ML21111A012
| ML21111A012 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/21/2021 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Rodriguez-Luccioni H | |
| References | |
| Download: ML21111A012 (34) | |
Text
Pre-Application Meeting April 21st, 2021 Saint Lucie Units 1 and 2 Subsequent License Renewal Application
2 Draft Application Materials Subject to Change
3 Agenda Saint Lucie Units 1 and 2 (PSL) Subsequent License Renewal (SLR) Project Team PSL Background General Topics Scoping and Screening Aging Management Programs (AMPs)
Time-Limited Aging Analyses (TLAAs)
Topics of Interest Questions Action Items
William Maher, Senior Project Director Michael Friedman, Engineering Lead Basil Pagnozzi, Engineering Rick Orthen, Environmental Licensing Lead Steve Franzone, Licensing Manager Paul Atkinson, Site SLRA Lead ENERCON James Wicks, Senior Project Manager Jack Hoffman, Project Design Lead Steve Hale, Technical Lead Jeffrey Head, Mechanical Lead Jim Hamlen, Electrical Lead Bruce Beisler, Civil Lead PSL SLR Project Team
5 Original license renewal application (LRA) approved on October 2, 2003 Current license expiration dates, 3/1/2036 (Unit 1), 4/6/2043 (Unit 2)
Based on draft NUREG-1801, Generic Aging Lessons Learned (GALL) 10 programs were updated to NUREG-1801, Generic Aging Lessons Learned (GALL),
Rev 0 as part of the application review process Inspection Procedure (IP) 71003 inspection completed 11/20/2015 (Unit 1) & 10/20/2017 (Unit 2)
Unit 1 entered PEO 3/1/2016 and Unit 2 will enter PEO 4/6/2023 NEI 14-12 AMP effectiveness review completed 1/25/2021 Phase 4 post-approval site inspection for license renewal TBD Licensed core power history, Units 1 and 2 2560 MWt, initial license 2700 MWt, Stretch Power uprate(U1-1981 & U2-1985) 3020 MWt, 11% Extended Power Uprate (EPU) (2012)
PSL Background
6 General Topics Submittal schedule ePortal Folder Structure Operating Experience (OE) including Keywords Application Development Procedures Application of lessons learned Handling of Proprietary Information Incorporation of New ISGs
7 General Topics Submittal schedule Third Quarter of 2021 ePortal Folder Structure Folder for each AMPs Added folders for Instructions, References, Special Topics Operating Experience (OE) including Keywords Latest available keywords utilized (198 total)
The AR search covered the period from 10/01/2010 to 10/01/2020 78,000 initial hits screened, over 1500 identified for further review Experience based interviews, in accordance with EPRI TR-110089, Experience-based Interview Process for Power Plant Management, conducted with AMP owners on-site and at the home office No new aging effects identified Application Development Procedures Application developed under Vendor QA program
8 Application of Lessons Learned Senior, LR/SLR experienced engineers in key positions Benchmarking against other successful SLR/LR utilities Review and incorporation of industry LR operating experience including implementation Incorporated lessons learned from NextEra fleet experience Review and incorporation of previous RAIs including Turkey Point, Peach Bottom and Surry SLRA reviews Including RAI cross reference table Will be available in ePortal as a reviewer's aid Handling of Proprietary Information Minimized in Application Developing Cross-Reference Matrix Discuss Reports under TLAAs General Topics
9 General Topics Application will incorporate:
SLR-ISG-2021-02-MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance, (ML20181A434)
SLR-ISG-2021-04-ELECTRICAL, Updated Aging Management Criteria for Electrical Portions of Subsequent License Renewal Guidance, (ML20181A395)
SLR-ISG-2021-03-STRUCTURES, Updated Aging Management Criteria for Structures Portions of Subsequent License Renewal Guidance ISG, (ML20181A381)
SLR-ISG-2021-01-PWRVI, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors, (ML20217L203)
10 Scoping and Screening Approach to 10 CFR 54.4(a)(2)
Nuclear Energy Institute (NEI) 17-01 Guidance Applicant should rely on the plants current licensing basis (CLB),
plant-specific and industry OE, as appropriate, and existing plant-specific engineering evaluations Refers to Appendix F of NEI 95-10, Rev. 6, for industry guidance Refers to SRP-SLR Report (NUREG-2192), Table 2.1-2 regarding hypothetical failures
11 Scoping and Screening Approach to 10 CFR 54.4(a)(2)
NEI 95-10, R6, App. F Guidance NRC staff position on 54.4(a)(2) scoping criterion taken from Grimes (NRC) to Nelson (NEI), dated March 15, 2002 Non-nuclear safety related (NNS) systems, structures and components (SSCs) meeting the scoping criterion of 54.4(a)(2) fall into 3 categories A plants CLB (missiles, cranes, flooding, high energy line break (HELB))
NNS SSCs directly connected to SR SSCs (piping systems)
NNS SSCs not directly connected to safety related (SR) SSCs Mitigative Option Preventive Option - PSL approach
12 Aging Management Programs AMP Summary AMPs with exceptions to GALL Results of AMP effectiveness review
13 Consistency with NUREG-2191 AMRs (SLRA Section 3)
Very consistent, >98% A through E notes (~2500-line items)
No new aging effects AMPs (Appendix B)
Goal is to maximize consistency Includes aging management effectiveness review of current LR AMPs Turkey Point, Surry, Peach Bottom RAIs addressed Separate section in each AMP basis document summarizes how the RAIs were addressed AMP basis documents will be available to reviewers on ePortal Aging Management Programs AMP Category AMPs Consistent with GALL AMPs Consistent with Enhancement AMPs with Exception AMPs with Exception and Enhancement Plant Specific AMPs Existing 36 4
27 0
4 1
New 12 12 0
0 0
0 Total AMPs 48
14 AMPs with exceptions to GALL XI.M3, Reactor Head Closure Stud Bolting Current bolting is high strength XI.M30, Fuel Oil Chemistry Some fuel oil tanks do not allow for internal inspection, complete draining and/or cleaning XI.M31, Reactor Vessel Material Surveillance Revision to capsule removal schedule XI.S3, ASME Section XI, Subsection IWF High strength bolting is utilized in some applications Aging Management Programs
15 Saint Lucie AMP effectiveness review, evaluated all AMPs - Completed Performed in accordance with NEI 14-12, Aging Management Program Effectiveness, in 2021 Review concluded that all AMPs continue to be effective with no failed elements Aging Management Programs
16 Time-Limited Aging Analyses List, including vendor support No Topical Reports Fluence Methodology Reactor Vessel Embrittlement Impact on surveillance capsule removal schedule
17 TLAA Description Resolution
[10 CFR 54.21(c)(1) Section]
Vendor Section REACTOR VESSEL NEUTRON EMBRITTLEMENT 4.2 Neutron Fluence Projections (iii) the effects of aging on the intended function will be adequately managed for the SPEO Westinghouse 4.2.1 Pressurized Thermal Shock (ii) projected to the end of the SPEO Westinghouse 4.2.2 Upper-Shelf Energy (ii) projected to the end of the SPEO Westinghouse 4.2.3 Adjusted Reference Temperature (ii) projected to the end of the SPEO Westinghouse 4.2.4 Pressure-Temperature Limits and LTOP Setpoints (iii) the effects of aging on the intended function will be adequately managed for the SPEO Westinghouse ENERCON 4.2.5 METAL FATIGUE 4.3 Metal Fatigue of Class 1 Components (iii) the effects of aging on the intended function will be adequately managed for the SPEO Westinghouse ENERCON Structural Integrity Associates Framatome 4.3.1 Metal Fatigue of Non-Class 1 Components (i) remains valid for the SPEO ENERCON 4.3.2 High-Energy Line Break Analyses (i) remains valid for the SPEO ENERCON Environmentally Assisted Fatigue (iii) the effects of aging on the intended function will be adequately managed for the SPEO Westinghouse ENERCON Structural Integrity Associates Framatome 4.3.3 Time-Limited Aging Analyses
18 TLAA Description Resolution
[10 CFR 54.21(c)(1) Section]
Vendor Section ENVIRONMENTAL QUALIFICATION (EQ) OF ELECTRICAL EQUIPMENT (iii) the effects of aging on the intended function will be adequately managed for the SPEO ENERCON 4.4 CONCRETE CONTAINMENT TENDON PRESTRESS Not Applicable 4.5 CONTAINMENT LINER PLATE, METAL CONTAINMENTS, AND PENETRATIONS FATIGUE (i) remains valid for the SPEO ENERCON 4.6 OTHER PLANT-SPECIFIC TLAAS 4.7 Leak-Before-Break of Reactor Coolant System Loop Piping (ii) projected to the end of the SPEO Westinghouse 4.7.1 Alloy 600 Instrument Nozzle Repairs (ii) projected to the end of the SPEO Westinghouse 4.7.2 Unit 1 Core Support Barrel Repair Plug Preload Relaxation (ii) projected to the end of the SPEO Westinghouse 4.7.3 Flaw Tolerance Evaluation for CASS Piping (ii) projected to the end of the SPEO SIA 4.7.4 Reactor Coolant Pump Flywheel (i) remains valid for the SPEO ENERCON 4.7.5 Reactor Coolant Pump Code Case N-481 (ii) projected to the end of the SPEO Westinghouse 4.7.6 Crane Load Cycle Limits (i) remains valid for the SPEO ENERCON 4.7.7 Time-Limited Aging Analyses
19 Time-Limited Aging Analyses Fluence Methodology Neutron transport followed the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel NRC approved methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET Methodology has been generically approved for calculations of exposure of the reactor pressure vessel (RPV) beltline Exposure Projections Based on 72 EFPY 10% positive bias on peripheral assembly power for projected cycles
20 Time-Limited Aging Analyses Reactor Vessel Embrittlement - Impact on surveillance capsule removal schedule Current approved withdrawal of capsules Unit 1 - Two capsules schedule for testing (~ 38 EFPY & 45 EFPY)
- Unit 1 limiting weld material is not in the program. However, we have embrittlement data > 72 EFPY fluence from a sister plant
- Standby capsule is damaged. FPL plans to revise schedule to include either damaged capsule (if tooling can be developed) or scheduled capsule as both have the same lead factor Unit 2 - One scheduled capsule that leads vessel. Remaining standby capsules have a lead factor of <1
- Unit 2 limiting material is plate and all materials are low copper Similar to Turkey Point & Point Beach, an incremental adjustment to the approved withdrawal schedule will allow sufficient material data and dosimetry for the end of the subsequent period of extended operation (SPEO)
21 Topics of Interest Reactor Vessel Internals Irradiation of Concrete Irradiation of Reactor Vessel (RV) supports
22 Reactor Vessel Internals (RVI) Gap Analysis for SLR Existing PSL RVI AMP based on the Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (MRP-227-1-A)
GALL-SLR Report (NUREG 2191) allows use of the existing RVI AMP if supplemented by a 60 to 80 Year gap analysis with MRP-227-A as the starting point SLR-ISG-2021-01-PWRVI will be incorporated into the PSL application Permits the use of MRP-227-1-A as the starting point for the Gap Analysis PSL RVI Gap Analysis for SLR will use MRP-227-1-A as the starting point Will utilize joint industry issue program documents to perform this analysis MRP-191 Rev 2 MRP-2018-022 Gap Analysis will consider all relevant industry OE FPL will continue to actively participate in the joint industry issue programs regarding RVI and update the program, as needed Topics of Interest - Reactor Vessel Internals
23 Design configuration The PSW is a 7.25 ft thick cylindrical reinforced concrete wall and surrounds the RPV Both Unit 1 and Unit 2 PSWs have the same dimensions and the same concrete properties.
Unit 2 PSW has slightly less vertical reinforcement steel than Unit 1 PSW However, Unit 2 PSW has Grade 60 reinforcement steel whereas Unit 1 PSW has Grade 40 steel.
5000 psi concrete Topics of Interest - Irradiation of Concrete
24 Design configuration The RPV is supported at three points on three steel beam-column assemblies within the reactor cavity.
The beams are embedded in the PSW approximately 6 ft on each end The column is bolted to the underside of the girder and to the reactor cavity floor Load transfer between the RPV system and the RPV support occur between the support shoe which is welded to the reactor nozzle and steel bearing plates designed into the top of the steel support beam Topics of Interest - Irradiation of Concrete
25 Topics of Interest - Irradiation of Concrete Elevation View of the Reactor Building, Section A-A of Plan View
26 Topics of Interest - Irradiation of Concrete Plan View of the Reactor Building at Elevation 18 ft
27 Topics of Interest - Irradiation of Concrete Plan View of RPV Supports
28 Maximum exposures on the inner surface of PSW at the end of the SPEO (72 EFPY) based on very conservative UFSAR numbers (PSL2 bounding)
Neutron fluence E > 0.1 MeV - 7.0 x 1019 n/cm2 Gamma dose - 3.2 x 1010 rads PSW exposures result in an assumed loss of strength as follows 100% strength loss for first 4.5 due to neutron fluence and radiation induced volumetric expansion (RIVE) 25% strength loss for an additional 25.5 for a total of 30 Evaluation demonstrates PSW maintains its structural integrity under CLB loading Topics of Interest - Irradiation of Concrete
29 Topics of Interest - Irradiation of RV Supports Plan View of RPV Supports
30 Topics of Interest - Irradiation of RV Supports
31 FPL will perform a qualitative assessment of the PSL Units 1 & 2 reactor pressure vessel (RPV) supports, as it pertains to the irradiation aging effects for the SPEO This assessment will provide the technical basis to support an inspection-based approach The assessment has two (2) elements:
Qualitative comparison of the technical attributes Inspection-base attributes Topics of Interest - Irradiation of RV Supports
32 Assessment topic areas include:
Compare Geometry & Materials Identify analogous components between PBN and PSL Compile all material types for downstream evaluation Compare Fracture Toughness Consider CMTRs and industry guidance as applicable Compare KIC at critical locations Compare Stresses Develop stresses from plant specific model Compare stress at critical locations Validate Inspection Plan Review current inspection capability Utilize above comparisons to evaluate RPV Inspection Program Topics of Interest - Irradiation of RV Supports
33 Questions
34 Action Items