from Mode 5 within 36 hours to Mode 4 within 12 hours.
    Assessment and Finding: In Mode 4 the stored energy of the reactor 
system would be only that associated with reduced decay heat energy and 
energy stored in the RCS components. Because of this, heat loads on the 
SWS will be greatly reduced from those associated with the DBA, i.e., a 
LOCA. Also, occurrence of a design bases LOCA is considered to be very 
unlikely to occur at anytime much less while operating in Mode 4. 
Indeed, the occurrence of a LOCA of any kind during operation in this 
Mode is considered highly unlikely. Plant risk is lower when operating 
in Mode 4 (not on SDC) than when operating in Mode 5; risk associated 
with SDC operation is avoided. Also, when operating in Mode 4 (not on 
SDC) there are more mitigation systems (e.g., HPI and EFW/AFW) 
available to respond to an IE that could challenge RCS inventory or 
decay heat removal, than when operating in Mode 5. These considerations 
ultimately lead to reduced challenges to the SWS when operating in Mode 
4 versus Mode 5, and

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therefore, the staff finds that the above requested change is 
acceptable.
3.2.13 TS 3.7.9 Ultimate Heat Sink (UHS)
    The UHS provides a heat sink for process and operating heat from 
safety related components during a transient or accident as well as 
during normal operation. The UHS has been defined as that complex of 
water sources, including necessary retaining structures (e.g., a pond 
with its dam, or a river with its dam), and the canals or conduits 
connecting the sources with, but not including, the cooling water 
system intake structures. The two principal functions of the UHS are 
the dissipation of residual heat after a reactor shutdown, and 
dissipation of residual heat after an accident. The UHS is the sink for 
heat removal from the reactor core following all accidents and 
anticipated occurrences (AOs) in which the unit is cooled down and 
placed on DHR. Its maximum post accident heat load occurs approximately 
20 minutes after a design basis LOCA. Near this time, the unit switches 
from injection to recirculation and the containment cooling systems are 
required to remove the core de