NL-21-0767, Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations for the 4th 10-Year Inservice Inspection Internal (VEGP-ISI-ALT-04-05, Version 1.0)
| ML21256A243 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/10/2021 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21256A242 | List: |
| References | |
| NL-21-0767 | |
| Download: ML21256A243 (64) | |
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Defects shall be removed or mitigated in accordance with the following requirements:
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Defect removal by mechanical processing shall be in accordance with IWA-4462.
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Defect removal by thermal methods shall be in accordance with IWA-4461.
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Defect removal or mitigation by welding or brazing shall be in accordance with IWA-4411.
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Defect removal or mitigation by modification shall be in accordance with IWA-4340. to NL-21-0767
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Westinghouse Non-Proprietary Class 3 CAW-21-5215 Page 1 of 3 COMMONWEALTH OF PENNSYLVANIA:
COUNTY OF BUTLER:
(1)
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(2)
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The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.
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Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable to NL-21-0767 CAW-21-5215, Westinghouse Affidavit Requesting Witholding of Proprietary Information
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Enclosed is:
CAW-21-5215 The enclosure contains information proprietary to Westinghouse Electric Company LLC (Westinghouse), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission (Commission) and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commissions regulations.
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Westinghouse Non-Proprietary Class 3 WCAP-18647-NP August 2021 Revision 0 Technical Basis for Westinghouse Embedded Flaw Repair of Vogtle Units 1 and 2 Reactor Vessel Head Penetrations to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3
- Electronically approved records are authenticated in the electronic document management system.
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA
© 2021 Westinghouse Electric Company LLC All Rights Reserved WCAP-18647-NP Revision 0 Technical Basis for Westinghouse Embedded Flaw Repair of Vogtle Units 1 and 2 Reactor Vessel Head Penetrations August 2021 Xiaolan Song*
RV/CV Design and Analysis Verifier:
Geoffrey M. Loy*
RV/CV Design and Analysis Reviewers: Anees Udyawar*
RV/CV Design and Analysis Approved: Lynn A. Patterson*, Manager RV/CV Design and Analysis to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 ii WCAP-18647-NP August 2021 Revision 0 FOREWORD This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a,c,e) associated with the brackets sets forth information which is considered proprietary.
The proprietary information and data contained within the brackets in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public disclosure in accordance with the Rules of Practice 10 CFR 2.390 and the information presented herein is safeguarded in accordance with 10 CFR 2.390. Withholding of this information does not adversely affect the public interest.
This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the Advisory Committee on Reactor Safeguards (ACRS) without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Companys proprietary interests.
Several locations in this topical report contain proprietary information. Proprietary information is identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is provided, using a standardized system. The proprietary brackets are labeled with three (3) different letters, a, c, and e per Westinghouse policy procedure BMS-LGL-84, which stand for:
- a. The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc. The prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.
- c. The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
- e. The information reveals aspects of past, present, or future Westinghouse-or customer-funded development plans and programs of potential commercial value to Westinghouse.
The proprietary information in the brackets has been provided in the proprietary version of this report (WCAP-18647-P Revision 0).
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Westinghouse Non-Proprietary Class 3 iii WCAP-18647-NP August 2021 Revision 0 RECORD OF REVISIONS Revision Date Revision Description 0
August 2021 Original Issue to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 iv WCAP-18647-NP August 2021 Revision 0 TABLE OF CONTENTS FOREWORD................................................................................................................................................ ii
1
INTRODUCTION........................................................................................................................ 1-1
2
TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO PENETRATION NOZZLES......................................................................................................... 2-1
2.1
ACCEPTANCE CRITERIA......................................................................................................... 2-1
2.1.1
Axial Flaws................................................................................................................................... 2-2
2.1.2
Circumferential Flaws................................................................................................................... 2-3
2.2
METHODOLOGY....................................................................................................................... 2-4
2.2.1
Geometry and Material................................................................................................................. 2-5
2.2.2
Finite Element Analysis................................................................................................................ 2-5
2.2.3
Loading Conditions....................................................................................................................... 2-6
2.2.4
Allowable Flaw Size Determination............................................................................................. 2-9
2.2.5
Stress Intensity Factors................................................................................................................. 2-9
2.2.6
Fatigue Crack Growth Prediction............................................................................................... 2-10
2.3
FRACTURE MECHANICS ANALYSIS RESULTS................................................................. 2-11
2.3.1
Maximum End-of-Evaluation Period Flaw Sizes....................................................................... 2-11
2.3.2
Allowable Initial Flaw Sizes for Penetration Nozzles................................................................ 2-12
3
TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO ATTACHMENT J-GROOVE WELD........................................................................................... 3-1
3.1
ACCEPTANCE CRITERIA......................................................................................................... 3-1
3.1.1
Section XI Appendix K................................................................................................................. 3-1
3.1.2
Primary Stress Limits.................................................................................................................... 3-2
3.2
METHODOLOGY....................................................................................................................... 3-2
3.2.1
Geometry and Material................................................................................................................. 3-3
3.2.2
Loading Conditions....................................................................................................................... 3-6
3.2.3
Stress Intensity Factors................................................................................................................. 3-6
3.2.4
J-R curve for Reactor Vessel Closure Head Material.................................................................... 3-7
3.2.5
Applied J-Integral......................................................................................................................... 3-8
3.2.6
Fatigue Crack Growth Prediction................................................................................................. 3-9
3.3
FRACTURE MECHANICS ANALYSIS RESULTS................................................................. 3-10
3.3.1
Results for Applied J-Integral and J-R Curve............................................................................. 3-10
3.3.2
Results for Fatigue Crack Growth into the Reactor Vessel Head............................................... 3-15
3.3.3
Results for Fatigue Crack Growth into the Repair Weld............................................................ 3-16
4
SUMMARY
AND CONCLUSIONS............................................................................................ 4-1
5
REFERENCES............................................................................................................................. 5-1
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Westinghouse Non-Proprietary Class 3 1-1 WCAP-18647-NP August 2021 Revision 0 1
INTRODUCTION Leakage and cracks have been reported from the reactor vessel closure head penetration nozzles in a number of plants that resulted in repairs or prompted the replacement of the reactor vessel closure head. The degradation of the reactor vessel closure head penetration nozzles increases the probability of a more significant loss of reactor coolant pressure boundary. This has led to the issuance of various regulatory requirements and guidelines in the United States imposing additional volumetric and surface examinations to supplement the existing visual inspections of the reactor vessel closure head as well as the penetration nozzles. The presence of axial cracks extending above and below the head penetration nozzle attachment J-groove welds was discovered in some of the leaking penetration nozzles. The cause of these axially oriented cracks has been determined to result from primary water stress corrosion cracking (PWSCC) that is driven by both the steady state operating stress and the residual stress resulting from the weld fabrication process. [
]a,c,e As a part of the inspection and repair efforts associated with the reactor vessel closure head inspection program at Vogtle Unit 1 and Unit 2, engineering evaluations have been performed in this report to support plant-specific use of the Westinghouse embedded flaw repair process in the repair of unacceptable flaws.
[
]a,c,e
[
]a,c,e Engineering evaluations were performed to determine the maximum flaw sizes that would satisfy the requirements in Section XI of the ASME Code [1] and be suitable to support the weld repair process.
The results presented in this report would enable the weld repair team to effectively determine the appropriate repair method.
Section XI repair rules allow the use of grinding to remove flaws, regardless of the edition of the Code.
The only requirement is to ensure that the excavated region still meets the stress limits of the original construction code, which wasSection III. Evaluations were performed in [2] to provide repair guidelines that may be used for removal of defects found on the surfaces of J-groove attachment welds and associated nozzles for the Vogtle Units 1 and 2 control rod drive mechanism (CRDM) and instrumentation port penetrations.
The technical basis of the embedded flaw repair process is documented in WCAP-15987-P [3], which has been reviewed and accepted by the NRC. The staff also concluded that WCAP-15987-P [3] is acceptable for referencing in licensing applications. As discussed in Appendix C of WCAP-15987-P [3], Westinghouse has developed the following three repair scenarios/method to address the most common types of flaws during the vessel head inspection:
Scenario 1: Axial or circumferential crack in the penetration nozzle inner surface to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 1-2 WCAP-18647-NP August 2021 Revision 0 Scenario 2: Axial crack in the penetration J-groove weld Scenario 3: Axial or circumferential crack in the penetration nozzle outer surface Figure 1-1 shows the repair for Scenario 1, and Figure 1-2 shows the repair for Scenario 2 and 3.
The purpose of this report is to provide plant-specific technical basis for the use of the embedded flaw repair process and to confirm that Vogtle Unit 1 and Unit 2 meet the criteria for application of the embedded flaw repair process stated in Appendix C of WCAP-15987-P [3]. Engineering evaluations were performed and the results are presented in this report to provide the maximum allowable initial embedded flaw sizes that could be repaired using the Westinghouse embedded flaw repair process and would satisfy the requirements in Section XI of the ASME Code [1]. The ASME Section XI Code of record for Vogtle Unit 1 and Unit 2 is 2007 Edition with 2008 Addenda [1]. Note that the methodology used in this report from the 2007 Edition with 2008 Addenda is the same up to the 2017 Edition of ASME Section XI Code, which is the most recent ASME Code edition approved by the NRC. The results presented in this report would support the use of the Westinghouse embedded flaw repair process as the repair option for the Vogtle Units 1 and 2 reactor vessel head penetration nozzles.
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Westinghouse Non-Proprietary Class 3 1-3 WCAP-18647-NP August 2021 Revision 0 Figure 1-1 General Schematic of the Embedded Flaw Repair to a Flaw in the Head Penetration Tube Inside Surface to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 1-4 WCAP-18647-NP August 2021 Revision 0 Figure 1-2 General Schematic of the Embedded Flaw Repair to a Flaw in the Head Penetration Tube Outside Surface, or to a Flaw in the Attachment Weld (J-Groove Weld) to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-1 WCAP-18647-NP August 2021 Revision 0 2
TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO PENETRATION NOZZLES This section provides a discussion on the technical basis for the use of the embedded flaw repair method for a flawed head penetration nozzle (i.e., flaws on the ID or OD of the head penetration nozzles (Scenario 1 and Scenario 3)). The technical basis for the use of the embedded flaw repair method for the flawed head attachment weld (Scenario 2) is provided in Section 3.
[
]a,c,e
[
]a,c,e The evaluation of the embedded flaw repair for the axial or circumferential crack on the penetration inner surface (Scenario 1) or outer surface (Scenario 3) began with the determination of an allowable end-of-evaluation period flaw size based on the acceptance criteria described in Section 2.1 for a flaw postulated to remain in the repaired penetration nozzle. [
]a,c,e 2.1 ACCEPTANCE CRITERIA Rapid, non-ductile failure is possible for ferritic materials at low temperatures, but is not applicable to the nickel-base alloy head penetration nozzle material, Alloy 600. Nickel-base alloy material is a high toughness material and plastic collapse would be the dominant mode of failure. [
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Westinghouse Non-Proprietary Class 3 2-2 WCAP-18647-NP August 2021 Revision 0
]a,c,e 2.1.1 Axial Flaws For axial flaws the allowable flaw depth is given by [
]a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-3 WCAP-18647-NP August 2021 Revision 0 2.1.2 Circumferential Flaws For circumferential flaws [ to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-4 WCAP-18647-NP August 2021 Revision 0
]a,c,e 2.2 METHODOLOGY The evaluation assumed that an unacceptable flaw has been detected on the surface of a penetration nozzle and that the embedded flaw repair process is used to seal the flaw from further exposure to the primary water environment. The evaluation began with the determination of an allowable end-of-evaluation period flaw size based on the acceptance criteria described in Section 2.1 for a flaw postulated to remain in the repaired penetration nozzle. [
]a,c,e The maximum initial flaw size in a penetration nozzle that can be repaired using to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-5 WCAP-18647-NP August 2021 Revision 0 the embedded flaw repair process can then be determined [
]a,c,e The following provides a discussion of the geometry, loading conditions, thermal transient stress analysis, and
[
]a,c,e used in the development of the plant specific technical basis for the embedded flaw repair process.
2.2.1 Geometry and Material There are seventy eight CRDM head penetration nozzles in the reactor vessel upper closure head with the same nozzle geometry but at different locations in the closure head [4.a and 5.a]. The outside radius and thickness for all Alloy 600 tubes are [
]a,c,e. The CRDM nozzle material is [
]a,c,e.
2.2.2 Finite Element Analysis The distributions of transient thermal and pressure stresses [
]a,c,e Reference [6] considers the welding residual stresses associated with original nozzle installation.
Subsequent to the welding residual stress analysis, the stresses that result from the [
]a,c,e in the presence of welding residual conditions are calculated. [
]a,c,e, including the welding residual stresses associated with original nozzle installation. [
]a,c,e Figure 2-2 shows the location of the stress cuts. [
]a,c,e of the circumferential and axial cracks postulated on the inside or outside of the nozzles. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-6 WCAP-18647-NP August 2021 Revision 0 Figure 2-2 Finite Element Model with Analytical Stress Cuts Identified 2.2.3 Loading Conditions The requirement for determining the maximum allowable end-of-evaluation period flaw size using the rules of Section XI is that the governing loadings from the normal, upset, emergency, and faulted conditions be considered. This is necessary because, as discussed in Section 2.1, different safety margins are used for the normal/upset conditions and the emergency/faulted conditions. A lower safety factor is used to reflect the lower probability of occurrence for the emergency/faulted conditions.
[
]a,c,e a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-7 WCAP-18647-NP August 2021 Revision 0
[
]a,c,e The thermal transients that occur in the upper head region are relatively mild because most of the water in the head region has already passed through the core region.
The flow in the upper head region is low compared to other regions of the reactor vessel, which mutes the effects of the operating thermal transients. The normal, upset and test transients considered for Vogtle Unit 1 and Unit 2 reactor vessel analyses [8 and 9] and the design cycles of the transients from Table 3.9.N.1-1 of Vogtle plants final safety analysis report (FSAR) [10] are summarized in Table 2-1. [
]a,c,e The operating licenses for both Vogtle units have been renewed and the original licensed operating terms have been extended by 20 years. The effect of the extended operating term on the number of transient cycles was evaluated as a Time-Limited Aging Analyses (TLAA) for license renewal in accordance with 10 CFR Part 54 and it was concluded in Section 19.4.2.1 of the FSAR [10] that the design cycles in Table 3.9.N.1-1 are conservative and bound 60 years of operation. Note that the fatigue crack growth evaluation performed herein will be applicable to 80 years of operation, given that the 80-year projected transients and cycles are also bounded by the design cycles.
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Westinghouse Non-Proprietary Class 3 2-8 WCAP-18647-NP August 2021 Revision 0 Table 2-1 Vogtle Unit 1 and Unit 2 Normal, Upset and Test Transients for FCG Analyses Transient Cycles(1)
Plant Heatup and Cooldown (100°F/hour) 200 Unit Loading, 0-15% of Full Power 500 Plant Loading @ 5% Full Power/min 11,200 Step Load Decrease of 10% Full Power 2,000 Step Load Increase of 10% Full Power 2,000 Plant Unloading @ 5% Full Power/min.
13,200 Unit Unloading (15% to 0%)
500 Reactor Trip with no Cooldown 230 Reactor Trip with Cooldown, No Safety Injection 160 Reactor Trip with Cooldown, Safety Injection 10 Large Step Load Decrease 200 Reduce Temp. Return to Power 2,000 Excessive Feedwater Flow 30 Control Rod Drop 80 Inadvertent Startup, Inactive Loop 10 Feedwater Cycling 2,000 Partial Loss of Flow 80 Inadvertent Depression 20 Inadvertent Safety Injection 60 Loss of Power 40 Loss of Load 80 Loop out of Service, Startup 70 Loop out of Service, Shutdown 80 Turbine Roll Test 20 Steady State Fluctuation -
Initial Fluctuations, +/-3oF, +/-25 psi 176,400(2)
Steady State Fluctuation -
Random Fluctuations, +/-0.5oF, +/-6 psi 3,000,000 Notes:
1.
Cycles are from Table 3.9.N.1-1 of Vogtle plants final safety analysis report (FSAR) [10].
2.
[
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Westinghouse Non-Proprietary Class 3 2-9 WCAP-18647-NP August 2021 Revision 0 2.2.4 Allowable Flaw Size Determination Allowable end-of-evaluation flaw sizes for axial and circumferential flaws with various aspect ratios (flaw length/flaw depth) in a CRDM penetration nozzle are calculated in accordance with the acceptance criteria discussed in Section 2.1. The allowable initial flaw sizes are subsequently determined by adjusting the allowable end-of-evaluation flaw sizes based on the results from the fatigue crack growth evaluation described in Section 2.2.6. Since the repaired flaws are embedded and sealed, they are not subjected to PWSCC.
2.2.5 Stress Intensity Factors One of the key elements in a crack growth analysis is the crack driving force or crack tip stress intensity factor, KI. This is based on the equations available in public literature. Both embedded and surface flaws are analyzed for repaired inside and outside surface flaws.
Outside and Inside Surface Flaws The stress intensity factors (SIF), KI, for the part through-wall surface cracks are calculated based on
[
]a,c,e. The stress distribution profile is represented by a 3rd order polynomial as shown below.
= 0+ 1 a t+ 2 a t
2
+ 3 a t
3 where:
0, 1, 2, and 3 are the stress profile curve fitting coefficients to be determined; a is the distance from the wall surface where the crack initiates; t is the wall thickness; and is the stress perpendicular to the plane of the crack.
The SIFs can be expressed in the general form as follows:
[
]a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-10 WCAP-18647-NP August 2021 Revision 0 Embedded Flaws The stress intensity factor calculation for an embedded flaw was based on [
]a,c,e This stress intensity factor expression for subsurface (embedded) flaws can be expressed [
]a,c,e 2.2.6 Fatigue Crack Growth Prediction With the application of the embedded flaw repair process, any postulated flaws in the reactor vessel head penetration tubes are sealed from the PWR environment; therefore, the only mechanism for crack growth would be due to fatigue crack growth.
The FCG analysis procedure involves postulating an initial flaw at the region of concern and predicting the growth of that flaw due to an imposed series of loading transients. The applied loads include pressure, thermal transients, and residual stresses. The normal and upset thermal transients as well as the associated design cycles considered in the fatigue crack growth analysis are shown in Table 2-1. The cycles are distributed evenly over 60 years of plant design life. The stress intensity factor range, ¨KI, that controls fatigue crack growth, depends on the geometry of the crack, its surrounding structure, and the range of applied stresses in the region of the postulated crack. Once ¨KI is calculated, the fatigue crack growth due to a particular stress cycle can be determined using a crack growth rate reference curve applicable to the material of the head penetration nozzle. Once the incremental crack growth corresponding to a specific transient is calculated for a small time period, it is added to the original crack size, and the analysis continues to the next time period and/or thermal transient. The procedure is repeated in this manner until all the significant analytical thermal transients and cycles known to occur in a given period of operation have been analyzed.
[
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Westinghouse Non-Proprietary Class 3 2-11 WCAP-18647-NP August 2021 Revision 0
]a,c,e 2.3 FRACTURE MECHANICS ANALYSIS RESULTS 2.3.1 Maximum End-of-Evaluation Period Flaw Sizes The maximum allowable end-of-evaluation period flaw sizes are determined for axial and circumferential surface flaws for postulated flaw aspect ratios (flaw length/flaw depth) of 2, 3, 6, and 10. The allowable flaw sizes are considered for all normal, upset, test, emergency, and faulted conditions and the most limiting allowable flaw sizes from these conditions are summarized in Table 2-2 and will be used in the generation of flaw evaluation charts.
Table 2-2 Maximum Allowable End-of-Evaluation Period Flaw Size Based on Section XI Location Aspect Ratio
(
)
Axial Allowable Flaw Size Circumferential Allowable Flaw Size a/t a (in.)
a/t a (in.)
CRDM Nozzle (t = 0.625) 2 0.75 0.469 0.75 0.469 3
0.75 0.469 0.75 0.469 6
0.75 0.469 0.61 0.381 10 0.75 0.469 0.48 0.300 to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-12 WCAP-18647-NP August 2021 Revision 0 2.3.2 Allowable Initial Flaw Sizes for Penetration Nozzles After the maximum allowable end-of-evaluation period flaw sizes are determined, [
]a,c,e First, the outside and inside surface flaws with aspect ratios of 2, 3, 6, and 10 are postulated. [
]a,c,e The results are also plotted in Figure 2-3 and Figure 2-4. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-13 WCAP-18647-NP August 2021 Revision 0 Table 2-3 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair Location Years of Operation Aspect Ratio (l/a)
Inside Surface Outside Surface Circumferential Flaw Axial Flaw Circumferential Flaw Axial Flaw a/t a (in.)
a/t a (in.)
a/t a (in.)
a/t a (in.)
Downhill Side 20 2
0.74 0.4625 0.74 0.4625 0.74 0.4625 0.73 0.4563 3
0.74 0.4625 0.71 0.4438 0.73 0.45625 0.68 0.4250 6
0.60 0.3750 0.62 0.3875 0.54 0.3375 0.51 0.3188 10 0.47 0.2938 0.52 0.3250 0.41 0.25625 0.43 0.2688 40 2
0.74 0.4625 0.73 0.4563 0.74 0.4625 0.72 0.4500 3
0.74 0.4625 0.68 0.4250 0.71 0.44375 0.63 0.3938 6
0.60 0.3750 0.56 0.3500 0.49 0.30625 0.45 0.2813 10 0.47 0.2938 0.46 0.2875 0.37 0.23125 0.37 0.2313 60 2
0.74 0.4625 0.69 0.4313 0.74 0.4625 0.71 0.4438 3
0.74 0.4625 0.65 0.4063 0.69 0.43125 0.59 0.3688 6
0.60 0.3750 0.51 0.3188 0.46 0.2875 0.41 0.2563 10 0.47 0.2938 0.42 0.2625 0.34 0.2125 0.34 0.2125 Uphill Side 20 2
0.74 0.4625 0.70 0.4375 0.74 0.4625 0.72 0.4500 3
0.73 0.4563 0.66 0.4125 0.73 0.4563 0.63 0.3938 6
0.60 0.3750 0.52 0.3250 0.55 0.3438 0.45 0.2813 10 0.47 0.2938 0.41 0.2563 0.43 0.2688 0.37 0.2313 40 2
0.74 0.4625 0.63 0.3938 0.74 0.4625 0.69 0.4313 3
0.73 0.4563 0.58 0.3625 0.71 0.4438 0.57 0.3563 6
0.60 0.3750 0.43 0.2688 0.52 0.3250 0.39 0.2438 10 0.47 0.2938 0.34 0.2125 0.4 0.2500 0.32 0.2000 60 2
0.74 0.4625 0.58 0.3625 0.74 0.4625 0.67 0.4188 3
0.73 0.4563 0.55 0.3438 0.7 0.4375 0.53 0.3313 6
0.60 0.3750 0.38 0.2375 0.48 0.3000 0.36 0.2250 10 0.47 0.2938 0.29 0.1813 0.37 0.2313 0.29 0.1813 to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-14 WCAP-18647-NP August 2021 Revision 0 Figure 2-3 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair - Downhill Side to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 2-15 WCAP-18647-NP August 2021 Revision 0 Figure 2-4 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair - Uphill Side to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-1 WCAP-18647-NP August 2021 Revision 0 3
TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO ATTACHMENT J-GROOVE WELD This section provides a discussion on the technical basis for the use of the embedded flaw repair method for the flawed head attachment weld (Scenario 2). [
]a,c,e A flaw evaluation was carried out by analyzing a planar flaw in the reactor vessel head the size of the J-groove weld size.
3.1 ACCEPTANCE CRITERIA 3.1.1 Section XI Appendix K The evaluation procedure and acceptance criteria used to demonstrate structural integrity of the reactor vessel closure head is contained in Appendix K of ASME Section XI Code [1] as well as Regulatory Guide 1.161 [13]. Although the original purpose of Appendix K was to evaluate reactor vessels with low upper shelf fracture toughness, the general approach in paragraph K-4220 is equally applicable to any region of the reactor vessel where the fracture toughness can be described with elastic plastic parameters. This approach to evaluate the integrity of a nuclear vessel has been developed over several years, and has been illustrated with a number of example problems [14] to demonstrate its use. The extension of this methodology to issues other than the low shelf fracture toughness issue is appropriate when service conditions (temperature) promote ductile behavior. The closure head region of the reactor vessel has the operating temperature of about 558 ºF. This would result in ductile behavior and therefore the use of elastic-plastic fracture mechanics method is appropriate.
The acceptance criteria are to be satisfied for each category of transients, namely, Service Load Level A (normal), Level B (upset, including test), Level C (emergency) and Level D (faulted) conditions and two criteria discussed below must be satisfied.
The first criterion is that the crack driving force must be shown to be less than the material toughness as follows:
Japplied < Jmaterial where Japplied is the J-integral value calculated for the postulated flaw under the applicable Service Level condition and Jmaterial is the J-integral characteristic of the material resistance to ductile tearing at a crack extension of 0.1 inch. For Level A and B conditions, a safety factor of 1.15 is conservatively applied to the Japplied per Reg Guide 1.161 [13] and ASME Section XI Appendix K Article K-4220 of ASME Section XI Code [1]. The factor of 1.15 needs only to be applied on pressure, however, in this evaluation it is applied to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-2 WCAP-18647-NP August 2021 Revision 0 to the J-integral calculated from the transient and residual stresses in addition to the normal operating pressure. For Level C and D conditions, the safety factor on Japplied is 1.0.
The second criterion is that the flaw must also be stable under ductile crack growth as follows:
a J applied w
w da dJmaterial at Japplied = Jmaterial
- where, Jmaterial = J-integral resistance to ductile tearing for the material.
a J applied w
w
Partial derivative of the applied J-integral with respect to flaw depth, a da dJmaterial
Slope of the J-R curve For Level A and B conditions, a safety factor of 1.25 is conservatively applied to the Japplied per Reg Guide 1.161 [13] and ASME Section XI Appendix K Article K-4220 of ASME Section XI Code [1]. The factor of 1.25 needs only to be applied on pressure, however, in this evaluation it is conservatively applied to the transient and residual stresses in addition to the normal operating pressure. For Level C and D conditions, the safety factor on Japplied is 1.0. Flaw stability is verified when the slope of the applied J-integral curve is less than the material J-integral curve at the point on J-R curve where the two curves intersect.
3.1.2 Primary Stress Limits In addition to satisfying the Section XI criteria, the primary stress limits of paragraph NB-3000 in Section III of the ASME Code [15] must be satisfied. The effects of a local area reduction that is equivalent to the area of the postulated flaw in the vessel head attachment weld must be considered by increasing the membrane stresses to reflect the reduced cross section. The allowable flaw depth was determined by evaluating the primary stress of the spherical head with reduced wall thickness using the maximum pressure of [
]a,c,e for all service conditions. The results show the allowable flaw depth is 1.933 inches.
3.2 METHODOLOGY Since the depth of a flaw in the attachment weld cannot be detected using current technology, the engineering evaluation for the embedded flaw repair process was performed to demonstrate the stability of an assumed hypothetical flaw that encompasses the entire attachment J-groove weld region in the reactor vessel head near the penetration nozzle. The criteria used to demonstrate the stability and structural integrity of the reactor vessel closure head is described in Section 3.1.1 as per the ASME Code [1] and Regulatory Guide 1.161 [13].
After the implementation of the embedded flaw repair process, [
]a,c,e That is, the flaw depth at the end of evaluation period should be below the 1.933 inches as to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-3 WCAP-18647-NP August 2021 Revision 0 determined in Section 3.1.2 such that primary stress limit of the ASME Code Section III, paragraph NB-3000 [15] is satisfied. In addition, it needs to be shown that the postulated flaw will not grow through the repair layer.
3.2.1 Geometry and Material The reactor vessel head is made of [
]a,c,e, with the following geometry:
[
]a,c,e The reactor vessel upper head nozzle attachment weld geometry for the nozzles used in this calculation is tabulated in Table 3-1 for the case without the weld fillet as shown in Figure 3-1. The weld dimensions in Table 3-1 are used for the fatigue crack growth and J-integral analyses for postulated flaws in the reactor vessel head. The height and width of the J-groove weld configurations are the built-up dimensions.
The weld depths for all penetration nozzles on the uphill and downhill sides are also provided based on the UT scanning of the Alloy 600 tubes [16]. The weld depth dimension a, as shown in Figure 3-1 and Table 3-2, includes the weld fillet and butter thickness. The weld dimensions in Table 3-2 are used in the fatigue crack growth analysis for the growth of postulated flaws through the weld repair layer. [
]a,c,e these flaw depths bound all penetration row weld depths a for Vogtle Unit 1 and Unit 2.
to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-4 WCAP-18647-NP August 2021 Revision 0 a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-5 WCAP-18647-NP August 2021 Revision 0 Figure 3-1 Definition of J-Groove Weld Dimensions a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-6 WCAP-18647-NP August 2021 Revision 0 3.2.2 Loading Conditions For the normal/upset condition, the reactor vessel closure head structural integrity evaluation is performed for all the transients in Table 2-1, and [
]a,c,e For the emergency and faulted condition evaluation, [
]a,c,e There are many head penetrations in the reactor vessel upper head, and [
]a,c,e The distribution of residual, transient thermal, and pressure stresses in the closure head region is obtained from detailed three-dimensional elastic-plastic finite element analyses of the head penetration nozzle region [6]. [
]a,c,e 3.2.3 Stress Intensity Factors J-Groove Weld Double Corner Crack in the Reactor Vessel Head Since the depth of a flaw in the attachment weld cannot be detected using current technology, it is conservatively assumed that the flaw in the attachment weld extends radially over the entire attachment weld. [
]a,c,e The stress intensity factor expression shown above is applicable for a range of flaw shapes, with the depth of the flaw defined as a, and the width of the flaw defined as c, as shown in Figure 3-2. This flexibility is necessary because this expression can be applied to different attachment J-groove weld shapes for Vogtle Units 1 and 2 closure head penetrations as shown in Table 3-1. The attachment J-groove weld shapes were based on the J-groove geometry shown in the head penetration nozzle drawings for Vogtle Units 1 and 2 [4 and 5]. [
]a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-7 WCAP-18647-NP August 2021 Revision 0 Figure 3-2 Corner Crack Geometry Embedded Flaw in the Reactor Vessel Head
[
]a,c,e The details of the method is discussed in Section 2.2.5 and thus not repeated here.
3.2.4 J-R curve for Reactor Vessel Closure Head Material One of the most important pieces of information for fracture toughness for pressure vessel and piping materials is the J-R curve of the material. The J-R stands for material resistance to crack extension, as represented by the measured J-integral value versus crack extension. Simply put, the J-R curve to cracking resistance is as significant as the stress-strain curve to the load-carrying capacity and the ductility of a material. Both the J-R curve and stress-strain curves are properties of a material.
[ to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-8 WCAP-18647-NP August 2021 Revision 0
]a,c,e Neutron irradiation has been shown to produce embrittlement that reduces the toughness properties of the reactor vessel ferritic steel material. The irradiation levels are very low in the reactor vessel closure head region and therefore the fracture toughness will not be measurably affected.
3.2.5 Applied J-Integral For small scale yielding, Japplied of a crack can be calculated by the Linear Elastic Fracture Mechanics (LEFM) method based on the crack tip stress intensity factor, KI, calculated as per Section 3.2.3. However, a plastic zone correction must be performed to account for the plastic deformation at the crack tip similar to the approach in Regulatory Guide 1.161 [13]. The plastic deformation ahead of the crack front is then regarded as a failed zone and the crack size is, in effect, increased. The KI-values can be converted to Japplied by the following equation:
where Kep is the plastic zone corrected K-value, and Ec=E/(1-Q2) for plane strain, E = Youngs Modulus, and Q = Poissons Ratio.
Kep is equal to the elastically calculated KI-value based on the plastic zone adjusted crack depth or size.
The plastic zone size, rp, is calculated by 1
6
where Sy is the yield strength of the material.
Assume that the crack depth is ao, the Kep can now be calculated based on a new crack length, ao + rp. For small scale yielding, this can be simplified as Kep = f KI where
Once the J-applied is calculated, stability for the postulated flaw in the attachment J-groove weld can be determined using the methodology described in Section 3.1.1. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-9 WCAP-18647-NP August 2021 Revision 0 3.2.6 Fatigue Crack Growth Prediction With the application of the embedded flaw repair process, any postulated flaws in the reactor vessel head penetration tubes or the attachment weld are sealed from the PWR environment; therefore, the only mechanism for crack growth would be due to fatigue.
The FCG analysis procedure involves postulating an initial flaw at the region of concern and predicting the growth of that flaw due to an imposed series of loading transients, using the same approach described in Section 2.2.6. The FCG curves used for [
]a,c,e and the embedded flaw beneath the repair weld are discussed below.
FCG Curve for the Reactor Vessel Closure Head: Carbon and Low Alloy Ferritic Steel The crack growth rate curves used in the analyses for [
]a,c,e are taken directly from Appendix A in the ASME Section XI Code [1] for ferritic steel material. With the repair weld any potential flaws in the J-groove weld (Alloy 182) are sealed from the primary water environment and the only applicable growth mechanism is fatigue crack growth in air environment; therefore, the analysis is performed for a surface flaw based on the limiting crack growth rate reference curve of the air environment. This curve is a function of the applied stress intensity factor range (KI) and the R ratio, which is the ratio of the minimum to maximum stress intensity factor during a thermal transient. The crack growth equation is given below:
where n is the slope of the log (da/dN) versus log (¨KI) curve and is equal to 3.07 for subsurface flaws.
Parameter Co is a scaling constant:
0 1.99 10 where ¨Kth is the threshold ¨KI value below which the fatigue crack growth rate is negligible and S is a scaling parameter. Both ¨Kth and S are a function of the R ratio (Kmin/Kmax). The calculation of crack tip stress intensity factor range (¨KI) also changes with R ratio when.
5.0 0 5.01 0.8 0 1.0 The calculation of crack tip stress intensity factor range (¨KI) also changes with R ratio when.
The calculation of S and for different R ratio ranges is summarized below:
x For 0 R 1 S = 25.72(2.88-R)-3.07 and ¨KI = Kmax - Kmin x
For R < 0 and 1.12 S=1 and ¨KI = Kmax - Kmin x
For -2 R 0 and 1.12 S=1 and ¨KI = Kmax x
For R<-2 and 1.12 S=1 and ¨KI = (1-R)Kmax/3 to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-10 WCAP-18647-NP August 2021 Revision 0
[
]a,c,e Note that a condition is imposed on A-4300(b)(1) of ASME Code Section XI in 10CFR 50.55a Codes and Standards and a factor of 0.8 is applied to the limit in defined in A-4300 of the ASME Code Section XI [1].
FCG Curve for the Repair Weld, Alloy 52/52M, Below the J-Groove Attachment Weld
[
]a,c,e 3.3 FRACTURE MECHANICS ANALYSIS RESULTS 3.3.1 Results for Applied J-Integral and J-R Curve For the J-integral calculation, the key aspects of the analysis are to demonstrate that the magnitude of J-applied is less than J-material at 0.1 inch crack extension, and the slope of the J-material curve is greater than the slope of the J-applied curve at the intersection of the Jmat and Japplied curves. This evaluation is performed for the postulated flaws encompassing the J-groove welds at all the nozzle locations. The weld dimensions are shown in Table 3-1. The results shows that for all the nozzle locations, the applied J-integral is less than material J-integral at 0.1 inch crack extension, as shown in Table 3-3 and Table 3-4. The slope of the J-material curve is also greater than the slope of the J-applied curve at the intersection of the J-applied and J-material curves for all the locations. Figures 3-3 and 3-4 show the plots for the penetration nozzle locations with the highest J-applied at 0.1 inch crack extension for Level A/B and Level C/D conditions, respectively. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-11 WCAP-18647-NP August 2021 Revision 0 Table 3-3 J-Integral Results for 0.1 inch Crack Extension on Downhill and Uphill Sides -
Level A/B Unit Pen. No. Penetration Angle Downhill Uphill Japplied(1)
(kip-in/in2)
Jmaterial(1)
(kip-in/in2)
Japplied(1)
(kip-in/in2)
Jmaterial(1)
(kip-in/in2) 1 1
0 0.573 1.156 0.809 1.156 2-5 11.4 0.420 1.156 0.771 1.156 6-9 16.2 0.577 1.156 0.790 1.156 10-13 18.2 0.581 1.156 0.798 1.156 14-17 23.3 0.610 1.156 0.814 1.156 18-21 24.8 0.611 1.156 0.841 1.156 22-29 26.2 0.617 1.156 0.849 1.156 30-37 30.2 0.637 1.156 0.877 1.156 38-41 33.9 0.671 1.156 0.924 1.156 42-49 35.1 0.679 1.156 0.938 1.156 50-53 36.3 0.688 1.156 0.953 1.156 54-61 38.6 0.701 1.156 0.982 1.156 62-65 44.3 0.739 1.156 1.083 1.156 66-73 45.4 0.752 1.156 1.104 1.156 74-78 48.7 0.779 1.156 1.113 1.156 1.194 1.265(2) 2 1
0 0.582 1.511 0.824 1.511 2-5 11.4 0.609 1.511 0.727 1.511 6-9 16.2 0.639 1.511 0.713 1.511 10-13 18.2 0.645 1.511 0.708 1.511 14-17 23.3 0.674 1.511 0.696 1.511 18-21 24.8 0.693 1.511 0.706 1.511 22-29 26.2 0.699 1.511 0.704 1.511 30-37 30.2 0.729 1.511 0.701 1.511 38-41 33.9 0.775 1.511 0.716 1.511 42-49 35.1 0.785 1.511 0.718 1.511 50-53 36.3 0.802 1.511 0.721 1.511 54-61 38.6 0.830 1.511 0.728 1.511 62-65 44.3 0.921 1.511 0.762 1.511 66-73 45.4 0.949 1.511 0.771 1.511 74-78 48.7 1.018 1.511 0.795 1.511 Notes:
1.
The applied and material J-integrals are conservatively calculated at maximum Level A/B temperature of [
]a,c,e unless otherwise noted.
2.
The material J-integrals calculated at the temperature of [
]a,c,e, which is closer to the actual temperature when the corresponding maximum applied J-integrals occur. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-12 WCAP-18647-NP August 2021 Revision 0 Table 3-4 J-Integral Results for 0.1 inch Crack Extension on Downhill and Uphill Sides -
Level C/D Unit Pen. No. Penetration Angle Downhill Uphill Japplied(1)
(kip-in/in2)
Jmaterial(1)
(kip-in/in2)
Japplied(1)
(kip-in/in2)
Jmaterial(1)
(kip-in/in2) 1 1
0 0.561 1.111 0.699 1.111 2-5 11.4 0.412 1.111 0.644 1.111 6-9 16.2 0.566 1.111 0.656 1.111 10-13 18.2 0.570 1.111 0.661 1.111 14-17 23.3 0.606 1.111 0.668 1.111 18-21 24.8 0.608 1.111 0.693 1.111 22-29 26.2 0.614 1.111 0.698 1.111 30-37 30.2 0.637 1.111 0.717 1.111 38-41 33.9 0.679 1.111 0.757 1.111 42-49 35.1 0.689 1.111 0.767 1.111 50-53 36.3 0.699 1.111 0.779 1.111 54-61 38.6 0.718 1.111 0.807 1.111 62-65 44.3 0.769 1.111 0.912 1.111 66-73 45.4 0.793 1.111 0.934 1.111 74-78 48.7 0.830 1.111 1.024 1.111 2
1 0
0.572 1.453 0.714 1.453 2-5 11.4 0.591 1.453 0.607 1.453 6-9 16.2 0.626 1.453 0.592 1.453 10-13 18.2 0.633 1.453 0.587 1.453 14-17 23.3 0.674 1.453 0.572 1.453 18-21 24.8 0.707 1.453 0.583 1.453 22-29 26.2 0.712 1.453 0.580 1.453 30-37 30.2 0.748 1.453 0.574 1.453 38-41 33.9 0.818 1.453 0.587 1.453 42-49 35.1 0.828 1.453 0.588 1.453 50-53 36.3 0.851 1.453 0.589 1.453 54-61 38.6 0.885 1.453 0.594 1.453 62-65 44.3 0.999 1.453 0.623 1.453 66-73 45.4 1.043 1.453 0.631 1.453 74-78 48.7 1.127 1.453 0.650 1.453 Note:
1.
The applied and material J-integrals are conservatively calculated at maximum Level C/D temperature of [
]a,c,e unless otherwise noted. to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-13 WCAP-18647-NP August 2021 Revision 0 Figure 3-3 Comparison of the Slopes for Applied and Material J-Integral versus Crack Depth Curves for the Case with the Highest Japplied at 0.1 inch Crack Extension - Level A/B Conditions to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-14 WCAP-18647-NP August 2021 Revision 0 Figure 3-4 Comparison of the Slopes for Applied and Material J-Integral versus Crack Depth Curves for the Case with the Highest Japplied at 0.1 inch Crack Extension - Level C/D Conditions to NL-21-0767 WCAP-18647-NP, Revision 0
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Westinghouse Non-Proprietary Class 3 3-15 WCAP-18647-NP August 2021 Revision 0 3.3.2 Results for Fatigue Crack Growth into the Reactor Vessel Head The FCG into the reactor vessel head is considered for the postulated cracks with the initial flaw size based on the J-groove weld depth from Table 3-1. The weld dimensions in Table 3-1 show the following:
1.
For downhill side, [
]a,c,e 2.
For uphill side, [
]a,c,e Therefore, the fatigue crack growth is performed for these two locations (i.e., the outermost nozzles), which bound all the other penetration nozzles. It is assumed that the initial aspect ratio is held constant as the flaw grows through the reactor head wall thickness.
The stress intensity factor is conservatively calculated based on [
]a,c,e and the fatigue crack growth law for the reactor vessel head carbon steel material described in Section 3.2.6 is used. The FCG results are shown in Figure 3-5, which shows that the postulated flaw will not reach the reactor vessel head primary stress limit (1.933 inches) after 60 years of growth.
Figure 3-5 Fatigue Crack Growth Prediction into the Reactor Vessel Shell for Postulated Flaws in the J-Groove Welds for the Bounding Penetration Angles of 48.7° on the Downhill and Uphill Sides Primary Stress Limit of 1.933 inches Most Limiting Crack Growth at Uphill Side Most Limiting Crack Growth at Downhill Side to NL-21-0767 WCAP-18647-NP, Revision 0
- This record was final approved on 8/26/2021 4:41:33 PM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 3-16 WCAP-18647-NP August 2021 Revision 0 3.3.3 Results for Fatigue Crack Growth into the Repair Weld The attachment weld (J-groove) repair is performed by [
]a,c,e The attachment weld is thus sealed, and the thickness of the reactor vessel shell is locally increased by [
]a,c,e In order to determine the durability of the repair weld, an embedded flaw based on the J-Groove weld geometry is postulated, which starts from [
]a,c,e beneath the free surface. The postulated flaw is an axial flaw with the aspect ratio (flaw length/flaw depth) of 2. This aspect ratio of 2 bounds all the aspect ratios for the uphill and downhill side attachment weld dimensions shown in Table 3-1. For the FCG analysis, the initial total flaw depth (2a) is assumed equal to the maximum uphill and downhill weld depths [
]a,c,e in Table 3-2. The crack growth results are summarized in Table 3-5 and it shows that the structural integrity of the repaired weld layer is expected to be maintained for at least 20 years of service life.
Table 3-5 Growth of Embedded Axial Flaw in J-Groove Weld Location Year Half Crack Depth (inch)
Remaining Repair Weld Thickness (inch)
Uphill Side 0
[
]a,c,e
[
]a,c,e 10
[
]a,c,e
[
]a,c,e 20
[
]a,c,e
[
]a,c,e Downhill Side 0
[
]a,c,e
[
]a,c,e 10
[
]a,c,e
[
]a,c,e 20
[
]a,c,e
[
]a,c,e 30
[
]a,c,e
[
]a,c,e to NL-21-0767 WCAP-18647-NP, Revision 0
- This record was final approved on 8/26/2021 4:41:33 PM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 4-1 WCAP-18647-NP August 2021 Revision 0 4
SUMMARY
AND CONCLUSIONS Engineering evaluations were performed to provide plant specific technical basis for the Westinghouse embedded flaw repair process that is associated with the reactor vessel head penetration nozzle inspection and contingency repair program for Vogtle Units 1 and 2.
The technical basis for the use of the embedded flaw repair process if unacceptable flaws are detected in the head penetration nozzles is provided in Section 2. Based on the results in Section 2.3, it is determined that unacceptable axial and circumferential flaws detected on the inside surface or outside surface of a head penetration nozzle can be repaired using the embedded flaw repair process by shielding them from the primary water environment. The maximum allowable initial axial and circumferential flaw sizes that can be repaired using the Westinghouse embedded flaw repair process are shown in Table 2-3 and Figures 2-3 and 2-4 for a plant service life up to 60 years.
The technical basis for the use of the embedded flaw repair process if indications or flaws are found in the head penetration attachment J-groove welds is provided in Section 3. Based on the results shown in Section 3.3, the evaluation documented herein has demonstrated that the embedded flaw repair process is a viable method for repairing flaws found in the attachment J-groove weld. The fracture mechanics evaluation demonstrated that a flaw postulated in the J-groove weld which encompasses the entire attachment J-groove weld shape is stable under the J-integral analysis. Furthermore, the reduced wall thickness considering the 60-year fatigue crack growth of the postulated flaw will meet the reactor vessel head primary stress limit minimum thickness requirement. The fatigue crack growth through the weld overlay repair layer demonstrates that a postulated flaw in the J-groove weld will not grow through the repair layer in less than 20 years. Therefore, it is technically justified to use the embedded flaw repair process as the repair option for the reactor vessel head penetration nozzle attachment J-groove welds since the criteria for application of such a process as stated in Appendix C of WCAP-15987-P Revision 2-P-A is met. to NL-21-0767 WCAP-18647-NP, Revision 0
- This record was final approved on 8/26/2021 4:41:33 PM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-1 WCAP-18647-NP August 2021 Revision 0 5
REFERENCES 1.
ASME Boiler & Pressure Vessel Code, 2007 Edition with 2008 Addenda,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.
2.
Westinghouse Letter, LTR-SDA-21-004, Revision 0, Vogtle Units 1 and 2 General Reactor Vessel Control Rod Drive Mechanism (CRDM) Penetration J-Groove Weld Repair Dimensional Requirements, February 1, 2021.
3.
Westinghouse Report WCAP-15987-P, Revision 2-P-A, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, December 2003.
4.
Combustion Engineering Drawings for Vogtle Unit 1:
a.
[
]a,c,e 5.
Combustion Engineering Drawings for Vogtle Unit 2:
a.
[
]a,c,e 6.
[
]a,c,e 7.
[
]a,c,e 8.
[
]a,c,e 9.
[
]a,c,e
- 10. Vogtle Electric Generating Plants Final Safety Analysis Report, Rev. 19, April 2015.
- 11. [
]a,c,e
- 12. NUREG/CR-6721, ANL-01/07, Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds, April 2001. to NL-21-0767 WCAP-18647-NP, Revision 0
- This record was final approved on 8/26/2021 4:41:33 PM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 5-2 WCAP-18647-NP August 2021 Revision 0
- 13. Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 ft-lb.
- 14. Development of Criteria for Assessment of Reactor Vessels with Low Upper Shelf Fracture Toughness, Welding Research Council Bulletin 413, July 1996.
- 15. ASME Boiler & Pressure Vessel Code, 1971 Edition with Addenda through the Summer of 1972,Section III, Rules for Construction of Nuclear Power Plant Component.
- 16. [
]a,c,e
- 17. American Petroleum Institute, API 579-1/ASME FFS-1 (API 579 Second Edition), Fitness-For-Service, June 2016.
- 18. E. D. Eason, J. E. Wright, E. E. Nelson, Multivariable Modeling of Pressure Vessel and Piping J-R Data, NUREG/CR-5729, MCS 910401, RF, R5, May 1991. to NL-21-0767 WCAP-18647-NP, Revision 0