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Category:Letter type:L
MONTHYEARL-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report L-PI-23-010, Annual Report of Individual Monitoring2023-04-27027 April 2023 Annual Report of Individual Monitoring L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2023-03-28028 March 2023 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-005, CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2023-03-0303 March 2023 CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) L-PI-23-001, Day Steam Generator Tube Inspection Report2023-01-30030 January 2023 Day Steam Generator Tube Inspection Report L-PI-22-047, Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report2022-12-21021 December 2022 Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2022-12-0202 December 2022 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-22-040, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-10-0606 October 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts2022-09-20020 September 2022 Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts L-PI-22-032, CFR 50.46 LOCA Annual Report2022-06-16016 June 2022 CFR 50.46 LOCA Annual Report L-PI-22-033, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles2022-06-10010 June 2022 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles L-PI-22-003, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0707 June 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-024, Supplement to Application for License Amendment to Implement 24-Month Operating Cycle2022-03-0707 March 2022 Supplement to Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-047, Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 22021-12-0707 December 2021 Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 2 L-PI-21-045, Response to Request for Additional Information Cooling Water System License Amendment Request2021-11-0404 November 2021 Response to Request for Additional Information Cooling Water System License Amendment Request L-PI-21-029, Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.12021-10-0707 October 2021 Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.1 L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions2021-10-0202 October 2021 License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions L-PI-21-032, Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island2021-09-30030 September 2021 Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island L-PI-21-016, Application for License Amendment to Implement 24-Month Operating Cycle2021-08-0606 August 2021 Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-027, 2020 10 CFR 50.46 LOCA Annual Report2021-06-28028 June 2021 2020 10 CFR 50.46 LOCA Annual Report L-PI-21-023, Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report2021-05-14014 May 2021 Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report L-PI-21-007, Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes2021-04-19019 April 2021 Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes L-PI-20-050, Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic2020-10-0707 October 2020 Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic L-PI-20-051, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2020-09-28028 September 2020 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-20-026, Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiativ2020-09-0101 September 2020 Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 L-PI-20-035, = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule2020-07-28028 July 2020 = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule L-PI-20-023, Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI)2020-06-10010 June 2020 Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI) L-PI-20-014, Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI2020-04-29029 April 2020 Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI L-PI-20-004, License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.132020-03-30030 March 2020 License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.13 L-PI-20-001, License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-12020-01-29029 January 2020 License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-1 L-PI-19-041, Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2019-12-23023 December 2019 Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-19-031, License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2019-12-16016 December 2019 License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b L-PI-19-040, License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency2019-10-0707 October 2019 License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency L-PI-19-038, Submittal of Revised Pressure and Temperature Limits Report2019-09-19019 September 2019 Submittal of Revised Pressure and Temperature Limits Report L-PI-19-037, Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals2019-09-16016 September 2019 Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals L-PI-19-025, Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP)2019-08-27027 August 2019 Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For...2019-08-0505 August 2019 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For... L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 L-PI-19-014, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-04-29029 April 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-PI-19-003, Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP)2019-02-0404 February 2019 Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-006, Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements2019-01-29029 January 2019 Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements L-PI-19-005, Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.692019-01-15015 January 2019 Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.69 L-PI-18-063, Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 8052018-12-0606 December 2018 Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 805 2024-01-02
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December 11, 2008 L-PI-08-111 10 CFR 54 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Responses to NRC Requests for Additional Information Dated December 1, 2008 Regarding Application for Renewed Operatinq Licenses By letter dated April 11, 2008, Northern States Power Company, a Minnesota Corporation, (NSPM) submitted an Application for Renewed Operating Licenses (LRA) for the Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. In a letter dated December 1, 2008, the NRC transmitted Requests for Additional Information (RAIs) regarding that application. This letter provides responses to those RAIs.
Enclosure 1 provides the text of each RAI followed by the NSPM response.
If there are any questions or if additional information is needed, please contact Mr. Eugene Eckholt, License Renewal Project Manager.
Summary of Commitments This letter contains no new commitments or changes to existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on December 11, 2008.
Michael D. Wadley Site Vice President, Prairie Island Nuclear Generating Plant Units 1 and 2 Northern States Power Company - Minnesota 1717 Wakonade Drive East - Welch, Minnesota 55089-9642 Telephone: 651.388.1121 k/3-5
Enclosure (1) cc:
Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Prairie Island Indian Community ATTN: Phil Mahowald Minnesota Department of Commerce
Enclosure 1 NSPM Responses to NRC Requests for Additional Information Dated December 1, 2008 RAI 2.4.1-1 Due to lack of clarity in the license renewal application (LRA) Tables 2.4.1-1 and 3.5.2-1, please confirm/clarify if the Spent Fuel Pool (SFP) Divider Gates, the SFP leak-chase channels, and the fuel transfer canal upending frame are structural components in the scope of license renewal and subject to an aging management review (AMR). If yes, include their scoping, screening and AMR results, as appropriate, or clarify the location in the LRA where these components are included. If not, please provide justification for exclusion.
NSPM Response to RAI 2.4.1-1 Spent Fuel Pool (SFP) Divider Gates are not in scope of license renewal since they perform no intended function. As discussed in USAR Section 10.2.2.3, to protect against complete loss of water in the spent fuel pool, spent fuel pool cooling system piping connections enter the top of the pool. The drain connection from the transfer canal to the CVSC holdup tank recirculation pump is at the canal's bottom. Even if the water in the transfer canal were completely drained with the SFP gate removed, the active portion of the spent fuel would not be uncovered. This is because the bottom of the gate connection in the wall separating the transfer canal from the spent fuel pool is at an elevation that would preclude complete drainage.
SFP leak-chase channels are in scope of license renewal. There components are located in the Auxiliary Building, are fabricated from stainless steel, and are located in an embedded-in-concrete environment. See LRA Table 2.4.1-1 on page 2.4-9 (i.e.,
stainless steel components), and Table 3.5.2-1 on page 3.5-77 (i.e., stainless steel components (embedded members)).
The fuel transfer canal upender (or tipping device) is in scope of license renewal. The upending frame is part of the fuel transfer tipping device identified in the LRA Section 2.4.3, page 2.4-18. See LRA Table 3.5.2-3 on pages 3.5-115 and 3.5-116 for aging management of the fuel transfer tipping devices.
RAI 2.4.3-1 In Updated Final Safety Analysis Report (UFSAR) Section 12.2.6, the applicant states that in order to assure the stability and prevent toppling and over-traveling of the containment polar crane or its components, the features incorporated in its design include: (i) up-kick lugs fastened to each truck; (ii)overturning locks fastened to each truck; and (iii) positive wheel stops. Also, in UFSAR Section 12.2.9, the applicant indicates that the spent fuel pool bridge crane, auxiliary building crane and the turbine building crane are protected against tipping, derailments and uncontrolled movements by features that include: (i) crane bridge and trolley being equipped with fixed, fitted rail yokes; and (ii) positive wheel stops and bumpers. From LRA Section 2.4.3, Table 2.4.3-1 and Table 3.5.2-3, it is not clear if the above noted structural components and fasteners of the cranes are included in-scope of license renewal and subject to an AMR.
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Enclosure 1 NSPM Responses to NRC Requests for Additional Information Dated December 1, 2008 Please confirm if these crane components have been screened in as items requiring an AMR. If yes, indicate where these items have been included in the LRA. If not, provide the technical bases for their exclusion.
NSPM Response to RAI 2.4.3-1 Structural components and fasteners for the containment polar crane (up-kick lugs, overturning locks, positive wheel stops), spent fuel pool bridge crane, auxiliary building crane, and the turbine building crane (fixed, fitted rail yokes, and positive wheel stops and bumpers) identified in Sections 12.2.6 and 12.2.9 of the USAR, are in-scope of License Renewal and subject to an AMR. They are included in the LRA description in Section 2.4.3 which characterized them as miscellaneous load carrying components, and in Table 2.4.3-1 under the component heading, "Cranes - Rails" and "Cranes -
Structural Girders." These components are further defined in Table 3.5.2-3 as "Cranes -
structural girders (load carrying structural members, welded and bolted connections
.... )," and "Cranes -rails (rails and associated welded and bolted connections .... )."
Bumpers are considered subcomponents of the crane structural assembly and are not explicitly called out.
RAI 2.4.7-1 In LRA Section 2.4.7, the system function listing under code RCV-04, "Reactor Containment Vessels and their internal structures provide shielding against high energy line breaks," indicates scoping under 10 CFR 54.4(a)(2), which corresponds to all non-safety related systems, structures and components, whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1). The comment under this item on LRA page 2.4-38 states that: "Reactor Containment Vessels and their internal structures are designed to withstand the effects of high energy line breaks without loss of function. Reinforced concrete walls and steel structures inside each Reactor Containment Vessel shield safety related equipment from the effects of a HELB." The NRC staff finds that the above stated structures and structural components are generally safety-related and are in scope in accordance with 10 CFR 54.4(a)(1). Please address the inconsistency.
NSPM Response to RAI 2.4.7-1 Criterion 10 CFR 54.4(a) (2), as it applies to Code RCV-04 on page 2.4-38 of the LRA, is used to describe the HELB protection function applicable to certain non-safety related concrete and steel structures inside each Reactor Containment Vessel including whip restraints and jet impingement shields whose only function is to provide HELB protection for safety related equipment. NEI 95-10, Appendix F, Section 3.4 states that:
"NSR whip restraints, jet impingement shields, blowout panels, etc., that are designed and installed to protect SR equipment from the effects of a HELB, are within the scope of license renewal per 10 CFR 54.4(a)(2)."
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Enclosure 1 NSPM Responses to NRC Requests for Additional Information Dated December 1, 2008 There are also concrete and steel structures inside the reactor containment vessels that perform a HELB function in combination with safety related functions such as missile protection and structural support to safety related components. In an attempt to avoid confusion, the HELB system function was only used to identify non-safety related structures whose only function is to provide HELB protection for safety related equipment. LRA Table 3.5.2-7 provides a list of safety related concrete and steel structures with multiple functions, one of which is HELB protection.
RAI 2.4.7-2 Because of lack of clarity in LRA Tables 2.4.2-1, 2.4.7-1, 3.5.2-2, 3.5.2-7 and the corresponding LRA sections, please indicate where in the LRA are the scoping, screening and AMR results of structural supports (vertical and lateral, as appropriate) for steam generators, reactor coolant pumps and the reactor vessel included. If these structural components were inadvertently not included, please provide their scoping, screening and AMR results, otherwise justify the exclusion.
NSPM Response to RAI 2.4.7-2 Supports for the reactor vessels, steam generators, and reactor coolant pumps are identified in the PINGP USAR, Section 12.2.4 and Table 12.2-1, as Class 1 structures consistent with Chapter Ill.B13.1 of NUREG-1 801. LRA Table 3.5.2-2 refers to them by the component type, "Support (... Class 1 vessels, exchangers, and pumps ...)." Only the Unit 2 steam generator supports and the Units 1 and 2 reactor coolant pump supports are installed using high strength bolts, and therefore Table 3.5.2-2 specifically identifies these supports for this application.
LRA Section 2.4.2 includes a list of in-scope component supports which includes pressure vessels, heat exchangers, and pumps, and LRA Table 2.4.2-1 combines all in-scope supports under the component heading, "Support."
RAI 2.4.8-1 Please confirm if there are any ductbanks and manholes in the yard that are safety-related or important-to-safety or required for regulated events that may be within the scope of license renewal and subject to an AMR. If there are, please provide their scoping, screening and AMR results.
NSPM Response to RAI 2.4.8-1 There are no ductbanks in scope of license renewal, and only one manhole is in scope and subject to an AMR. The single manhole, in scope for the SBO regulated event, is located about 100 feet west of the Security Building. It provides access to splices in the 13.8 kV cables that run from the switchyard to the Cooling Tower Equipment House.
License Renewal Boundary drawing LR-1 93817, entitled, "PINGP Site Layout of the 3
Enclosure 1 NSPM Responses to NRC Requests for Additional Information Dated December 1, 2008 Owner Controlled Area," provides its location (Item 57, coordinate D6). LRA Section 2.4.8 provides a description of the manhole structure, and Table 2.4.8-1 identified its components as "Concrete" and "Steel Components." Table 3.5.2-8 further defines the concrete portion of the structure as "Concrete (... cable vault...)," and its metal components as "Steel components (... miscellaneous structures/equipment items ...)."
The aging effects for the manhole structure are managed by the Structures Monitoring Program based on the results of the AMR.
RAI 2.4.11-1 Section 1.3.2 of the UFSAR states that the plant screenhouse houses the cooling water pumps, fire pumps, circulating water pumps, trash racks and traveling screens. Due to lack of clarity in LRA Tables 2.4.11-1 and 3.5.2-11, please confirm the inclusion or exclusion of the trash racks and traveling screens as structural components within the scope of license renewal and subject to an AMR. If they were not included as an oversight, please provide a description of their scoping and AMR. If they are included elsewhere in the LRA, please indicate the location. If they are excluded from the scope of license renewal and AMR, please provide the basis for their exclusion.
NSPM Response to RAI 2.4.11-1 The trash racks and traveling screen support components are in scope of License Renewal, and the aging effects are managed by the RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants Program. See LRA Table 2.4.11-1 which identifies the components as "Steel Components" and see Table 3.5.2-11 which further defines the components as "Steel components (Screenhouse trash racks, safeguards traveling screen frames, safeguards bay gates, fasteners ...)." The traveling screen portion of the screen assembly is active and therefore, does not require an AMR.
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