ML19318F291
| ML19318F291 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Mcguire, Catawba, Harris, Brunswick, Robinson, McGuire  |
| Issue date: | 11/18/2019 |
| From: | Duke Energy Corp |
| To: | Andrew Hon NRC/NRR/DORL/LPL2-2 |
| Hon A, NRR/DORL/LPL2-2, 415-8480 | |
| References | |
| Download: ML19318F291 (20) | |
Text
Duke Energy Pre-Submittal Meeting - November 18, 2019
Request for Alternative RA-19-0352 Duke Energy Request for Alternative for Reactor Vessel Closure Stud Exam Extensions Brunswick Steam Electric Plant (BNP)
Catawba Nuclear Station (CNS)
Harris Nuclear Power Plant (HNP)
McGuire Nuclear Station (MNS)
Oconee Nuclear Station (ONS)
Robinson Steam Electric Plant (RNP) 2
Participants Mark Pyne, ISI Engineering Programs Manager - Duke Energy Austin Keller, Senior Nuclear Engineer - Duke Energy Art Zaremba, Fleet Licensing Manager - Duke Energy Chet Sigmon, Senior Nuclear Engineer - Duke Energy Robert Grizzi, Program Manager - EPRI Gary Stevens, Technical Executive - EPRI Douglas Kull, Principal Technical Leader - EPRI John Broussard, Principal Engineer - Dominion Engineering, Inc.
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Key Goals for This Meeting:
Brief NRC on Dukes proposed Request for Alternative, project scope, and proposed timeline Ensure common understanding for Duke Energys request, technical scope, and regulatory expectations Obtain NRC feedback prior to formal submittal 4
Agenda I.
Description of Request II. Regulatory Requirements and Guidance III. Proposed Alternative IV. EPRI Technical Report V. Conclusion / Schedule 5
I. Description of Request
Background:
Affected components: Reactor Vessel Closure Studs ASME Section XI, Table IWB-2500-1, Exam Category B-G-1, Item B6.20 require volumetric exams once during each Section XI Interval with exam volume shown in Figure IWB-2500-12.
Additionally, per Note (7) of Table IWB-2500-1 (B-G-1), when bolts or studs are removed for examination, surface examination meeting the acceptance criteria of IWB-3515 may be substituted for volumetric examination. The surface examination area is the stud external surface along the entire engaged length of the stud, as shown in ASME Section XI Figure IWB-2500-12.
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I. Description of Request 7
II. Regulatory Requirements and Guidance 10 CFR 50.55a Codes and Standards ASME Section XI IWB-2500(a), Table IWB-2500-1, Examination Category B-G-1, Item No. B6.20 10 CFR 50.55a(z)(1) 8
III. Proposed Alternative Duke Energy is requesting an alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 (B-G-1), Examination Category B-G-1, Item B6.20.
The proposed alternative is to extend the frequency of reactor vessel closure stud volumetric or surface examinations past the remainder of the currently licensed operating periods for the plants listed in Table 1. The current licensing periods for these plants are summarized in Table
===2.
Duration of Proposed Alternative===
The proposed alternative is requested for the remainder of the currently licensed 60 year operating periods for the plants listed in Tables 1 and 2.
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III. Proposed Alternative Table 1: Plants Included in This Request for Alternative and Their Current ISI Intervals and Applicable ASME Code Section XI Editions/Addenda 10 Plant/Unit(s)
ISI Interval ASME Section XI Code Edition/Addenda Interval Start Date Interval End Date Brunswick Steam Electric Plant Units 1 and 2 (BWR)
Fifth 2007 Edition, Through 2008 Addendum 05/11/2018 05/10/2028 Catawba Nuclear Station Units 1 and 2 (PWR)
Fourth 2007 Edition, Through 2008 Addendum 08/19/2015 12/06/2024 (Unit 1) 02/24/2026 (Unit 2)
H.B. Robinson Steam Electric Plant Unit 2 (PWR)
Fifth 2007 Edition, Through 2008 Addendum 07/21/2012 02/19/2023 McGuire Nuclear Station Units 1 and 2 (PWR)
Fourth 2007 Edition, Through 2008 Addendum 12/01/2011 (Unit 1) 07/15/2014 (Unit 2) 11/30/2021 (Unit 1) 12/14/2024 (Unit 2)
Oconee Nuclear Station Units 1, 2, and 3 (PWR)
Fifth 2007 Edition, Through 2008 Addendum 07/15/2014 07/15/2024 Shearon Harris Nuclear Power Plant Unit 1 (PWR)
Fourth 2007 Edition, Through 2008 Addendum 09/09/2017 09/08/2027
III. Proposed Alternative Table 2: Current ISI Intervals and License Periods 11 Plant/Unit Current License Period End Date Date of Last Category B-G-1 Examination Duration of Alternative (years)
Brunswick Unit 1 (BWR) 09/08/2036 03/30/2018 18 Brunswick Unit 2 (BWR) 12/27/2034 04/11/2017 17 Catawba Unit 1 (PWR) 12/05/2043 12/05/2015 28 Catawba Unit 2 (PWR) 12/05/2043 03/19/2015 28 H. B. Robinson Unit 2 (PWR) 07/31/2030 02/17/2012 18 McGuire Unit 1 (PWR) 03/03/2041 03/31/2013 27 McGuire Unit 2 (PWR) 03/03/2043 09/19/2015 27 Oconee Unit 1 (PWR) 02/06/2033 11/12/2014 18 Oconee Unit 2 (PWR) 10/06/2033 10/22/2015 17 Oconee Unit 3 (PWR) 07/19/2034 04/30/2016 18 Shearon Harris Unit 1 (PWR) 10/24/2046 04/27/2009 37
IV. EPRI Report Technical Basis for Dukes Request:
The technical basis for Dukes Request for Alternative is contained in an EPRI Report o EPRI Report 3002014589, Technical Basis for Optimization of the Volumetric Examination Frequency for Reactor Vessel Studs. EPRI, Palo Alto, CA: November, 2018 (publicly available)
This report is specific to Item No. B6.20 inspections for RPV studs, and includes the following:
(1) Introduction
Describes the overall approach and provides the scope of the document and its applicability. Background is also provided for RPV studs (2) Degradation Mechanisms and Experience (3) Stress Analysis (4) Flaw Tolerance Assessment (5) Conclusions and Plant-Specific Applicability Brief summaries of Sections (2) through (5) of the report follow 12
IV. EPRI Report (2) Degradation Mechanisms and Experience
Describes the literature review performed to identify degradation mechanisms that may contribute to RPV stud degradation and flaw propagation
Based on the 2007 expert panel on Proactive Materials Degradation, three degradation mechanisms were considered for RPV studs: erosion corrosion (steam cutting), fatigue, and stress corrosion cracking (SCC)
The most significant degradation issue was SCC due to Molykote lubricant combined with elevated strength of bolting materials Maximum tensile strength for existing industry studs is generally less than 170 ksi Procedurally prohibited Molykote lubricant decades ago Generic Safety Issue 109 on RV Closure was resolved and closed by NRC
Stuck studs most prevalent OE - not considered applicable to degradation issues
A description of current examination requirements is also provided
Requires volumetric examination every Interval
Surface examination may be substituted for volumetric when studs are removed PWR studs are typically removed every outage BWR studs are typically not removed every outage 13
IV. EPRI Report (3) Stress Analysis (1/2)
Performed generic finite element analysis of the closure region including the closure head and flange, upper vessel and flange, and closure stud and nut/washer stack PWR and BWR models developed and evaluated separately
Substantial differences in diameter and thickness for RPV head in PWR vs BWR
These differences play a prominent role in stud bending stresses Models developed using inputs provided by utilities from industry survey
Combinations of model parameters were used to identify a single bounding configuration for each of the PWR and BWR geometries
The bounding model parameter was identified as the head shell radius-to-thickness ratio (R/t)
The maximum R/t was used along with the minimum stud shank diameter
One-stud circumferential slice of the RPV closure region was modeled Conservatively simulates tensioning all studs at once - maximizes stud bending stress 14
IV. EPRI Report (3) Stress Analysis (2/2)
PWR Model BWR Model 15 Evaluated typical BWR and PWR RPV closure transients for peak stresses and maximum changes in stud stresses
PWR model transients/stress states
Preload, hydrotest, operation, heatup, cooldown, plant loading/unloading, small step increase/decrease, loss of load, loss of flow, reactor trip
BWR model transients/stress states:
Preload, hydrotest, operation, inner seal leakage, startup, shutdown, loss of feedwater pumps, preop blowdown
IV. EPRI Report (4) Flaw Tolerance Assessment
A deterministic flaw tolerance assessment was performed
No probabilistic evaluations were performed for this study; the results are based on bounding deterministic assessments
Calculated allowable crack depths
Calculated the projected fatigue crack growth for hypothetical initial fatigue cracks
Computed the time for the final flaw sizes to reach the allowable flaw sizes
The following results were calculated
PWR: 0.445" after 80 years < 1.06" maximum allowable
BWR: reaches 0.789 maximum allowable after 37.9 years Therefore, PWR studs have 80 years of flaw tolerance, BWR studs have 38 years of flaw tolerance 16 Limiting PWR Size 0.00 0.15 0.30 0.45 0.60 0.75 0.90 1.05 1.20 0
10 20 30 40 50 60 70 80 Crack Depth (in)
Time (yr)
Limiting BWR Size 0.00 0.15 0.30 0.45 0.60 0.75 0.90 1.05 1.20 1.35 1.50 0
10 20 30 40 50 60 70 80 Crack Depth (in)
Time (yr)
IV. EPRI Report (5) Conclusions and Plant-Specific Applicability
This section of the report provides conclusions from the technical evaluations and 8 criteria that are to be met for the technical bases to be considered applicable on a plant-specific basis:
1.
The RPV head inner radius-to-head thickness ratio is smaller than 34.8 for BWRs and 14.5 for PWRs, which is the bounding ratio used in the underlying evaluation 2.
The RPV head shell inner radius-to-stud diameter ratio is smaller than 22.4 for BWRs and 14.0 for PWRs, which is the bounding ratio used in the underlying evaluation 3.
The applicable transients are bounded by the transients used in the EPRI report analysis 4.
The number of transients projected through the end of the applicable operating period is less than the number of transients used in the EPRI report analysis 5.
All RPV studs remain in service and are successfully tensioned 6.
All RPV studs are fabricated from material consistent with low susceptibility to SCC 7.
RPV studs are specified as SA-540, Grades B23 or B24 material, or the RPV stud material specification is consistent with all SA-540 Grade B23/B24 requirements 8.
No leakage from the RPV flange has been observed since the most recent volumetric/surface examination 17
V. Conclusion / Schedule Conclusions Regarding the EPRI Report for Dukes Request:
The plant-specific criteria for application of the EPRI report will be satisfied for all of the Duke plants included in the Request for Alternative The requested inspection frequency extension is based on prior satisfactory inspection histories for all plants included in the request The results of the EPRI generic flaw tolerance analyses conclude that the existing ASME Section XI inspection interval of 10 years can be extended to as much as 80 years from the last examination without compromising plant or public safety Dukes Request for Alternative requests an extension for RPV stud examination to the end of the currently-licensed operating period for all plants included in the request
This extended time period ranges from 17 to 37 years, depending on the plant
These are generally a factor of 2 less than the periods established in the EPRI Report 18
V. Conclusion / Schedule Duke Energy is requesting an alternative to the examination requirements of ASME Section XI Table IWB-2500-1 (B-G-1), Examination Category B-G-1, Item B6.20. The request is for no further examinations of the studs for the remainder of the currently licensed 60 year operating periods.
Request 1 year NRC review 19
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