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Enclosure 4, Attachment 1: ANP-3679NP, Revision 0, Topical Report
	ML18291A836
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Site:	Surry
Issue date:	06/30/2018
From:	; Framatome, Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Reactor Regulation
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Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 NON-PROPRIETARY REFERENCE DOCUMENTS AND REDACTED VERSIONS OF PROPRIETARY REFERENCE DOCUMENTS (PUBLIC VERSION)

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 Attachment 1 .

ANP-3679NP, REVISION 0 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

framatome Low Upper-Shelf Toughness- Fracture 79 AN~-~a NP Rev1s1on 0 Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years Topical Report June 2018 Framatome Inc.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 Equivalent Margins Analysis-Analysis of Record .................................. 1-3 2.0 REGULATORY REQUIREMENTS .................................................................... 2-1 2.1 Regulatory Requirements ..................... .' ................................................. 2~1 2.2 Compliance with 10 CFR 50 Appendix G and Acceptance Criteria .................................................................................................... 2-2 2.2.1 Acceptance Criteria Levels A and B ............................................. 2-3

3.0 DESCRIPTION

OF SURRY REACTOR VESSELS ........................................... 3-1 4.0 MATERIAL PROPERTIES ............................................................................. :... 4-1 4.1 J-lntegral Resistance Model ................................................................... 4-1 4.2 Mechanical Properties of Weld Metals .................................................... 4-3 4.2.1 Mechanical Properties for the Surry Reactor Vessels .................. 4-4 5.0 FRACTURE MECHANICS ANALYSIS .............................................................. 5-1 5.1 Methodology ........................................................................................... 5-1 5.2 Procedure for Evaluating Levels A and B Service Loadings ................... 5-2 5.3 Evaluation for Flaw Extension ................................................................. 5-6 5.3.1 Reactor Vessel Shell Welds ......................................................... 5-6 5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds ....................... :................................................................... 5-7 5.4 Evaluation for Flaw Stability .................................................................... 5-7 5.4.1 Reactor Vessel Shell Welds ......................................................... 5-8 5.4.2 Reactor Vessel Transition Welds and RV Nozzle Welds ......................_..................................................................... 5-8 6.0

SUMMARY

AND CONCLUSIONS .................................................................... 6-1 6.1 Reactor Vessel Shell Welds .................................................................... 6-1 6.2 Reactor Vessel Transition Welds and RV Nozzle Welds ........................ 6-1

7.0 REFERENCES

.................................................................................................. 7-1 8.0 CERTIFICATION ............................................................................................... 8-1 APPENDIX A ...............................................................................................................A-1

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page iii List of Tables Table 3-1 Reactor Vessel Weld Locations--Copper Content and 80-Year Fluence Projections ................................................................................... 3-2 Table 4-1 Parameters in Jd Model 48 ....................................................................... 4-3 Table 4-2 Mechanical Properties of Surry RV Shell Materials ................................... 4-4 Table 5-1 Reactor Vessel Shell Dimensions and Operating Conditions .................... 5-9 Table 5-2 Flaw Evaluation Summary for RV Shell Regions ..................................... 5-10 Table 5-3 Reactor Vessel Nozzle Belt Dimensions ................................................. 5-11 Table 5-4 Flaw Evaluation Summary of Surry Upper Transition and RV Nozzle-to-Shell Welds ......................................................................................... 5-12 Table 5-5 Applied J-lntegral versus Flaw Extensions of Surry Controlling RV Shell Weld (SA-1526) .............................................................................. 5-13 Table 5-6 Mean & Lower Bound J-R Curve Values for Surry Controlling RV Shell Weld (SA-1526) .............................................................................. 5-14 Table A-1 Model 48, Range of Test Data .................................................................. A-3 Table A-2 New B&WOG Specimen Data for Model Assessment ............................... A-6 Table A-3 Jd Model Coefficients (Models 48, 58, and 68) ...................................... A-11 Table A-4 EMA Reconciliation for Limiting RV Shell Welds-Models 48 and 68 .... A-19

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years

Topical Report Page iv List of Figures Figure 3-1 Reactor Vessel-Surry Unit 1 .................................................................... 3-3 Figure 3-2 Reactor Vessel-Surry Unit 2 .................................................................... 3-4 Figure 5-1 J-lntegral versus Flaw Extension for Surry 1 and 2 Controlling Reactor Vessel Shell Weld (SA-1526) ..................................................... 5-15 Figure 5-2 J-lntegral versus Flaw Extension for Surry Axial Weld (SA-1585)

Near Transition ........................................................................................ 5-16 Figure A-1 8AW-2251A, Appendix 8, Figure 3-1 ....................................................... A-5 Figure A-2 Jct (0.1) vs Fluence 8&WOG J-R Model 48 and New Test Data (Normalized to Standard Conditions) .......................................................A-8 Figure A-3 Original and New Data and Model Fit Normalized at Standardized Conditions vs f:J.a .................................................................................... A-12 Figure A-4 Original and New Data and Model Fit Normalized at Standard Conditions vs Fluence .. :..........*............................................................... A-13 Figure A-5 Model 68 Residuals vs Fitted Values ..................................................... A-14 Figure A-6 Model 68 Standardized Residuals vs Fitted Values ............................... A-15 Figure A-7 Normal Q-Q Plot of Standardized Residuals .......................................... A-16 Figure A-8 Comparison of Models 48, 58, and 68 at Standard Conditions ............. A-18

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at 80-Years Topical Report Page v Nomenclature (If applicable)

Acronym Definition B&W Babcock and Wilcox B&WOG Babcock and Wilcox Owners Group CvUSE Charpy Upper Shelf Energy EFPY Effective Full Power Years EMA Equivalent Margins Analysis INF Inlet Nozzle Forging Jd J deformation J-R J-integral Resistance LAR License Amendment Request ONF Outlet Nozzle Forging PWROG Pressurized Water Reactor Owners Group RV Reactor Vessel RVWG Reactor Vessel Working Group SLR Subsequent License Renewal SRP Standard Review Plan Sy Yield Strength

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page vi ABSTRACT This topical report presents the results of an equivalent margins analysis (EMA) considering Levels A and B service loads for high copper Linde 80 weld metals using conservative BO-year fluence estimates for Surry Units 1 and 2. This report applies to the following Westinghouse-designed reactor vessels fabricated by B&W: Surry Units 1 and 2. Note that the Surry EMA reported herein is technically identical to the Surry EMA reported in BAW-2192P, Supplement 1, Revision 0, which was submitted to the NRC by the PWROG on December 15, 2017. That is, Sections 1.0 through 7.0 of ANP-3679P were generated by extracting Surry-specific results from Sections 1.0 through 7.0 of BAW-2192, Supplement 1, Revision 0. Appendix A to ANP-3679P, B&WOG J-R Model-Data Analysis and Empirical Model Development, is identical to Appendix A of BAW-2192, Supplement 1, Revision 0, with the exception that references to plants other than Turkey Point 3 and 4 were removed from Sections A.1, A.2, and A.4.

The analytical procedure used in this supplement is in accordance with ASME Section XI, Appendix K, Subarticle K-1200. EMA results are reported for all reactor vessel weld locations with BO-year fluence projections that exceed 1.0 E+17 n/cm 2 (E> 1.0 MeV).

The ASME Section XI, acceptance criteria for Levels A & B Service Loads for all reactor vessel shell welds are satisfied. The acceptance criteria for Levels A & B Service Loads for RV transition welds and RV nozzle welds are also satisfied:

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years Topical Report Page vii The EMA utilizes the 8&WOG J-integral resistance (J-R) Model 48 reported in 8AW-2192PA, Appendix B. Model 48 was developed based on fracture toughness test data obtained through approximately 1990, with specimen fluence that ranges from 0.0 to 8.45E18 n/cm 2 . Consistent with 8AW-2192PA, Revision 00, the 8&WOG J-R Model 48 is used for Linde 80 welds and Rotterdam welds. Eighty-year fluence estimates for Surry Units 1 and 2 at the T/10 location exceeds 8.45E18 n/cm 2 (i.e., for Surry weld SA-1526 fluence equals 8.16E18 n/cm 2 at 1/4T and 1.083E19 n/cm 2 at T/10) and use of Model 48 to estimate J-integral resistance values, including the associated model uncertainty, for BO-years is made by extrapolation of the model. To assess the model extrapolation uncertainty, Model 48 is compared to new fracture toughness test data (1990 to 2017) irradiated to fluence ranging from 8.0E18 n/cm 2 to 5.8E19 n/cm 2 . The majority of test data fell above the Model 48 mean and all of the test data fell above the Model 48 mean minus 2 standard* error band. Therefore, use of Model 48 and associated uncertainty to extrapolate J-integral resistance for 80-year fluence applications was determined to be appropriate. This assessment is reported in Appendix A herein.

To further substantiate the use of Model 48, all of the original fracture toughness data used to develop Model 48 was combined with new fracture toughness data, using the same model form, and a new Model 68 was generated. Model 68 was found to be essentially equivalent to Model 48 with respect to model mean and 2 standard errors.

The EMA results reported herein using Model 48 were reconciled to Model 68, with little or no change to the EMA results. Model 68 development and the EMA reconciliation to Model 48 are reported in Appendix A.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 1-1

1.0 INTRODUCTION

The purpose of this topical report is to present an equivalent margins analysis (EMA) considering Levels A and B service loads for high copper Linde 80 weld metals and applicable non-Linde 80 welds using fluence values expected at BO-years (subsequent license renewal--SLR). This topical report applies to the following Westinghouse-designed reactor vessels fabricated by B&W/Rotterdam: Surry Units 1 and 2. Note that the Surry EMA reported herein is technically identical to the Surry EMA reported in BAW-2192P, Supplement 1, Revision 0, which was submitted to the NRC by the PWROG on December 15, 2017. That is, Sections 1.0 through 7.0 of ANP-3679P were generated by extracting Surry-specific results from Sections 1.0 through 7.0 of BAW-2192, Supplement 1, Revision 0. Appendix A to ANP-3679P, B&WOG J-R Model-Data Analysis and Empirical Model Development, is identical to Appendix A of BAW-2192, Supplement 1, Revision 0, with the exception that references to plants other than Turkey Point 3 and 4 were removed from Sections A.1, A.2, and A.4.

Equivalent margins analyses for the plants within the scope of this report are reported for all reactor vessel weld locations with BO-year fluence projections that exceed 1.0 E+17 n/cm 2 (E> 1.0 MeV) [2]. Upper shelf energy evaluations at reactor vessel base metal locations with BO-year fluence projections greater than 1.0 E+17 n/cm 2 , if needed, will be addressed separately in the Surry Units 1 and 2 subsequent license renewal application. The EMA utilizes the B&WOG J-integral resistance (J-R) model reported in BAW-2192PA, Appendix B, [1].

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 1-2 The following groups are used for the welds within the scope of this report:

Reactor Vessel Shell Welds-circumferential and longitudinal welds within the intermediate and lower shell assemblies for Surry Units 1 and 2 (also referred to as Surry reactor vessels). There are no geometric discontinuities at these weld locations and all reactor vessel shell welds surround the effective height of the active core. These locations have historically been considered "beltline" or "beltline region" as defined by 10 CFR 50, Appendix G. All reactor vessel shell welds are Linde 80 welds with the exception of Surry Unit 2 weld R-3008 (Figure 3-2), which is a Rotterdam weld.

Transition Welds and RV Nozzle Welds-welds that are located above and below the reactor vessel shell welds that may experience 80-year fluence greater than 1.0 E+17 n/cm 2 [2] and, must consider the effects of neutron irradiation embrittlement.

In addition, the transition welds are located at geometric discontinuities (e.g., lower shell to lower head and upper shell to nozzle belt forging). These locations may or may not have been included as part of the 10 CFR 50 Appendix G [4] "beltline" definition for 60-years for the participating plants. All transition welds and RV nozzle welds (also referred to as RV nozzle-to-shell welds) are Linde 80 welds with the exception of the following: Surry Unit 1 transition weld J726 (Figure 3-1 ), Surry Unit 2 transition weld L737 (Figure 3-2), and Surry Unit 2 RV outlet nozzle-to-nozzle belt forging welds, which are Rotterdam welds.

The EMA evaluations in this report are for all weld locations expected to receive fluence

> 1.0E17 n/cm 2 [2] at 80 years. The use of the terms "beltline" and/or "extended beltline" are not used in this report.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at 80-Years Topical Report Page 1-3 The 60-year EMA summary reports for Surry Units 1 and 2 are reported in Section 1.1.

Section 2.0 provides the current NRC regulatory requirements for the EMA. Section 3.0 provides a description of the reactor vessels within the scope of this report, with illustrations of reactor vessel welds that are evaluated for equivalent margins in Figure 3-1 and Figure 3-2. Section 4.0 provides the material properties that are required for the EMA, and Section 5.0 presents the results of the EMA. Section 6.0 provides the summary and conclusions, Section 7.0 lists all references and Appendix A provides the technical basis for the use of B&WOG J-R Model 48 for the EMA reported herein (includes the development of new Model 68).

1.1 Equivalent Margins Analysis-Analysis of Record BAW-2192PA, Revision 00 [1] provided the EMA analysis of record for Levels A and B service loads for Surry Units 1 and 2. For 60 years, Surry Units 1 and 2 reported a plant-specific evaluation. The summary *reports for EMA analyses of record are as follows.

Surry Units 1 and 2 The Surry Units 1 and 2 current licensing basis equivalent margins analysis at 48 EFPY is summarized in Section 3.2.3 of NRC document "SURRY POWER STATION, UNIT NOS. 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING REACTOR VESSEL HEATUP AND COOLDOWN CURVES FOR 48 EFFECTIVE FULL-POWER YEARS,"

Adams Accession number ML11110A111 [9]. The NRC SER of the 48 EFPY P-T limits references the Dominion submittal entitled, VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 UPDATE TO NRC REACTOR VESSEL INTEGRITY DATABASE AND EXEMPTION REQUEST FOR ALTERNATE MATERIAL PROPERTIES BASIS PER 10 CFR 50.60(b) [10]. Specifically, Attachment 4 to Reference [10] includes AREVA document BAW-2494, Revision 1, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Units 1 and 2 for Extended Life through 48 Effective Full Power Years.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 2-1 2.0 REGULATORY REQUIREMENTS 2.1 Regulatory Requirements In accordance with 10 CFR 50 Appendix G [4], IV, A, 1., Reactor Vessel Upper Shelf Energy Requirements are as follows:

a. Reactor vessel beltline materials must have Charpy upper-shelf energy in the transverse direction for base material and along the weld for weld material according to the ASME Code, of no less than 75 ft-lb (102 J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68 J), unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. This analysis must use the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a (b)(2) at the time the analysis is submitted.

b. Additional evidence of the fracture toughness of the beltline materials after exposure to neutron irradiation may be obtained from results of supplemental fracture toughness tests for use in the analysis specified in section IV.A.1.a.

c. The analysis for satisfying the requirements _of section IV.A.1 of this appendix must be submitted, as specified in § 50.4, for review and approval on an individual case basis at least three years prior to the date when the predicted Charpy upper-shelf energy will no longer satisfy the requirements of section IV.A.1 of this appendix, or on a schedule approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate.

When the reactor vessels within the scope of this report were fabricated, charpy V-notch testing of the reactor vessel welds were in accordance with the original construction code, which did not specifically require charpy V-notch tests on the upper shelf.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 2-2 Applicable construction codes are as follows:

Surry--ASME B&PV Code, Section Ill, 1968 Edition through Winter 1968 Addenda (UFSAR Table 4.1-9)

In accordance with NRC Regulatory Guide 1.161 [15], the NRC has determined that the analytical methods described in ASME Section XI, Appendix K, provide acceptable guidance for evaluating reactor pressure vessels when the Charpy upper-shelf energy falls below the 50 ft-lb limit of Appendix G to 10 CFR Part 50. However, the NRC staff noted that Appendix K does not provide information on the selection of transients and gives very little detail on the selection of material properties. Selection of the limiting design transient (i.e., cooldown at 100 F/h) is consistent with BAW-2192PA [1], Section 5.3. Section 4.1 of this report includes a summary of the B&WOG J-integral resistance model, and Section 4.2 provides mechanical properties of weld metals.

The Linde 80 and Rotterdam weld locations that are included within the scope of this report (i.e., weld locations with SO-year projected fluence > 1.0E+17 n/cm2)are all assumed to have upper shelf energy values below 50 ft-bs and thus require an equivalent margins analysis.

2.2 Compliance with 10 CFR 50 Appendix G and Acceptance Criteria The analyses reported herein are performed in accordance with the 2007 Edition with 2008 Addenda [16] of Section XI of the ASME Code, Appendix K. The current edition of ASME Section XI listed in 10 CFR 50.55a is the 2013 Edition [17]. With regard to Appendix K, there are no differences between.the 2007 Edition with 2008 Addenda and the 2013 Edition of ASME Section XI, and hence these ASME Section XI, Appendix K analyses are equally applicable to the 2013 Edition of the ASME Code.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years Topical Report Page 2-3 The material properties used in this analysis are based on ASME Section II, Part D, 2007 Edition with 2008 Addenda. The only change in the material properties listed in the 2013 Edition of ASME Section II for the applicable properties is the coefficient of thermal expansion for stainless steel at 600°F; this value was changed from 9.8E-6 in/in/°F to 9.9E-6 in/in/°F. This does not impact the Levels A and B evaluation reported herein.

2.2.1 Acceptance Criteria Levels A and B ASME Section XI [17], Subarticle K-2200, provides the following acceptance criteria for Levels A and B Service Conditions:

a. When evaluating adequacy of the upper shelf toughness for the weld material for Levels A and B Service Loadings, an interior semi-elliptical surface flaw with a depth one-quarter of the wall thickness and a length six times the depth shall be postulated, with the flaw's major axis oriented along the weld of concern, and the fla*w plane oriented in the radial direction. When* evaluating adequacy of the upper shelf toughness for the base material, both interior axial and circumferential flaws with depths one quarter of the wall thickness and lengths six times the depth shall be postulated, and toughness properties for the corresponding orientation shall be used. Smaller flaw sizes may be used when justified. Two criteria shall be satisfied:

1. The applied J-integral evaluated at a pressure 1.15 times the accumulation pressure as defined in the plant specific Overpressure Protection Report, with a structural factor of 1 on thermal loading for the plant specific heatup and cooldown conditions, shall be less than the J-integral of the material at a ductile flaw extension of 0.1 in. (2.5 mm).

2. Flaw extensions at pressures up to 1.25 times the accumulation pressure of K-2200(a)(1) shall be ductile and stable, using a structural factor of 1 on thermal loading for the plant specific heatup and cooldown conditions.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at SO-Years Topical Report Page 2-4

b. The J-integral resistance versus flaw extension curve shall be a conservative representation for the vessel material under evaluation.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 3-1

3.0 DESCRIPTION

OF SURRY REACTOR VESSELS The Surry reactor vessels and applicable weld locations are shown in Figures 3-1 and 3-2. All weld locations evaluated for equivalent margins in this report are identified by an asterisk (*) in each Figure. Plant-specific weld copper and nickel content and 80-year fluence projections data needed for the equivalent margins analysis are provided in Table 3-1. The fluence projections are reported for all reactor vessel weld locations that are expected to exceed 1.0E+17 n/cm 2 at 80 years. The 80-year fluence

projections are conservative estimates based on detailed transport calculations completed by Westinghouse Electric Corporation using a methodology that is in compliance with Regulatory Guide 1.190 (WCAP -18028-NP, Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2).

Note that all fluence values are inside wetted surface with the exception of selected locations marked by an* (clad/base metal interface). Copper and nickel content of the reactor vessel shell welds is consistent with EMA analyses of record reported in Section 1.1; the copper and nickel content for transition welds and RV nozzle-to-nozzle belt forging welds reported in Table 3-1 were obtained from either the EMA analysis of record or a search of Surry reactor vessel fabrication reports.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 3-2 Table 3-1 Reactor Vessel Weld Locations--Copper Content and BO-Year Fluence Projections Surry Unit 1, BO-Year Fluence (E > 1.0 MeV)

SA-1493(Wire Ht. (IS) 1.50E+18 0.19 0.57 Nozzle Shell (NS) to Outlet Nozzle 8T1762 Forging Welds SA-1494(Wire Ht. (IS) 1.50E+18 0.16 0.57 8T1554B SA-1526 (Wire (IS) 1.50E+18 0.34 0.68 Ht. 299L44 NS to Inlet Nozzle Forging Welds SA-1580 (Wire (IS) 1.50E+18 0.19 0.57 Ht. 8T1762 NS to Intermediate Shell (IS) Gire. J726 (*) 7.98E+18 0.33 0.10 Weld ire Ht. 25017 SA-1494(Wire Ht. (*)1.33E+19 IS Long. Welds (Both) 8T1554) 0.16 0.57 IS to Lower Shell (LS) Gire. Weld (*)6.67E+19 SA-1585 (ID 40% 0.22 0.54 (Wire Ht. 72445) 0 SA-1650 NA IS to LS Gire. Weld (OD 60%) 0.22 0.54 ire Ht. 72445 SA-1494 (*)1.34E+19 LS Long. Weld (1) (Wire Ht. 0.16 0.57 8T1554)

LS Long. Weld (2) SA-1526 (Wire (*)1.34E+19 Ht. 299L44) 0.34 0.68 Sur Unit 2, BO-Year Fluence (E > 1.0 MeV)

Nozzle Shell (NS) to Outlet Nozzle (IS) 1.50E+18 Rotterdam 0.35 1.0 For in Welds WF-4 (Wire Ht. (IS) 1.50E+18 0.19 0.57 8T1762 NS to Inlet Nozzle Forging Welds WF-8 (Wire Ht. (IS) 1.50E+18 0.19 0.57 8T1762 NS to Intermediate Shell (IS) Gire. L737 (Wire Ht. (*) 9.21E+18 4275) 0.35 0.10 Weld IS Long. Weld (1), and (2) (100% SA-1585 (Wire (*) 1.36E+ 19 0.22 0.54 and OD 50% Ht. 72445 WF-4 (Wire Ht. (*) 1.36E+19 IS Long. Weld (2) (ID 50%) 0.19 0.57 8T1762 R3008 (Wire Ht. (*) 7.67E+19 IS to Lower Shell (LS) Gire. Weld 0227) 0.187 0.545 WF-4 (Wire Ht. (*) 1.37E+19 LS Long. Weld (Both) 0.19 0.57 8T1762

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at BO-Years Topical Report Page 3-3 Figure 3-1 Reactor Vessel-Surry Unit 1 Nozzle to Shell Weld* .

J726* (Rotterdam) Weld

.J.:::>--- Weld SA-1494*

Intermediate Shell (Plate) 144"

- - - - + - - - + - - + - + - - - - Weld SA-1494*

i---t--t----Weld SA-1526*

48.3" Lower Shell (Plate)

Equivalent Margins Analysis performed for these Linde 80 and Rotterdam Welds.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 3-4 Figure 3-2 Reactor Vessel-Surry Unit 2 Nozzle to Shell Weld*

i-~_ _JL__ _ .-----J~---- L737* (Rotterdam) Weld

- - + - - - - Weld SA-1585*

,__ _.....__ __._----1-.,__ _ _ Weld SA-1585 Outside 50%

WF - 4* Inside 50%

144" - - - - Intermediate Shell (Plate)

H-.-L._-.,..-ll---,L---14i-- R3008* (Rotterdam) Weld i--1--+----- Weld WF - 4*

i-----i----i--+---i--< Weld WF - 4* Inside 63%

WF - 8 Outside 37%

48.3' Lower Shell (Plate)

Equivalent Margins Analysis performed for these Linde 80 and Rotterdam Welds.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-1 4.0 MATERIAL PROPERTIES 4.1 J-lntegral Resistance Model The J-integral resistance model for Mn-Mo-Ni/Linde 80 welds in the reactor vessels of the B&WOG RVWG plants were developed using a large J-resistance data base. A detailed description of this model is provided in Appendix B of BAW-2192PA [1],

Revision 00. This model was developed using specimens irradiated to 8.45E+18 n/cm 2 ,

and the range of applicability of the model was extended (qualitatively) to approximately 1.90 E+19 n/cm 2 in Appendix B, Figure 3-1, to BAW-2251A [5]. See Appendix A of this report for a discussion of the extension of the range of applicability of the B&WOG J-R model to fluence values expected at 80 years for Surry Units 1 and 2. Consistent with BAW-2192PA, Revision 00, this J-R model is used for Linde 80 welds and Rotterdam welds.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 4-2 The coefficients a, d, and C4 are provided in Table 4-1. As required by ASME Section XI, ASME K-3300, when evaluating the vessel for Levels A, B, and C Service Loadings, the J-integral resistance versus crack-extension curve (J-R curve) shall be a conservative representation of the toughness of the controlling beltline material at upper shelf temperatures in the operating range. As such, the Jd correlation minus 2 standard errors is used for evaluation of Levels A & B service loadings (i.e., equation (1) multiplied by [ ] ).

As discussed in Appendix B to BAW-2192PA, the J-R curve was generated from a J-integral database obtained from the same class of material with the same orientation as the applicable reactor vessel materials using correlations for the effects of temperature, chemical composition, and fluence level. Crack extension was by ductile tearing with no cleavage. This complies with the ASME Code,Section XI, K-3300.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years Topical Report Page 4-3 Table 4-1 Parameters in Jd Model 48 4.2 Mechanical Properties of Weld Metals The following subsections provide representative properties for the Surry reactor vessels. The temperature dependent mechanical properties are developed from the 2007 Edition with 2008 Addenda of the ASME Code (Section II, Part D) for the reactor base metal and cladding (the ASME Code does not provide separate mechanical properties for base and weld metal). Both ASME Code minimum and representative irradiated yield strengths are also provided. The mechanical properties such as weld metal yield strengths typically used were the irradiated properties but in some cases the ASME Code minimum properties were conservatively considered. The irradiated material properties used herein are consistent with those used for the plants 60-year license renewal low upper shelf toughness analysis submittals (See Section 1.1 above).

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 4-4 4.2.1 Mechanical Properties for the Surry Reactor Vessels The Surry reactor vessels are fabricated using A-533 Grade B Class 1 (Mn-1/2Mo-1/2Ni) Low Alloy Steel (LAS) and stainless steel (18Cr-8Ni) cladding materials. Table 4-3 provides the Young's modulus (E), the mean coefficient of thermal expansion (a), and the yield strength (Sy) for the RV shell.

For the Surry reactor vessel shell welds, transition welds, and RV nozzle welds, the normal operating steady state condition cold leg temperature value is [

] (Table 5-1). The yield strength value for SA-1526 at [ ]

corresponding to a value of [ ] was used in the analysis.

Table 4-2 Mechanical Properties of Surry RV Shell Materials Weld RV Base Metal Metal Temp. E a Sy SA-1526 (OF) (ksi) (in/in/°F) (ksi) (ksi) 70 29000 7.00E-06 50.0 [ ]

200 28500 7.30E-06 47.0 [ ]

300 28000 7.40E-06 45.5 [ ]

400 27600 7.60E-06 44.2 [ ]

500 27000 7.70E-06 43.2 [ ]

600 26300 7.80E-06 42.1 [ ]

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-1 5.0 FRACTURE MECHANICS ANALYSIS 5.1 Methodology In accordance with ASME Section XI, Appendix K [16], Subarticle K-1200, the following analytical procedure was used for Levels A & B Service Loads.

a. The postulated flaws in the reactor vessel shell welds, the transition welds as well as RV nozzle-to-shell welds were postulated in accordance with the acceptance criteria of Subarticle K-2200.

b. Loading conditions at the locations of the postulated flaws were determined for Levels A and B Service Loadings. For Levels A and B Service loadings the equations to calculate the stress intensity factor (SIF) due to pressure and thermal gradients for a given pressure and cooldown rate are given in Article K-4210. Consistent with Section 5 of BAW-2192PA [1], the accumulation pressure is taken as ten percent above the design pressure and the maximum cooldown rate is 100°F/hr. In the area of the nozzle-to-shell weld, applied loadings consist of pressure, thermal, and attached piping reactions.

c. Material properties, including E, a, cry, and the J-integral resistance curve (J-R curve), were determined at the locations of the postulated flaws. Young's modulus, mean coefficient of thermal expansion and yield strength are addressed in Section 4.2. The J-R curve is discussed in Section 4.1.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-2

d. The postulated flaws were evaluated in accordance with the acceptance criteria of Article K-2000. Requirements for evaluating the applied J-integral are provided in Subarticle K-3200, and for determining flaw stability in Subarticle K-3400.

Subarticle K-3500(a) invokes the procedure provided in Subarticle K-4200 (K-4220) for evaluating the applied J-integral for a specified amount of ductile flaw extension. Three permissible evaluation methods to address flaw stability are described in Subarticle K-3500(b). The evaluation method selected herein is the J-R curve crack driving force diagram procedure described in Subarticle K-4310.

5.2 Procedure for Evaluating Levels A and B Service Loadings For RV shell regions remote from structural discontinuities, the applied J-integral is calculated in accordance with Appendix K, Subsubarticle K-4210, using an effective flaw depth to account for small scale yielding at the crack tip, and evaluated per K-4220 for upper-shelf toughness and per K-4310 for flaw stability, as outlined below.

(1) For an axial flaw of depth a, the stress intensity factor due to internal pressure is calculated with a structural factor (SF) on pressure using the following:

K 1P = (SF)p( l + ~; }mi) 05 F..

where F; = 0.982+ 1.00{; )'. 0.20,; (;),; 0.50 (2) For a circumferential flaw of depth a, the stress intensity factor due to internal pressure is calculated with a structural factor (SF) on pressure using the following:

where

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at BO-Years Topical Report Page 5-3 0.20 ~ (;) ~ 0.50 (3) For an axial or circumferential flaw of depth B, the stress intensity factor due to radial thermal gradients are calculated using the following:

0 ~ (CR) ~ 100°F/hr where for SA-508, Class 2 or SA-533, Grade B, Class1 steels the material coefficient Cm is defined as:

Ea Cm=( ) =0.0051, 1-v d CR= cooldown rate (°F/hr), and (4) The effective flaw depth for small scale yielding, Be, is calculated using the following:

2 a -a+

e

(-1 6,r J[K cry+K 1

P 1

t]

(5) For an axial flaw of depth Be, the stress intensity factor due to internal pressure is:

where I,;' = 0.982+1.00{ :* J,

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-4 (6)

For a circumferential flaw of depth Be, the stress intensity factor due to internal pressure is:

where 0.20,;; ( :*),;; 0.50 (7) For an axial or circumferential flaw of depth Be, the stress intensity factor due to radial thermal gradients is:

~2.5' K1t =Cm CR1t F;,

I (

0 ~ (CR) ~ 100°F/hr where (8) The J-integral due to applied loads for small scale yielding is calculated using the following:

(K ' +K ' )2 JI = 1000 Ip ' It E

where

, E E =--

1-v2 (9) Evaluation of upper-shelf toughness at a flaw extension of 0.10 in. is performed for a flaw depth, a= 0.25t + O.lOin.,

using

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-5 SF= 1.15 p = Pa where Pa is the accumulation pressure for Levels A and B. Service Loadings, such that where J1 = the applied J-integral for a safety factor of 1.15 on pressure, and a safety factor of 1.0 on thermal loading Jo.1 = the lower bound J-integral resistance at a ductile flaw extension of 0.10 in.

(10) Evaluation of flaw stability is performed through use of a crack driving force diagram procedure, by comparing the slopes of the applied J-integral curve and the lower bound J-R curve. The applied J-integral is calculated for a series of flaw depths corresponding to increasing amounts of ductile flaw extension. The applied pressure is the accumulation pressure for Levels A and B Service Loadings, Pa, and the safety factor (SF) on pressure is 1.25. Flaw stability at a given applied load is verified when the slope of the applied J-integral curve is less than the slope of the J-R curve at the point on the J-R curve where the two curves intersect.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-6 For both the Surry reactor vessels, the applied J-integrals at the nozzle-to-shell welds and the upper transition welds were determined using stresses from a detailed three-dimensional finite element analysis. Path line stresses were used to determine applied J-integrals that included a plastic zone correction to account for small scale yielding.

Based on the results of analysis performed for another B&W fabricated vessel it was deemed that the effects of structural discontinuities at the lower transition welds need not be explicitly addressed.

5.3 Evaluation for Flaw Extension The applied J-integrals for the RV shell welds, the RV transition welds, and the RV nozzle welds are calculated as discussed in Section 5.2.

5.3.1 Reactor Vessel Shell Welds The basic reactor vessel shell geometry and design pressure along with operating condition temperature information for each of the Surry reactor vessels (also referred to as three groups of reactor vessels) is provided in Table 5-1. Initial flaw depths equal to

% of the vessel wall thickness are analyzed for Levels A and B service loadings following the procedure outlined in Section 5.2 and evaluated for acceptance based on values for the J-integral resistance of the material from the Linde 80 J-R model discussed in Section 4.1. For each reactor vessel, calculations are initially carried out to identify the controlling weld such that subsequent detailed low upper shelf toughness flaw evaluations can be performed using the controlling weld.

The results of the plant specific flaw evaluations for each of the RV shell welds are presented in Table 5-2. From the results of the evaluation in Table 5-2, the controlling RV shell welds can be observed. The controlling welds are determined by noting the minimum ratio of the material J-resistance (Jo. 1) to the applied J-integral (J 1) (also referred to as "margin").

~

I Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-7 For the Surry reactor vessels, the controlling RV shell weld is the SA-1526 longitudinal weld of Surry Unit 1. This weld is located in the lower shell. The minimum margin (J 0.1/J 1) is [ ] , which is higher than the minimum acceptable value of 1.0.

5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds The reactor vessel nozzle welds are located in the substantially thicker cylindrical shell section (reinforced to account for the inlet/outlet RV nozzle openings and typically referred to as the nozzle belt), located above the reactor vessel shell welds. The reactor vessel nozzle belt dimensions are reported in Table 5-3.

For the Surry reactor vessels upper RV transition and RV nozzle weld regions, the J-applied results with the safety factor of 1.15 on applied pressure are compared against the lower bound J-integral resistance at a ductile flaw extension of 0.1 inches (Jo.1) in Table 5-6. The most limiting ratio or margin (Jo.1/J1) is [ ] due to postulating an axial flaw in the long itud in al weld SA-1585 located near the base of the taper transition section. Weld SA-1585 has the highest copper content of any longitudinal weld in the intermediate shells at Surry. The fluence value corresponding to the base of the taper transition section was utilized. The lower transition welds are located approximately four feet below the bottom of the core and are predicted to receive less than the threshold fluence value of 1.0E17 n/cm 2 at 80-years As such, these lower transition welds were not evaluated in this report.

5.4 Evaluation for Flaw Stability The flaw stability analysis is performed by calculating the applied J-integrals for various amounts of flaw extension with a safety factor (on pressure) of 1.25. The resulting applied J-integral curve can then be compared against the lower bound J-R curve for the weld metal. It is noted that applied J-integrals are also calculated with a safety factor on pressure of 1.15 for illustration of the Jo.1/J1 margin with respect to the lower bound J-R curve at a flaw extension of 0.1 inch.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-8 5.4.1 Reactor Vessel Shell Welds The applied J-integral values for the controlling weld of the Surry reactor vessels (SA-1526) is similarly calculated and shown in Table 5-5 with the corresponding mean and lower bound J-R curve values shown in Table 5-6. The resulting J-applied curves are then compared against the lower boun.d J-R curve for this material in Figure 5-1. An evaluation line at a flaw extension of 0.1 inch is included to confirm the results of Table 5-2 by showing the margin between the applied J-integral with the safety factor of 1.15 and the lower bound J-integral resistance of the material. The requirement for ductile and stable crack growth is demonstrated by Figure 5-1 since the slope of the applied J-integral curve for a safety factor of 1.25 is less than slope of the lower bound J-R curve at the point where the two curves intersect.

5.4.2 Reactor Vessel Transition Welds and RV Nozzle Welds For the Surry reactor vessels, the controlling weld and flaw orientation is the axial flaw in the longitudinal weld SA-1585 of Surry Unit 2 as discussed previously in Section 5.3.2. The applied J-integral for this axial flaw with a safety factor of 1.25 on pressure at various flaw extensions is plotted with the lower bound J-resistance curve (mean J-R curve provided for information only) in Figure 5-2. The slope of the applied J-integral is less than the slope of the lower bound J-resistance curve at the point of intersection, which demonstrates that the flaw is stable as required by ASMESection XI, Appendix K.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-9 Table 5-1 Reactor Vessel Shell Dimensions and Operating Conditions

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-10 Table 5-2 Flaw Evaluation Summary for RV Shell Regions

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-11

. Table 5-3 Reactor Vessel Nozzle Belt Dimensions

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-12 Table 5-4 Flaw Evaluation Summary of Surry Upper Transition and RV Nozzle-to-Shell Welds

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-13 Table 5-5 Applied J-lntegral versus Flaw Extensions of Surry Controlling RV Shell Weld (SA-1526)

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 5-14 Table 5-6

Mean & Lower Bound J-R Curve Values for Surry Controlling RV Shell Weld (SA-1526)

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years ,

Topical Report Page 5-15 Figure 5-1 J-lntegral versus Flaw Extension for Surry 1 and 2 Controlling Reactor Vessel Shell Weld (SA-1526)

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-16 Figure 5 J-lntegral versus Flaw Extension for Surry Axial Weld (SA-1585) Near Transition

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at BO-Years Topical Report Page 6-1 6.0

SUMMARY

AND CONCLUSIONS 6.1 Reactor Vessel Shell Welds The ASME Section XI, acceptance criteria for Levels A & B Service Loads for all reactor vessel shell welds are satisfied. The results of the limiting welds for Surry Units 1 and 2 are reported below.

The limiting RV shell weld is Surry Unit 1 axial weld SA-1526. With factors of safety of 1.15 on pressure and 1.0 on thermal loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (J0.1 )-

(Figure 5-1). The ratio J0.1/J1 = [ ] is greater than the required value of 1.0.

With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect (Figure 5-1).

6.2 Reactor Vessel Transition Welds and RV Nozzle Welds The acceptance criteria for Levels A & B Service Loads for RV transition welds and RV nozzle welds are satisfied. The results of the limiting weld considering transition welds and RV nozzle welds (inlet and outlet) for Surry are reported below.

The limiting weld for the Surry Units 1 and 2 is the longitudinal weld SA-1585 near the base of the transition section. With factors of safety of 1.15 on pressure and 1.0

on thermal loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (J0.1)-(Figure 5-2). The ratio J0.1/J1 = [ ] is greater than the required value of 1.0. This is due to postulating an axial flaw in the longitudinal weld of Surry Unit 2 while using the highest copper content of any intersecting longitudinal weld in the intermediate shell.

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at 80-Years Topical Report Page 6-2

With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect (Figure 5-2).

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 7-1 7 .0 REFERENCES

1. BAW-2192PA, Revision 00, "Low Upper Shelf Toughness Fracture Analysis of Reactor Vessels of B&W Owners Group Reactor Vessel Working Group for Levels A and B Conditions," April 1994, ADAMS Accession (Legacy) 9406240261 (P), 9312220294 (NP).

2. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components.

3. NUREG-2192, Volume 2, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, ADAMS Accession Number ML17188A158.

4. Code of Federal Regulations, Title 10, Part 50 - Domestic Licensing of Production and Utilization Facilities, Appendix G - Fracture Toughness Requirements, Federal Register Vol. 60. No. 243, December 19, 1995.

5. AREVA Document BAW-2251A, "Demonstration of the Management of Aging Effects for the Reactor Vessel, The B&W Owners Group Generic License Renewal Program," August 1999, ADAMS Accession Number 9909300150.

6. Not used

7. Not used

8. Not used.

9. SURRY POWER STATION, UNIT NOS. 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING REACTOR VESSEL HEATUP AND COOLDOWN CURVES FOR 48 EFFECTIVE FULL-POWER YEARS,"

Adams Accession number ML 1111 OA 111.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page 7-2 10.VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 UPDATE TO NRC REACTOR VESSEL INTEGRITY DATABASE AND EXEMPTION REQUEST FOR ALTERNATE MATERIAL PROPERTIES BASIS PER 10 CFR 50.60(b),

Adams accession number ML061650080.

11. Not used.

12. Not used.

13. Not used

14. Not used

15. NRC Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels With Charpy Upper Shelf Energy Less Than 50 ft-lb.

16. 2007 Edition (with 2008 Addenda) ASME & Boiler Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear power Plant Components, Appendix K.

17.2013 Edition ASME & Boiler Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, Appendix K.

18. BAW-1975, Applicability of the HSST Program Second and Third Irradiation Series Data to the Integrity of Nuclear Reactor Vessels, A.L.

Lowe, June 1987.

19. NUREG-1766, Safety Evaluation Report Related to the License Renewal of North Anna Power Station, Units 1 and 2 and Surry Power Station, Units 1 and 2, ADAM Accession Number ML030160825.

20. NUREG/CR-5729, Multivariable Modeling of Pressure Vessel and Piping J-R Data, May 1991, E. D. Eason, J. E. Wright, E. E. Nelson.

21. BAW-1543, Revision 4, Supplement 6-A, Supplement to the Master Integrated Reactor Vessel Surveillance Program, June 2007.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 7-3 22.ANP-3680P, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years.

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 8-1.

8.0 CERTIFICATION This report is an accurate description of the low upper-shelf toughness fracture analysis of Surry Units 1 and 2 reactor vessels.

dav_ ~ d-/,/-;ia/.5 Mark Rine el Nuclear Analysis Unit This report has been reviewed and is an accurate 'description of the low upper-shelf toughness fracture analysis of reactor vessels of Surry Un!ts 1 and 2.

~~ '1/llf /18.

Ashok Nana. ,

Component Analysis, Fracture and Materials Unit Verification of independent review.

~~::.,,___r:::~~;.__,,,,_~lP/j~j, David Cofflin

'Yi 8 Component Analysis, Materials Unit This report is approved for release.

Maya Chandra ekar NSSS Project Management

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-1 APPENDIX A APPENDIX B B&WOG J-R MODEL-DATA ANALYSIS AND EMPIRICAL MODEL DEVELOPMENT A.1 Background

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-2

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page A-3 Table A-1 Model 48, Range of Test Data

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years To ical Re art Pa e A-4

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-5 Figure A-1 BAW-2251A, Appendix B, Figure 3-1

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at BO-Years Topical Report Page A-6 A.2 New B&WOG J-LJa Data and Comparison to B&WOG J-R Model Table A-2 New B&WOG Specimen Data for Model Assessment

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years To ical Re ort Pa e A-7

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-8 Figure A-2 Jd (0.1) vs Fluence B&WOG J-R Model 48 and New Test Data (Normalized to Standard Conditions)

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page A-9 A.3 New B&WOG J-R Model

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years To ical Re art Pa e A-10

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page A-11 Table A-3 Jd Model Coefficients (Models 48, 58, and 68)

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at BO-Years Topical Report Page A-12 Figure A-3 Original and New Data and Model Fit Normalized at Standardized Conditions vs Aa

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-13 Figure A-4 Original and New Data and Model Fit Normalized at Standard Conditions vs Fluence

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page A-14 Figure A-5 Model 68 Residuals vs Fitted Values

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years Topical Report Page A-15 Figure A-6 Model 68 Standardized Residuals vs Fitted Values

Framatome Inc. ANP-3679NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-16 Figure A-7 Normal Q-Q Plot of Standardized Residuals

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-17 A.4 Model 68 Reconciliation to EMA Results Presented in Section 6.0

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-18 Figure A-8 Comparison of Models 48, SB, and 68 at Standard Conditions

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at BO-Years To ical Re ort Pa e A-19 Table A-4 EMA Reconciliation for Limiting RV Shell Welds-Models 48 and 68

Framatome Inc. ANP-3679NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years Topical Report Page A-20 A.5 Summary and Conclusions

v t e Letter Sequence Request
EPID:L-2018-RNW-0023, Dominion'S Presentation on Surry'S SLRA Safety Meeting-Final (for April 17, 2018) (Open) EPID:L-2018-RNW-0024, Dominion'S Presentation on Surry'S SLRA Safety Meeting-Final (for April 17, 2018) (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Meeting, Meeting, Meeting Questions 30 May 2019 2 August 2019 14 August 2019 5 September 2019 Answers 27 June 2019 17 July 2019 3 September 2019 19 September 2019 Results Approval... Other: ML18297A045, ML18319A184, ML18319A252, ML18345A244, ML18351A051, ML19015A243, ML19046A433, ML19098A810, ML19100A254, ML19107A021, ML19128A079, ML19140A449, ML19169A329, ML19198A059, ML19248C228, ML19267A042, ML19270E741, NRC-2018-0247, Federal Register Notice - Surry Power Station, Unit Nos. 1 and 2 - Correction and Extension of Date to Request a Hearing and Petition for Leave to Intervene Regarding the Virginia Electric and Power Company'S Application for Subsequent
MONTHYEAR2019-01-14 [Table View]

ML18291A836

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1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

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