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Enclosure 4, Attachment 2, ANP-3680NP, Revision 0, Topical Report
	ML18291A837
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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 Attachment 2 ANP-3680NP, REVISION 0 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

r-framatome Low Upper-Shelf Toughness Fracture ANP-3680NP Revision 0 Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report June 2018 Framatome Inc.

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Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-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 C and D ............................................ 2-3

3.0 DESCRIPTION

OF SURRY REACTOR VESSELS ........................................... 3-1 4.0 MATERIAL PROPERTIES AND LEVELS C&D SERVICE LOADINGS ........................................................................................................ 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 4.3 Levels C and D Service Loadings ........ :.................................................. 4-5 4.3.1 Surry ............................................................................................ 4-5 5.0 FRACTURE MECHANICS ANALYSIS .............................................................. 5-1 5.1 Methodology ........................................................................................... 5-1 5.2 Procedure for Evaluating Levels C and D Service Loadings ................... 5-2 5.2.1 Processing of Transient Time-History Data .................................. 5-3 5.2.2 Temperature Range for Upper Shelf Fracture Toughness Evaluations ................................................................ 5-3 5.2.3 Cladding Effects ........................................................................... 5-4 5.3 Evaluation for Levels C and D Service Loadings .................................... 5-7 5.3.1 Reactor Vessel Shell Welds ......................................................... 5-8 5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds ......................................................................................... 5-10 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 ........................ 6-2

7.0 REFERENCES

....... :.......................................................................................... 7-1 8.0 CERTIFICATION ............................................................................................... 8-1

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-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 3-2 Reactor Vessel Shell Dimensions .............................................................. 3-3 Table 3-3 Reactor Vessel Nozzle Belt Dimensions ................................................... 3-3 Table 4-1 Parameters in Jd Model 48 ....................................................................... 4-3 Table 4-2 Mechanical Properties of Surry RV Materials ............................................ 4-4 Table 5-1 Surry J-lntegral versus Flaw Extension for Levels C & D Service Loadings .................................................................................................. 5-11 Table 5-2 Surry J-R Curves for Evaluation of Levels C & D Service Loadings ........ 5-12 Table 5-3 Surry-Levels C & D Results for Nozzle to Shell and Upper Transition Welds ...................................................................................................... 5-13

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page iv List of Figures Figure 3-1 Reactor Vessel-Surry Unit 1 .................................................................... 3-4 Figure 3-2 Reactor Vessel-Surry Unit 2 .................................................................... 3-5 Figure 4-1 Surry Steam Line Break Transients ........................................................... 4-6 Figure 5-1 Surry-Kl versus Crack Tip Temperature for Levels C & D Service Loadings .................................................................................................. 5-14 Figure 5-2 Surry-J-lntegral versus Flaw Extension for Levels C & D Service Loadings .................................................................................................. 5-15 Figure 5-3 Surry- Level C & D Applied J Integral vs Crack Tip Temperature for the Outlet Nozzle to Shell Weld ............................................................... 5-16 Figure 5-4 Surry-Levels C & D Applied J Integral vs Crack Extension for the Outlet Nozzle to Shell Weld ..................................................................... 5-17

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-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 ONF Outlet Nozzle Forging PWROG Pressurized Water Reactor Owners Group RV Reactor Vessel RVWG Reactor Vessel Working Group SLR Subsequent License Renewal Sy Yield Strength TSs Technical Specifications USE Upper Shelf Energy

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for ,

Levels C & D Service Loads at BO-Years Topical Report Page vi ABSTRACT This topical report presents the results of an equivalent margins analysis (EMA) considering Levels C and D service loads for high copper Linde 80 weld metals and applicable non-Linde 80 welds using conservative BO-year fluence estimates for Surry Units 1 and 2 . 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-2178P, 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-3680P were generated by extracting Surry Unit 1 and 2-specific results from Sections 1.0 through 7.0 of BAW-2178P, Supplement 1, Revision 0. The B&WOG J-integral resistance model is discussed in Appendix A to ANP-3679P, which is identical to Appendix A of BAW-2192P, Supplement 1, Revision 0, with the exception that references to plants other than Surry Units 1 and 2 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, with selection of design transients consistent with the guidance in Regulatory Guide 1.161, Section 4.0. 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 C & D Service Loads for all reactor vessel shell welds are satisfied. The acceptance criteria for Levels C & D Service Loads for RV transition welds and RV nozzle welds are also satisfied.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page vii Consistent with 8AW-2178PA, Revision 0, the EMA reported herein utilizes the 8&WOG J-integral resistance (J-R) Model 48 reported in 8AW-2192PA, Appendix 8.

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 .

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 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 80-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 8AW-2192P, Supplement 1, Appendix A and ANP-3679P, Rev. 0, Appendix A.

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 8AW-2192P, Supplement 1, Appendix A and ANP-3679P, Rev. 0, Appendix A.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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 C and D 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 8&W-designed and Westinghouse-designed reactor vessels fabricated by 8&W/Rotterdam: Surry Units 1 and 2. Note that the Surry EMA reported herein is techni_cally identical to the Surry EMA reported in 8AW-2178P, Supplement 1, Revision O [7], which was submitted to the NRG by the PWROG on December 15, 2017. That is, Sections 1.0 through 7.0 of ANP-3680P were generated by extracting Surry Unit 1 and 2-specific results from Sections 1.0 through 7.0 of 8AW-2178P, Supplement 1, Revision 0. The 8&WOG J-integral resistance model is discussed in Appendix A to ANP-3679P [3], which is identical to Appendix A of 8AW-2192P, Supplement 1, Revision 0, with the exception that refere,nces to plants other than Surry Units 1 and 2 were removed from Sections A.1, A.2, and A.4.

The EMA reported herein utilizes the 8&WOG J-integral resistance (J-R) Model 48 reported in 8AW-2192PA, Appendix 8. Justification for use of Model 48 for 80-year fluence is addressed in ANP-3679P, Rev. 0, Appendix A [3].

Equivalent margins analyses are reported for all reactor vessel weld locations with 80-year fluence projections that exceed 1.0 E+17 n/cm2 (E> 1.0 MeV). Upper shelf energy evaluations at reactor vessel base metal locations with 80-year fluence projections greater than 1.0 E+17 n/cm2, if needed, will be addressed separately in the Surry Units 1 and 2 subsequent license renewal application.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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/cm2 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 [5] "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/cm2 [19] at 80 years. The use of the terms "beltline" and/or "extended beltline" are not used in this report.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 1-3 The 60-year EMA summary reports Surry Units 1 and 2, are reported in Section 1.1.

Section 2.0 provides the current NRG regulatory requirements for the EMA. Section 3.0 provides a description of all reactor vessels within the scope of this report, with illustrations of applicable reactor vessel welds in Figures 3-1 and 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, and Section 7.0 lists all references. ANP-3679P, Rev. 0, Appendix A [3] provides the technical justification for the use of B&WOG J-R Model 48 for the EMA reported herein.

1.1 Equivalent Margins Analysis-Analysis of Record BAW-2178PA, Revision 00 [1] provided the EMA analysis of record for Levels C and D service loads for Surry Units 1 and 2. For 60 years, Surry Units 1 and 2 reported plant-specific evaluations. 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 NRG 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 ML 1111 OA 111 [1 O]. The N RC 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) [11]. Specifically, Attachment 4 to Reference [11] 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-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 2-1 2.0 REGULATORY REQUIREMENTS 2.1 Regulatory Requirements In accordance with 10 CFR 50 Appendix G [5], 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.

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years

Topical Report Page 2-2 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. The applicable construction code is as follows.

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

In accordance with NRG Regulatory Guide 1.161 [16], the NRG 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 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 design transients and selection of material properties are addressed in Sections 3.0 and 4.0.

The Linde 80 and Rotterdam weld locations that are included within the scope of this report (i.e., weld locations with 80-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 [17] 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 [18]. 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-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 2-3 The materialproperties 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.BE-6 in/in/°F to 9.9E-6 in/in/°F. At the limiting time points in the Level C & D analysis, where cladding effects are included, the temperature of the cladding is well below 600°F, and thus this change does not impact the low upper shelf toughness analysis reported herein.

2.2.1 Acceptance Criteria Levels C and D ASME Section XI [17], Subarticles K-2300 and K-2400, provide acceptance criteria for Levels C and D Service Conditions. Consistent with BAW-2178PA [1], the evaluations reported herein will utilize acceptance criteria applicable to Level C Service Loadings as summarized below.

a. When evaluating adequacy of the upper shelf toughness for the weld material for Level C Service Loadings, interior semi-elliptical surface flaws with depths up to 1110th of the base metal wall thickness, plus the cladding thickness, with total depths not exceeding 1 in. (25 mm), and a surface length 6 times the depth, shall be postulated, with the flaw's major axis oriented along the weld of concern, and the flaw plane oriented in the radial direction. When evaluating adequacy of the upper shelf toughness for the base material, both interior axial and circumferential flaws shall be postulated, and toughness properties for the corresponding orientation shall be used. Flaws of various depths, ranging up to the maximum postulated depth, shall be analyzed to determine the most limiting flaw depth. Smaller maximum flaw sizes may be used when justified. Two criteria shall be satisfied:

1. The applied J-integral shall be less than the J-integral of the material at a ductile flaw extension of 0.10 in. (2.5 mm), using a structural factor of 1 on loading.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 2-4

2. Flaw extensions shall be ductile and stable, using a structural factor of 1 on loading.

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

The above Level C acceptance criteria will be conservatively imposed on the Level D transients defined in Section 4.3.1. In addition, for information purposes only, the acceptance criteria applicable to the Level D Service Loadings as summarized below will be reported for Level D transients.

a. When evaluating adequacy of the upper shelf toughness for Level D Service Loadings, flaws as specified for Level C Service Loadings shall be postulated, and toughness properties for the corresponding orientation shall be used. Flaws of various depths, ranging up to the maximum postulated depth, shall be analyzed to determine the most limiting flaw depth. Flaw extensions shall be ductile and stable, using a factor of.safety of 1.0 on loading.

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

c. The extent of stable flaw extension shall be less than or equal to 75% of the vessel wall thickness, and the remaining ligament shall not be subject to tensile instability.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years

Topical Report Page 3-1

3.0 DESCRIPTION

OF SURRY REACTOR VESSELS The Surry reactor vessels with applicable weld locations are shown in Figure 3-1 and Figure 3-2. All weld locations evaluated for equivalent margins in this report are identified by an asterisk(*) in each figure. Weld copper and nickel content and BO-year fluence projections data needed for the equivalent margins analysis are provided in Figure 3-1. The BO-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- WCAP-1 B02B-NP, Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2).

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 analyses of record or a search of Surry reactor vessel fabrication reports.

The dimensions of the reactor vessel shell geometry for the Surry reactor vessels are provided in Table 3-2. Similarly, the dimensions for the reactor vessel nozzle belt region located above the reactor vessel shell course for each of these three groups of reactor vessels are given in Table 3-3.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years Topical Report Page 3-2 Table 3-1 Reactor Vessel Weld Locations--Copper Content and 80-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

(*)1.33E+19 IS Long. Welds (Both). 0.16 0.57 IS to Lower Shell (LS) Gire. Weld (*)6.67E+19 0.22 0.54 1040%

NA IS to LS Gire. Weld (OD 60%) 0.22 0.54 Wire 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 Rotterdam 3 (IS) 1.50E+18 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. (*) 9.21 E+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-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 3-3 Table 3-2 Reactor Vessel Shell Dimensions Table 3-3 Reactor Vessel Nozzle Belt Dimensions

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 3-4 Figure 3-1 Reactor Vessel-Surry Unit 1

~ - - Nozzle to Shell Weld*

J726* (Rotterdam) Weld

_E>-- Weld SA-1494*

Intermediate Shell (Plate) 144" i-----+-------t--+-+---- Weld SA-1494*

__.-t-t---- Weld SA-1526*

Lower Shell (Plate)

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

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 3-5 Figure 3-2 Reactor Vessel-Surry Unit 2 Nozzle to Shell Weld*

L737* (Rotterdam) Weld

,_.,__,__ _ _ _ Weld SA-1585*

ICORE Weld SA-1585 Outside 50%

,,----+--+-+-----------< WF - 4* Inside 50%

144"

,___ _ _ Intermediate Shell (Plate) 1---t--r-'~--r---1---.r'----.-.- R3008* (Rotterdam) Weld

.__+--1----- Weld WF - 4*

+---+---+-+----< 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-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years

. Topical Report Page 4-1 4.0 MATERIAL PROPERTIES AND LEVELS C&D SERVICE LOADINGS 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 model (J-R model) data base. A detailed description of this model is provided in Appendix B of BAW-2192PA [2], Revision 0. 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 [6]. See Appendix A of BAW-2192P, Revision 0, Supplement 1, 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 Oconee reactor vessels, Surry reactor vessels, and Turkey Point reactor vessels. Consistent with BAW-2178PA, Revision 00, this J-R model is used for Linde 80 welds and Rotterdam welds.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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, 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. When evaluating the vessel for Level D Service Loadings, the J-R curve shall be a best estimate 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 Level C Service Loadings (i.e., equation (1) multiplied by [ ] ) while the unaltered Jd correlation would be used to evaluate Level D Service Loadings.

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 using correlations for effects of temperature, chemical composition, and fluence level.

Crack extension was by ductile tearing with no cleavage. This complies with ASME Section XI, K-3300.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-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 Ill) for the reactor base metal and cladding (the ASME Code does not provide separate mechanical properties for base and weld metal). 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. At the limiting time points in the Levels C & D analysis where cladding effects are included the temperature of the cladding is well below 600°F, and thus this change does not impact the EMA analyses summarized herein.

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 4-4 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 60-year license renewal low upper shelf toughness analysis submittals created for the plants (See Section 1.1 above).

4.2.1 Mechanical Properties for the Surry Reactor Vessels The Surry reactor vessels are fabricated using A-533 Grade 8 Class 1(Mn-1/2Mo-1/2Ni)

Low Alloy Steel (LAS) and stainless steel (18Cr-8Ni) cladding materials. Table 4-2 provides the Young's modulus (E), the mean coefficient of thermal expansion (a), and the yield strength (Sy) for the RV base metal and weld material and the E and a properties for the RV cladding material.

Table 4-2 Mechanical Properties of Surry RV Materials Weld RV Base Metal Metal Cladding Temp. E a Sy SA-1526 E a (OF) (ksi) (in/in/°F) (ksi) (ksi) (ksi) (in/in/°F) 70 29000 7.00E-06 50.0 [ ] 28300 8.50E-06 200 28500 7.30E-06 47.0 [ ] 27500 8.90E-06 300 28000 7.40E-06 45.5 [ ] 27000 9.20E-06 400 27600 7.60E-06 44.2 [ ] 26400 9.50E-06 500 27000 7.70E-06 43.2 [ ] 25900 9.70E-06 600 26300 7.80E-06 42.1 [ ] 25300 9.80E-06

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 4-5 4.3 Levels C and D Service Loadings 4.3.1 Surry Levels C and D Service Loadings were developed for the Surry reactor vessels for the two steam line break transients identified below.

Level D: Steam Line Break (SM-0979)

Steam Line Break (SSDC 1.3 SLB)

The pressure and temperature steam line break transients for the Surry reactor vessels are illustrated in Figure 4-1.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years Topical Report Page 4-6 Figure 4-1 Surry Steam Line Break Transients

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-1 5.0 FRACTURE MECHANICS ANALYSIS 5.1 Methodology In accordance with ASME Section XI, Appendix K [17], Subarticle K-1200, the following analytical procedure was used for Levels C & D Service Loadings:

a. 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 Subarticles K-2300 and K-2400.

b. Loading conditions at the locations of the postulated flaws were determined for Levels C and D Service Loadings.

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.

d. The postulated flaws were evaluated in accordance with the acceptance criteria of Article K-2000 by calculating the applied J-integral according to the procedure provided by Subarticle K-5210. The applied J-integral was then evaluated to satisfy the criteria for flaw extension in Subarticle K-5220 and flaw stability in Subarticle K-5300.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-2 5.2 Procedure for Evaluating Levels C and D Service Loadings The evaluation for the Levels C and D service loadings is performed as follows:

1 For each transient described in Section 4.3, calculate stress intensity factors for a 1

semi-elliptical flaw of depth up to / 10 of the base metal wall thickness, as a function of time, due to internal pressure and radial thermal gradients with a factor of safety of 1.0 on loading. The applied stress intensity factor, K1, calculated for each of these transients is compared to the KJc upper-shelf toughness curve of the weld material. The transient for which the applied K1 most closely approaches the KJc curve is chosen as the limiting transient, and the critical time in the limiting transient selected for further evaluation occurs at the point where K, most closely approaches the KJc curve.

2 At the critical transient time, develop a crack driving force diagram with the applied J-integral and J-R curves plotted as a function of flaw extension. The adequacy of the upper-shelf toughness is evaluated by comparing the applied J-integral with the J-R curve at a flaw extension of 0.10 in. Flaw stability is assessed by examining the slopes of the applied J-integral and J-R curves at the points of intersection.

3 Verify that the extent of stable flaw extension is no greater than 75% of the vessel wall thickness by determining when the applied J-integral curve intersects the mean J-R curve.

4 Verify that the remaining ligament is not subject to tensile instability. The internal pressure p shall be less than P,, where P, is the internal pressure at tensile instability of the remaining ligament. The pressure at instability, P1, is given in K-5300, Appendix K of ASME Section XI for both axial and circumferential flaws.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-3 5.2.1 Processing of Transient Time-History Data For the Surry reactor vessels, the applied J-integrals at the nozzle to shell welds and the upper transition weld were determined from three-dimensional finite element analysis. The through-thickness path line stresses were subsequently used to calculate stress intensity factors and the applied J-integrals were determined based on consideration of small scale yielding. For the controlling reactor vessel shell weld, stress intensity factors were calculated using the one-dimensional, finite element thermal and closed form stress models and linear elastic fracture mechanics methodology of the PCRIT computer code.

5.2.2 Temperature Range for Upper Shelf Fracture Toughness Evaluations Upper-shelf fracture toughness is determined through use of Charpy V-notch impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level.

Similarly, fracture toughness can be addressed in three different regions on the temperature scale, i.e., a lower-shelf toughness region, a transition region, and an upper-shelf toughness region. Fracture toughness of reactor vessel steel and associated weld metals are conservatively predicted by the ASME initiation toughness curve, K1e, in the lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, KJe, is derived from the upper-shelf J-integral resistance model described in Section 4.1. The upper-shelf toughness then becomes a function of fluence, copper content, temperature, and fracture specimen size. When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperature. Since the present analysis addresses the low upper-shelf fracture toughness issue, only the upper-shelf temperature range, which begins at the intersection of Kie and the upper-shelf toughness curves, KJe, is considered.

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-4 5.2.3 Cladding Effects

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-5

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-6

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-7 5.3 Evaluation for Levels C and D Service Loadings The type of analysis models and computer code used to evaluate the RV shell welds, the RV transition welds and the RV nozzle welds for Levels C & D service loads are addressed in Section 5.2.1.1. Section 4.3.1 addresses the specific types of transient events analyzed. The applied J-integral for the RV shell welds, the RV transition and nozzle welds, due to these Levels C & D transient events, is calculated and evaluated as discussed in Section 5.2.

The transition region toughness and upper shelf toughness are discussed in Section 5.2.2. Transition region toughness is obtained from the ASME Section XI equation for crack initiation, Kie= 33.2 + 20.734 exp[0.02(T - RTNoT)]

using the applicable RTNDT value for a flaw depth of 1110th the waHthickness, where:

= transition region toughness, ksiv'in T = crack tip temperature, °F Upper shelf toughness Kie is derived from the J-integral resistance model of Section 4.1 for a flaw depth of 1/1 oth the wall thickness, a crack extension of 0.10 inch, and the applicable fluence value at the crack tip:

K1c =

where upper-shelf region toughness, ksiv'in Jo.1 = J-integral resistance at Da = 0.1 in.

Using the above equations, the transition and upper shelf toughness values as a function of temperature are determined for the controlling weld and Levels C and D service loading conditions.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-8 5.3.1 Reactor Vessel Shell Welds For the Surry reactor vessels, the controlling weld was identified to be the SA-1526 RV shell weld based on the results of Levels A and B Service Loadings. This controlling weld is therefore evaluated at a flaw depth of 1/10 the base metal thickness for Levels C and D Service Loadings. The two main steam line break (SLB) transients (SSDC 1.3 SLB & SM-0979) identified in Section 4.3.2 and illustrated in Figure 4-1 are evaluated.

The analysis is performed using the PCRIT Code.

Figure 5-1 shows the variation of the applied stress intensity factor, K1, for these two transient cases as a function of the crack tip temperature. This figure also shows the transition region toughness Kie curve and the mean and lower bound upper-shelf toughness KJc curves with crack tip temperature. The Kie curve is determined using the Adjusted Reference Temperature (ART) value at 1/1 oth of the wall thickness for the limiting weld SA-1526, which at BO-years is [ ] . The symbols on the K1 curves for each of the two transient cases indicate points in time at which PCRIT solutions are available. The SSDC SLB is identified as the limiting transient since it most closely approaches the KJc limit of the weld. All subsequent analysis was therefore based on evaluation of this transient case. In the upper-shelf toughness range, the SSDC SLB K1 curve is closest to the lower bound KJc curve at 10.0 minutes into the transient. This time is selected as the critical time in the transient at which to perform the flaw evaluation for Levels C and D Service Loadings.

The additional stress intensity factor attributable to the cladding, K1c1ad, at 10.0 minutes (limiting time point) into the SSDC SLB transient is determined to be [ ] at .

a flaw depth corresponding to 1/1 oth of the wall thickness.

Applied J-lntegrals are calculated for the controlling weld for various flaw depths in Table 5-1 using stress intensity factors from PCRIT for the SOC SLB (at 10.0 min.) and adding [ ] to account for cladding effects. Stress intensity factors are converted to J-integrals by the previously reported plane strain relationship.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-9 Since the RV shell weld for the Surry reactor vessels is [ ] inch thick as given in Table 3-2, the initial flaw depth of 1/10 of the wall thickness is [ ] inches. The flaw extension is calculating by subtracting this depth from the built-in PCRIT flaw depths. The results along with the mean and lower bound J-R curves developed in Table 5-2, are plotted in Figure 5-2. An evaluation line is used at a flaw extension of 0.10 in. to show that the applied J-integral is less than the lower bound J-integral of the material, as required by ASME Section XI, Appendix K.

The applied J-integral at a flaw extension of 0.1 inch is determined to be [ ]

as reported at the base of Table 5-1. The associated margin using the lower bound and mean J-R curve values (from Table 5-2) for Levels C and D conditions, respectively are also shown at the base of Table 5-1. The margin for Level C Service Loading is

[ ] and the margin for Level D Service Loading is [ ] . In accordance with ANP-3679P, Appendix A [3], the margin for Level C loading using J-R Model 68 is

[ ] .

The requirements for ductile and stable crack growth are demonstrated by Figure 5-2 since the slope of the applied J-integral curve is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

Referring to Figure 5-2, the Level D Service Loading requirement that the extent of stable flaw extension be no greater than 75% of the vessel wall thickness is easily satisfied since the applied J-integral curve intersects the mean J-R curve at a flaw extension that is only a small fraction of the wall thickness.

The last requirement for Level D Conditions is that the internal pressure p shall be less than P1, the internal pressure at tensile instability of the remaining ligament. The calculations for P1 were determined for an axial flaw with various flaw depths up to 1.26 inches. The P1 values calculated are in excess of 9000 psi, which is far greater than any expected pressure. The remaining ligament is therefore not subject to tensile instability.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years Topical Report Page 5-10 5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds The reactor vessel upper and lower transition welds are located above and below the reactor core, respectively (see Figure 3-1 and Figure 3-2). The RV nozzle welds are located above the upper transition weld in the substantially thicker cylindrical section (reinforced to account for the inlet/outlet RV nozzle openings). The reactor vessel nozzle belt dimensions are reported in Table 3-3.

For the Surry reactor vessels the applied J-integrals for the nozzle to shell and upper transition welds were evaluated for Levels C and D Service Loadings. Both transients shown in Figure 4-1 are evaluated. The bounding results with safety factor of 1.0 on the applied pressure are compared with the lower bound J-integral resistance at a ductile flaw extension of 0.1 inches in Table 5-3. The outlet nozzle is seen to be limiting and has a margin of [ ] . The applied J-integral vs crack tip temperature for each transient (SSDC SLB and SM-0979 SLB) is plotted in Figure 5-3 for the outlet nozzle, along with the temperature dependent mean and lower bound J0 . 1 curves. As can be seen all points of the transient remain below the lower bound J 0 .1 . Additionally, Figure 5-3 shows the Kie fracture toughness using an RTNDT of [ ] (lowest of nozzle to shell welds maximizes the upper shelf range), converted to an equivalent J using K1/l(E/(1-v 2 )); the intersection of this curve with the J 0.1 curves establishes the upper shelf temperature range. The applied J-integral at the limiting time point at various flaw extensions is plotted with the lower bound J-resistance curve in Figure 5-4; 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 ASME Section XI, Appendix K.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-11 Table 5-1 Surry J-lntegral versus Flaw Extension for Levels C & D Service Loadings

- _J

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-12 Table 5-2 Surry J-R Curves for Evaluation of Levels C & D Service Loadings

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-13 Table 5-3 Surry-Levels C & D Results for Nozzle to Shell and Upper Transition Welds

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-14 Figure 5-1 Surry-Kl versus Crack Tip Temperature for Levels C & D Service Loadings

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-15 Figure 5-2 Surry-J-lntegral versus Flaw Extension for Levels C & D Service Loadings

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-16 Figure 5-3 Surry- Level C & D Applied J Integral vs Crack Tip Temperature for the Outlet Nozzle to Shell Weld

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 5-17 Figure 5-4 Surry-Levels C & D Applied J Integral vs Crack Extension for the Outlet Nozzle to Shell Weld

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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 C & D Service Loads for all reactor vessel shell welds are satisfied. ASME Section XI, Appendix K, Level C acceptance criteria (Subarticle K-2300), relative to the ratio of applied J-integral to J-integral of the material and use of the lower bound J-integral resistance curve, were conservatively imposed on the Level D transients evaluated in Section 5.0, although ASME Section XI, Subarticle K-2400(b) permits use of a best estimate J-integral resistance curve for Level D Service Loadings. The results of the limiting welds for Surry Units 1 and 2 are reported below.

The limiting weld among the Surry reactor vessel shell welds is Surry Unit 1 longitudinal weld SA-1526. The limiting transient for Level C & D service Loads is the SSDC 1.3 steam line break.

With a factor of safety of 1.0 on loading, the applied J-integral (J 1) for the limiting reactor vessel shell weld (Surry Unit 1, SA-1526) is less than the lower bound J-integral of the material at a ductile flaw extension of 0.10 inch (Jo. 1) with a ratio Jo.1/J1 of [ ] , which is greater than the required value of 1.0. Using the mean J-R curve permitted by Subarticle K-2400 for this Service Level D transient, the ratio Jo.1/J1 is [ ] . In accordance with BAW-2192, Supplement 1, Appendix A [3],

the margin for Level C loading using J-R Model 68 is [ ] .

With a factor of safety of 1.0 on loading, flaw extensions are ductile and stable for the limiting reactor vessel shell weld (SA-1526) since the slope of the applied J-integral curve is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

For weld SA-1526 it was demonstrated that flaw growth is stable at much less than 75% of the vessel wall thickness. It has also been shown that the remaining ligament is sufficient to preclude tensile instability.

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 6-2 6.2 Reactor Vessel Transition Welds and RV Nozzle Welds The ASME Section XI, acceptance criteria for Levels C & D Service Loads for all reactor vessel transition welds and reactor vessel nozzle welds are satisfied. ASME Section XI, Appendix K, Level C acceptance criteria (Subarticle K-2300), relative to the ratio of applied J-integral to J-integral of the material and use of the lower bound J-integral resistance curve, were conservatively imposed on the Level D transients evaluated in Section 5.0, although ASME Section XI, Subarticle K-2400(b) permits use of a best estimate J-integral resistance curve for Level D Service Loadings. The results of the limiting welds for Surry Units 1 and 2 are reported below.

The upper transition weld and RV inlet and outlet nozzle-to-shell welds were evaluated for Levels C and D Service Loadings. The limiting transient for Level C & D service loads is the SSDC 1.3 steam line break.

With a factor of safety of 1.0 on loading, the applied J-integral (J1) for the RV nozzle-to-shell welds and upper transition weld are less than the lower bound J-integral of the material at a ductile flaw extension of 0.10 inch (J 0.1) with the following ratios for Jo.1/J1: [ ] for the RV outlet nozzle-to-shell weld, [ ] for the RV inlet nozzle-to-shell weld, and [ ] for the upper transition weld (i.e. axial flaw in longitudinal weld SA-1585 at the intersection with circumferential weld L737).

With a factor of safety of 1.0 on loading, flaw extensions are ductile and stable for the limiting RV outlet nozzle-to-shell weld (i.e., limiting location considering RV nozzle-to-shell welds and upper transition weld).

For the RV outlet nozzle-to-shell weld it was demonstrated that flaw growth is stable at much less than 75% of the vessel wall thickness. Tensile instability was not explicitly calculated but because this section of the reactor vessel is thicker compared to the RV shell welds, it is considered to be bounded by the RV shell location.

Framatome Inc. ANP-3680NP Revision O Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 7-1 7 .0 REFERENCES

1. BAW-2178PA, Revision 00, "Low Upper Shelf Toughness Fracture Analysis of Reactor Vessels of B&W Owners Group Reactor Vessel Working Group for Level C and D Conditions," April 1994, ADAMS Accession (Legacy) 9406290288 (P).

2. 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 8 Conditions," April 1994, ADAMS Accession (Legacy) 9406240261 (P), 9312220294 (NP).

3. Framatome Document ANP-3679P, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels A & 8 Service Loads at 80-Years.

4. Not used

5. 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.

6. 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.

7. BAW-2178, Supplement 1, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C & D Service Loads.

8. Not used

9. Not used

Framatome Inc. ANP-3680NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at BO-Years Topical Report Page 7-2 10.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.

11. 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.

12. Not used

13. Not used

14. Not used

15. Not used

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

17. 2007 Edition (with 2008 Addenda) ASME & Boiler Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear power Plant Components, Appendix K.

18.2013 Edition ASME & Boiler Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear power Plant Components, Appendix K.

19. RIS-2014-11, NRC REGULATORY ISSUE

SUMMARY

2014-11 INFORMATION ON LICENSING APPLICATIONS FOR FRACTURE TOUGHNESS REQUIREMENTS FOR FERRITIC REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS.

Framatome Inc. ANP-3680NP Revision 0 LQWJ)pp(:lrc.Sh~lf ToughnessFrc1cture Mechanics Analysis of Surry Units 1 and 2 Reactor Vessels for Levels C & D 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 vessels.

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 Units* 1 and 2. *

~ ~'--. fta/t<r/tv.

Ashok Nana

Component Analysis, Fracture and Materials Unit *

Verification of independent review.

~~-=-P----~~~ri/L4f1,s David Cofflin Component Analysis racture and Materials Unit This report is approved for release.

Maya handrashekh NSSS Project Mana

Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 Attachment 3 PWROG-17033-NP, REVISION 1 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Update for Subsequent License Renewal: WCAP-13045, "Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam 1

Supply Systems"

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWROG-17033-NP Revision 1 Update for Subsequent License Renewal: WCAP-13045, "Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems" PA-MSC-1498, Revision 0 June 2018 Authors: Geoffrey Loy*

Structural Design and Analysis Ill Reviewer: Anees Udyawar

Structural Design and Analysis Ill Approved: Lynn Patterson*, Manager Structural Design and Analysis Ill James Molkenthin*, Program Director PWR Owners Group PMO

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii ACKNOWLEDGEMENTS This report was developed and funded by the PWR Owners Group under the leadership of the participating utility representatives of the Materials Committee.

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PWROG-17033-NP June 2018 Revision 1

I --

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V PWR Owners Group United States Member Participation* for PA-MSC-1498, Revision O [4]

Participant Utility Member Plant Site(s) Yes No Ameren Missouri Callaway (W) X American Electric Power D.C. Cook 1 & 2 (W) X Arizona Public Service Palo Verde Units 1, 2, & 3 (CE) X Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X North Anna 1 & 2 (W) X Dominion VA Surry 1 & 2 (W) X Catawba 1 & 2 (W) X Duke Energy Carolinas McGuire 1 & 2 (W) X Oconee 1, 2, & 3 (B&W) X Robinson 2 (W) X Duke Energy Progress Shearon Harris (W) X Entergy Palisades Palisades (CE) X Entergy Nuclear Northeast Indian Point 2 & 3 (W) X Arkansas 1 (B&W) X Entergy Operations South Arkansas 2 (CE) X Waterford 3 (CE) X Braidwood 1 & 2 (W) X Byron 1 & 2 (W) X Exelon Generation Co. LLC TMI 1 (B&W) X Calvert Cliffs 1 & 2 (CE) X Ginna (W) X Beaver Valley 1 & 2 (W) X FirstEnergy Nuclear Operating Co.

Davis-Besse (B&W) X St. Lucie 1 & 2 (CE) X Turkey Point 3 & 4 (W) X Florida Power & Light \ NextEra Seabrook (W) X Pt. Beach 1 & 2 (W) X Luminant Power Comanche Peak 1 & 2 (W) X PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi PWR Owners Group United States Member Participation* for PA-MSC-1498, Revision O [4]

Participant Utility Member Plant Site(s) Yes No Omaha Public Power District Fort Calhoun (CE) X Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X PSEG - Nuclear Salem 1 & 2 (W) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Farley 1 & 2 (W) X Southern Nuclear Operating Co.

Vogtle 1 & 2 (W) X Sequoyah 1 & 2 (W) X Tennessee Valley Authority Watts Bar 1 & 2 (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Xcel Energy Prairie Island 1 & 2 (W) X

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PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii PWR Owners Group International Member Participation* for PA-MSC-1498, Revision 0 Participant Utility Member Plant Site(s) Yes No Asco 1 & 2 (W) X Asociaci6n Nuclear Asc6-Vandell6s Vandellos 2 (W) X AxpoAG Beznau 1 & 2 (W) X Centrales Nucleares Almaraz-Trillo Almaraz 1 & 2 (W) X EDF Energy Sizewell B (W) X Dael 1, 2 & 4 (W) X Electrabel Tihange 1 & 3 (W) X Electricite de France 58 Units X Elektriciteits Produktiemaatschappij Zuid- Borssele 1 (Siemens) X Nederland Eletronuclear-Eletrobras Angra 1 (W) X Emirates Nuclear Energy Corporation Barakah 1 & 2 X Eskom Koeberg 1 & 2 (W) X Hokkaido Tomari 1, 2 & 3 (MHI) X Japan Atomic Power Company Tsuruga 2 (MHI) X Mihama 3 (W) X Kansai Electric Co., LTD Ohi 1, 2, 3 & 4 (W & MHI) X Takahama 1, 2, 3 & 4 (W & MHI) X Kori 1, 2, 3 & 4 (W) X Hanbit 1 & 2 (W) X Korea Hydro & Nuclear Power Corp.

Hanbit 3, 4, 5 & 6 (CE) X Hanul 3, 4 , 5 & 6 (CE) X Genkai 2, 3 & 4 (MHI) X Kyushu Sendai 1 & 2 (MHI) X Nuklearna Electrarna KRSKO Krsko (W) X RinghalsAB Ringhals 2, 3 & 4 (W) X Shikoku lkata 1, 2 & 3 (MHI) X Taiwan Power Co. Maanshan 1 & 2 (W) X

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii TABLE OF CONTENTS 1 PURPOSE ...................................................................................................................... 1-1 2 BACKGROUND ............................................................................................................. 2-1 2.1 OPERATING EXPERIENCE AND APPLICATION OF LATEST FRACTURE TOUGHNESS DATABASE ................................................................................. 2-2 3 STABILITY ANALYSIS AND FRACTURE TOUGHNESS ............................................... 3-1 3.1 FRACTURE TOUGHNESS DETERMINED IN WCAP-13045 ............................ 3-1 3.2 FRACTURE TOUGHNESS BASED ON NUREG/CR-4513, REVISION 2 ......... 3-3 4 FATIGUE CRACK GROWTHANALYSIS ...................................................................... .4-1 5 CONCLUSIONS .................................................................................*............................ 5-1 6 REFERENCES ............................................................................................................... 6-1 APPENDIXA : ASME CODE SECTION XI CODE CASE N-481 ............................................ A-1 PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1 PURPOSE The purpose of this Topical Report (TR) is to extend the fracture mechanics integrity evaluation in WCAP-13045 [1], "Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems," through Subsequent License Renewal (SLR), 80 years of operation. In the past, plants have used the generic fracture mechanics evaluation performed in WCAP-13045 to comply with the requirements of ASME Code Case N-481 for plant-specific license renewal applications to extend plant operations to 60 years. In this TR, the fracture mechanics evaluation provides the justification for performing visual inspections, in lieu of volumetric inspections, for reactor coolant pump (RCP) casings as incorporated in the ASME Code Section XI, and extend the applicability of WCAP-13045 to 80 years of operation for all plants with Westinghouse pump casings.

ASME Section XI Table IWB-2500-1, Examination Categories [2] required performing periodic volumetric inspections of the welds of the primary loop pump casings in nuclear power plants.

Since these inspections result in radiation exposure to the personnel performing the inspections and require significant resources to perform the inspections, the ASME Code Committee approved Code Case N-481 [3] in March 1990 (see Appendix A of this report) that allows an alternative to the volumetric inspection requirement. The NRC endorsed Code Case N-481 in Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability ASME Code Section XI Division 1," in April 1992.

Westinghouse design primary loop pump casings are heavy-wall cast stainless steel. A volumetric inspection of the full thickness of the welds using the ultrasonic test method from the outside diameter surface is impractical due to the severe attenuation associated with the large grain structures. A volumetric inspection of the full thickness of the welds would require unconventional approaches (inside diameter and outside diameter ultrasonic testing or radiographic testing) that require access to the internal side of the pump casing.

ASME Code Case N-481 [3] allowed replacing the volumetric examination of primary loop pump casings with a fracture mechanics-based integrity evaluation supplemented by specific visual inspections. WCAP-13045 [1] contains the integrity evaluation that was performed to demonstrate compliance with ASME Code Case N-481 for 40 years of operation.

Since WCAP-13045 [1] was issued in September 1991, the ASME Code tables have been updated over time to be consistent with the guidance in Code Case N-481 to require visual inspections of the primary loop pump casings. In March 2004, Code Case N-481 was annulled by ASME, and the information in Code Case N-481 was implemented into the 2008 Addenda of

ASME Code Section XI. Note that the ASME Section XI 2000 Addenda replaced the pump casing weld B-L-1 volumetric examinations with visual examinations, while the ASME Section XI 2008 Addenda eliminated the pump casing weld (B-L-1) examinations completely. The only required examination is a visual examination of the pump casing (B-L-2) when the pump is disassembled for maintenance or repair.

The technical basis for WCAP-13045 was based on experience with evaluations performed for an assumed 40 year life. Due to the SLR program to extend an operating license to 80 years,

. PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 the integrity evaluations in WCAP-13045 were reviewed and confirmed to be applicable for 80 years of service. The fracture mechanics integrity evaluation in this report, as well as the requirements in Code Case N-481 (now incorporated into the ASME Code Section XI) were reviewed to confirm that the visual inspections for pump casings continue to remain valid for an 80 year life.

WCAP-13045 [1] contains the fracture mechanics based integrity evaluation for cast austenitic stainless steel pump casings as required by Code Case N-481 [3]. The evaluations co.ntained in WCAP-13045 are applicable to Westinghouse design RCPs. There are eight different models of pumps: Models 63, 70, 93, 93A, 93A-1, 930, 100A, and 1000. Models 63, 70, 93 and 930, which have a tangent outlet nozzle. Models 93A and 93A-1 have outlet nozzles that are radially orientated. Models 1OOA and 1000 are similar to the general design of the Model 93, except for a radially oriented outlet nozzle that is consistent with the Model 93A. Models 93, 93A and 93A-1 are the most commonly used RCPs in Westinghouse Nuclear Steam Supply System (NSSS) plants.

WCAP-13045 contains a model that is representative of each of the outlet nozzle configurations that was used for a 30 finite element stress analysis and fracture evaluation (the inlet nozzles are reasonably axisymmetric with the pump casing proper). The representative models chosen were the Model 93A (radial outlet nozzle) and Model 93 (tangential outlet nozzle). The material of the pump casings is fabricated from SA-351 CF8, except for the pumps of three plants which were fabricated from SA-351 CF8M. The SA-351 CF8 and CF8M are known to be susceptible to thermal aging.

WCAP-13045 addressed thermal aging of the cast austenitic stainless steel pump casings (CF8 and CF8M) by using end of life (40 years) fracture toughness values for all Westinghouse design pump casings. The fracture toughness criteria were established using the lowest toughness for each pump component. This report justifies the continued use of the end of life fracture toughness values determined in WCAP-13045 for 80 years of service.

The fracture toughness is used in WCAP-13045 as part of the elastic plastic fracture mechanics (EPFM) analysis based on the J-integral approach; therefore, it is necessary to confirm the fracture toughness values for an 80 year evaluation, and demonstrate that the EPFM analysis continues to remain valid for 80 years. The J-integral evaluation also used bounding loads that covered a wide range of pump casing nozzle loads from the various different plants. This report only confirms the toughness properties for 80 years, and not the applied loads, since the applied loadings in the J-integral analysis considered in WCAP-13045 will not be impacted by license extension to 80 years of operation because they are not time dependent. Note that the fracture toughness properties based on the CF8M (high molybdenum content) material is more susceptible (limiting) to thermal aging than the CF8 material. Therefore, the CF8M material is used in the fracture mechanics evaluation in this TR; and the conclusions for the CF8M fracture toughness determined in this report also apply to the CF8 material.

Fatigue crack growth evaluations were also determined in WCAP-13045 for the high stress outlet nozzle crotch regions. Various crack sizes were considered in the evaluations, and based on the evaluations, it was demonstrated that the fatigue crack growth was small. The PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3 discussion in this report also justifies that the fatigue crack growth evaluations contained in WCAP-13045 are still applicable for the 80 year design life.

The following two items were reviewed to confirm the applicability of WCAP-13045 [1] for 80 years of service:

1. Confirm that the fracture toughness (J 1c - J at crack initiation and Jmax - J at maximum crack extension) used in WCAP-13045 [1] and the associated tearing modulus (Tmat) for the stability analysis are applicable for 80 years of service. The postulated 1/4T flaw size used in the stability analysis is compared to the final flaw size due to fatigue crack growth.

2. Confirm that the generic fatigue crack growth (FCG) analysis performed in WCAP-13045 is applicable for 80 years of service, specifically; the stresses, stress intensity factor (SIF) equations, transient definitions and cycles, and the FCG rate.

The stability calculations in WCAP-13045 were reviewed based on the applicability of the fracture toughness for SLR in Section 3 of this report. The fatigue crack growth analysis in WCAP-13045 was reviewed based on for FCG rates, stresses, and transient definitions and cycles in Section 4. The final conclusions of this report are provided in Section 5, with all cited references provided in Section 6.

Appendix A of this report also provides the ASME Code Section XI Code Case N-481, for reference, which was approved and published in March 1990, and later annulled in March 2004, as the requirements of the code case were incorporated in the 2000 Addenda and 2008 Addenda of the ASME Code Section XI IWB-2500.

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 the basis on which the information is considered proprietary. These code letters are listed with their meanings in BMS-LGL-84 [12].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 BACKGROUND Plants have used the generic fracture mechanics evaluation performed in WCAP-13045 to comply with the requirements of ASME Code Case N-481 for the license renewal applications to extend plant operations to 60 years. Before Code Case N-481 was annulled and the requirements of the Code Case were incorporated into ASME Code Section XI, utilities were furthermore required to perform a plant-specific evaluation using the generic fracture mechanics analysis in WCAP-13045 to demonstrate compliance with N-481. Following the incorporation of Code Case N-481 into Section XI, no further fracture mechanics evaluations have been performed as plants were not required to perform volumetric or routine internal visual examinations of RCP casing welds, and the only ASME Code Section XI inspection requirement is to perform external surface examinations of the pump casing welds and visual examinations of the internal surfaces of the pump casing welds when the RCP is disassembled for other reasons (e.g. maintenance or refurbishments).

The thermal aging and fatigue crack growth methodology in WCAP-13045 was reviewed and is provided in this TR to demonstrate that the time-limiting aging analysis aspects of WCAP-13045 continue to remain bounding and acceptable for 80 years of plant life. The NRC has approved several plant license renewal applications for 60 years of operation that utilized the generic fracture mechanics analyses and conclusions in WCAP-13045 as discussed below:

1. The NRC Safety Evaluation Report (Section 4.4.4) for Salem Units 1 and 2 [13]

discusses the use of the generic fracture mechanics analysis in WCAP-13045 to meet the requirements of ASME Code Case N-481 for a 60 year license renewal. The NRC staff concluded that the generic analysis in WCAP-13045 is applicable to the Salem design of the RCP casings. The generic analysis, WCAP-13045, bounds the plant-specific analysis, WCAP-16957-P [17], as approved by the staff in the SER [13]. The analysis was shown to remain valid for extended operation.

2. In the 60 year license renewal NRC Safety Evaluation Report for D.C. Cook Units 1 and 2, NUREG-1831, Section 4.7.2 [14], the NRC staff discusses the use of WCAP-13045 to satisfy the requirements of ASME Code Case N-481 based on a plant-specific analysis, WCAP-13128 [18]. In Section 4.7.2.4 of [14], the NRC concludes that the time-limited aging analysis (TLAA) regarding ASME Code Case N-481 that was provided is acceptable.

3. The 60 year license renewal Safety Evaluation Report (Section 4.3.2.10) for Diablo Canyon Units 1 and 2 [15] discusses that WCAP-13045 was used to demonstrate compliance with Code Case N-481 based on a plant-specific evaluation, WCAP-13895

[19]. In Section 4.3.2.10.4 of [15], the NRC concludes that it was shown that the aging of the RCP pump casings will be adequately managed for extended operation.

4. Based on the Sequoyah 60 license renewal Safety Evaluation Report [16], the NRC concluded that the plant's LRA did not need to include a TLAA related to Code Case N-481 because the licensing basis is per the ASME Code Section XI Edition, which no longer relies on N-481 for ISi (in-service inspection interval) requirements. Thus, it was PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 not necessary to perform any TLAA analysis for RCP pump casings and N-481, as it does not meet Criterion 4 or 6 in 10 CFR 54.3(a).

In summary, WCAP-13045 has been reviewed by NRC to support 60 year license renewal applications for several plants that used Code Case N-481 as a basis for their ISi examination programs. This report reviews the TLAA aspects of WCAP-13045 to demonstrate its continued applicability for 80 years of plant operation.

2.1 OPERATING EXPERIENCE AND APPLICATION OF LATEST FRACTURE TOUGHNESS DATABASE The Westinghouse RCP casings have an operating history that demonstrates the inherent flaw tolerance and structural stability of the pump casings. No detectable service-induced flaws nor discernable degradation of the cast austenitic stainless steel (CASS) pump casings and welds in the Westinghouse pump operating history have been identified.

The fracture mechanics evaluation contained in this TR (later discussed in Section 3) considers the latest fracture toughness correlations that have been developed for the CASS pump casings. The end of life fracture toughness properties for the pump casing materials are determined based on NUREG/CR-4513, Revision 2, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR systems," by Omesh Chopra, published in May 2016 [11].

Revision 2 of NUREG/CR-4513 provides a large database for CASS material and thermal aging, and builds on the work performed in 1994 for Revision 1 of NUREG/CR-4513. In 1994, the Argonne National Laboratory (ANL) completed an extensive research program to evaluate the extent of thermal aging of cast stainless steel materials [11 ]. The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a database, both from data within ANL and from international sources, of approximately 85 compositions of cast stainless steel exposed to a temperature range of 290-400°C (550-750°F) for up to 58,000 hours0 days 0 hours 0 weeks 0 months (6.5 years).

In 2015, the work done by ANL was augmented, and the fracture toughness database for CASS materials was aged to 100,000 hours0 days 0 hours 0 weeks 0 months at 290-350°C (554-633°F). The methodology for estimating fracture properties has been extended to cover CASS materials with a ferrite content of up to 40%. From this database in NUREG/CR-4513, Revision 2, ANL developed lower bound correlations for estimating the extent of thermal aging of cast stainless steel [11 ].

ANL developed the fracture toughness estimation procedures by correlating data in the database conservatively. After developing the correlations, ANL validated the estimation procedures by comparing the estimated fracture toughness with the measured value for several cast stainless steel plant components that were removed from service. The procedure developed by ANL in [11] was used for the end of life fracture toughness values contained in this TR. The ANL research program was sponsored and the procedure was accepted by the NRC.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 STABILITY ANALYSIS AND FRACTURE TOUGHNESS 3.1 FRACTURE TOUGHNESS DETERMINED IN WCAP-13045 The fracture mechanics integrity evaluation is based on the Elastic Plastic Fracture Mechanics (EPFM) methodology as discussed in Sections 10 and 11 of WCAP-13045 [1]. The EPFM is determined for a postulated 1/4T (1/4 thickness) flaw size with a six-to-one (6: 1) aspect ratio.

This particular flaw size is consistent with the guidelines of Code Case N-481 Part (d)(5). The location of the postulated 1/4T flaws are either at highest stressed region, regions of significant stress concentrations, or locations in welds not affected by discontinuities such as nozzles.

Additional discussion on flaw postulation is contained in Section 9 of WCAP-13045.

The criterion for establishing stability is based on the fracture toughness of the pump casings, as well as the tearing modulus, T, as discussed in Section 10.4 of WCAP-13045 and shown below:

A crack is stable if either

1) Japplied < J1c or if

2) Japplied > J1c then T applied < Tmaterial and Japplied :s; Jmaximum The applied toughness (Japplied or Japp) and applied tearing modulus (Tapplied or Tapp) are calculated with the EPRI handbook methodology [1 O], based on various combinations of loading parameters and material properties for the various pump designs as discussed in WCAP-13045.

The Japp and Tapp are not impacted by a design life to 80 years as they are not time dependent; however, the fracture toughness material parameters such as the crack initiation toughness (J1c),

which is based on a J-integral resistance curve, the maximum fracture toughness (Jmaximum or Jmax), and the tearing modulus (Tmaterial or Tmat) need be reviewed to confirm that these parameters are not impacted by 80 years of operation.

The end of service (40 years) fracture toughness (J1c, T mat, and Jmax) of the pump casings are calculated in Section 5 of WCAP-13045 [1]. The lower bound toughness criteria selected from among all the pump casing (and welds) and models (Models 63, 70, 93, 93A, 93A-1, 930, 1OOA, and 100D) are given in Table 5-1 of WCAP-13045. The minimum fracture toughness values based on the most limiting SA-351 CFBM component from Table 5-1 of WCAP-13045 which were considered in this TR are shown in Table 1 below. The fracture toughness properties for CFBM in WCAP-13045 bound the CFS material, because the thermal aging is more limiting for the CFBM material than the CFS material.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 Table 1: End-of-Service Fracture Toughness from WCAP-13045 (Table 5-1) [1]

Material J 1c (in-lb/in 2

) T mat (dimensionless) Jmax (in-lb/in 2)

[ t,c,e [ ]a,c,e [ ]a,c,e Model 93 SA-351 CF8M Chemistry: [

]a,c,e Note: The values J 1c, T mat, and Jmax shown for the CF8M material above are the same values in Table 5-1 and on page A-30 of WCAP-13045.

The fracture toughness data in Table 1 (from Table 5-1 of WCAP-13045) is based on the chemistry data from Appendix A (page A-30) of WCAP-13045. Based on the chemistry of the limiting heat of cast austenitic stainless steel (i.e., Silicon (Si), Chromium (Cr), Molybdenum (Mo), Nickel (Ni), Carbon (C), Manganese (Mn), and Nitrogen (N) in percent weight, and percent delta ferrite), the fracture toughness is calculated in accordance with WCAP-10931 [5], Slama

[6], and WCAP-10456 [7].

Based on the results in Slama [6] and WCAP-10456 [7], the minimum (saturated) fracture toughness properties were obtained after [

rc,e Therefore, the fracture toughness properties shown in Table 1 (per Table 5-1 of WCAP-13045) are at the full-aged saturated condition material values, and are therefore applicable to 80 years of service life, because the resulting minimum (saturated) properties are reached by

[ rc,e Therefore, the EPFM stability analysis and conclusions in WCAP-13045 [1], based on the saturated fracture toughness values in Table 1, are applicable for 80 years of operation.

The minimum fracture toughness values based on Table 5-1 of WCAP-13045 (as shown in Table 1) were compared with the latest fracture toughness correlations for thermal aging of cast austenitic stainless steel CF8M material per NUREG/CR-4513, Revision 2 [11]; see Section 3.2 below. This comparison confirms that the fracture toughness values in WCAP-13045 [1] are bounding and applicable to the 80 year design life.

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--~I WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 3.2 FRACTURE TOUGHNESS BASED ON NUREG/CR-4513, REVISION 2 In this section of the TR, a calculation is performed to determine the J,e, Tmat and Jmax values based on latest industry guidelines and fracture toughness correlations for thermal aging of cast austenitic stainless steel from NUREG/CR-4513 Revision 2 [11] for the most limiting CF8M pump casing material, using the limiting heat specific chemistry values provided in Table 1 (originally from WCAP-13045). The fracture toughness values calculated in accordance with NUREG/CR-4513 will be compared to the values determined in Table 1, from WCAP-13045.

This comparison of fracture toughness values will demonstrate if the toughness properties from WCAP-13045 remain bounding and acceptable for 80 years. Note that the impact of thermal aging on the CFS fracture toughness properties is less than that for the CF8M material; therefore the evaluation herein only considered the CF8M material because it bounds the fracture toughness values for the CFS material.

The following equations are contained in NUREG/CR-4513, Revision 2 [11] and are applicable to the CF8M material. The calculated fracture toughness values based on [11] are shown in Table 2.

Creq =Cr+ 1.21 (Mo) + 0.48(Si) - 4.99 = (Chromium equivalent)

Nieq = (Ni) + 0.11 (Mn) - 0.0086(Mn)2 + 18.4(N) + 24.5(C) + 2. 77 = (Nickel equivalent)

De =100.3(Creq I Nieq )2-170.72(Creq / Nieq )+74.22 = (Ferrite Content)

The elements are in percent weight and De is ferrite in percent volume.

The saturation room temperature (RT) impact energies for the cast stainless steel materials are determined from the chemical composition.

For CF8M steel with < 10% Ni, the saturation value of RT impact energy Cvsat (J/cm2 ) is the lower value determined from:

log10CVsat = 0.27 + 2.81 exp (-0.022~)

where the material parameter ~ is expressed as:

~ = De (Ni + Si + Mn)2(C + 0.4N)/5.0 and using:

log 10 Cvsat = 7.28 - 0.011 De - 0.185Cr - 0.369Mo - 0.451 S- 0.007Ni - 4. 71 (C + 0.4N)

For CF8M steel with 2:'.: 10% Ni, the saturation value of RT impact energy Cvsat (J/cm 2) is the lower value determined from:

log10CVsat = 0.84 + 2.54 exp (-0.047~)

where the material parameter ~ is expressed as:

~ = De (Ni + Si + Mn)2(C + 0.4N)/5.0 and using:

log 10CVsat = 7.28- 0.011De- 0.185Cr- 0.369Mo -0.451Si - 0.007Ni-4.71(C + 0.4N)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 The saturation J-R curve at RT, for static-cast CF8M steel is given by:

Jd = 1.44 (CVsat) 1"35 (Llat for CVsat < 35 J/cm 2

for CVsat ~ 35 J/cm 2

Jd = 16 (CVsat)°- 67 (Llat n = 0.20 + 0.08 10910 (CVsat) where Jd is the "deformation J" in kJ/m 2 and Lla is the crack extension in mm.

The saturation J-R curve at 290-320°C (554-608°F), for static-cast CF8M steel is given by:

98 Jd = 5.5 (CVsat)°" (Llat for CVsat < 46 J/cm 2 41 Jd = 49 (CVsat)°" {Llat for CVsat ~ 46 J/cm2 n = 0.19 + 0.07 log10 {CVsat) where Jd is the "deformation J" in kJ/m 2 and Lla is the crack extension in mm.

t,c,e The tearing modulus, T material is calculated by T =

dJ/da

Ela/, where dJ/da is the slope of the J-R curve, Eis elastic modulus, and at is the flow strength (average of the yield strength and ultimate strength). Applying the NUREG/CR-4513, Revision 2 [11] correlations, the fracture toughness properties are given in Table 2.

NUREG/CR-4513, Revision 2 discusses that the fracture toughness correlations used for the full aged condition are applicable to plants that have been operating for greater than or equal to 15 Effective Full Power Years (EFPY) for the CF8M material. The Westinghouse NSSS plants have been operating for greater than 15 EFPY; therefore, the use of the fracture toughness correlations discussed above is applicable to the fully aged or saturated conditions.

Therefore, the fracture toughness values based on the original methodology in WCAP-13045 are limiting (see Table 1) as compared to the fracture toughness values based on NUREG/CR-4513 (see Table 2). By calculating the latest industry correlation for the fracture toughness values for aged cast austenitic stainless steel from NUREG/CR-4513, it can be concluded that the aged toughness values in Table 1 (from Table 5-1 of WCAP-13045) are bounding and limiting for 80 years of operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 2: NUREG/CR-4513 [11] Fully-Aged (Saturated) Fracture Toughness Material J1c (in-lb/in 2) Tmat (dimensionless) Jmax (in-lb/in 2) rc,e t,c,e [ ]a,c,e Model 93 SA-351 CF8M [ [

Chemist[Y (from Table 1): [

]a,c,e ASME Code properties are used for E (modulus of elasticity) = 25.6 x 106 psi and Yield Strength =

19,350 psi. Ultimate strength= 67,000 psi per Table 6-3 ofWCAP-13045, which is a close approximation to the ASME Code values. The values are for the cold leg temperature, approximately 550°F.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 FATIGUE CRACK GROWTH ANALYSIS Two fatigue crack growth analyses were originally performed in Section 12 of WCAP-13045 [1],

one for a postulated crack in the highest stressed outlet nozzle knuckle in a Model 93A pump casing and the other in the highest stressed outlet nozzle knuckle in a Model 93 pump casing.

The FCG analysis for the Model 93A pump was performed prior to the publication of the stainless steel FCG rate in the ASME Code Section XI; therefore, the FCG rate in [Equation 1]

from [9] was initially considered in WCAP-13045. However, the FCG analysis for a Model 93A RCP was calculated based on [Equation 2], which was an updated version of [Equation 1] for stainless steel in a water environment. [Equation 1] in WCAP-13045 was provided for information purposes and is as follows:

!~ = 5.4 x 10- 12 Cl<ett) 4 .48 (inches/cycle) [Equation 1]

Keff = KmaxC1-R) 0*5 Kmax and Kmin is in units of ksi v'm The FCG rate used for the Model 93 postulated flaw in [1] was based on Figure C-3210-1 of Article C-3000 of Section XI of the ASME Code (1989 Edition) and is shown in Equation 2 below:

~ =CFS E (L\K) 3*30 (inches/cycle) [Equation 2]

Where: C = 2.42 x 10-20 , F = 1 for temperatures below 800°F, S = the R ratio correction (see the definition of 'S' in Equation 3 below), R = the ratio of the minimum stress intensity factor (SIF) to the maximum SIF, L\K = the range of stress intensity factors (psi vin), and E = 2 based on stainless steel in water [8].

The current NRC approved 2013 Edition of the ASME Code Section XI Appendix C fatigue crack growth for stainless steel is shown below. A factor of 2 is applied to the da/dN rate to account for the environmental effects of a postulated flaw in water as discussed in [8]:

!~ = 2*C 0 S (L\K1) 3*3 (inches/cycle) [Equation 3]

where C0 = 10 11 [-10.009 + 8.12x10-4 T - 1.13x10-5 T2 + 1.02x10-9 T 3]

T = Temperature (°F) = 550°F (see Table 11-1 and Table 11-6 of [1])

S = 1.0 for R::;; 0, S = 1 + 1.8R when O < R < 0.79, and S = -43.35 + 57.97R when 0.79 < R < 1

.6.K1 is in units of ksi vin, R = KminlKmax There are no significant differences between the current stainless steel FCG rate in water,

[Equation 3] and the FCG rate used in WCAP-13045 [1], and [Equation 2] for the Model 93 PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 postulated flaws. The difference between the current stainless steel FCG rate in water [Equation 3] and the FCG rate considered in WCAP-13045, [Equation 1] for the Model 93A postulated flaw is also insignificant. Therefore, the existing FCG rates in Section 12 of WCAP-13045 are acceptable based on current industry standards for fatigue crack growth for stainless steel material in a water environment.

Other inputs required for an FCG analysis are the stress intensity factors, stresses, transient cycles and transient definitions. The stress intensity factor (SIF) correlations used for the FCG analysis in WCAP-13045 are consistent with the current correlations provided in 2013 Edition of ASME Code Section XI Appendix A. The transient stresses used in the FCG analysis are generic and encompass the various pump models. These stresses are not impacted for 80 years of operation, as plant operations for this extended period will not vary greatly from current operations. The number of predicted cycles for 80 years of service for each applicable transient is assumed to be bounded by the number of transient cycles considered in the 40 year life of the plant and shown in Table 12-2 of WCAP-13045. However, to ensure conservatism in the results presented in this TR, the FCG cycles for 40 years were doubled for 80 years to account for any large differences in the transient cycles. It was concluded that the final flaw size would still be less than the stability flaw size for this region as discussed below.

The calculated FCG for four flaw sizes in the outlet nozzle knuckle region of a Model 93A pump casing is given in Table 12-1 of WCAP-13045. The calculated FCG for three flaw sizes in the outlet nozzle knuckle region of a Model 93 pump casing is given in Table 12-3 of WCAP-13045.

One of the flaw size cases for FCG was for an initial flaw depth of 0.3"; this particular flaw depth was the maximum acceptable flaw size in the Acceptance Standards in Table IWB-3518-2 (for pressure retaining welds in pump casings) up to the 2007 Edition of the ASME Code Section XI.

The flaw depth of 0.3" is also the maximum acceptable flaw size in the Acceptance Standards Table IWB-3519.2-2 (for pump casings) in later editions of the ASME Code Section XI.

Therefore, the flaw depth of 0.3" was an important flaw size case to consider in WCAP-13045 for the flaw tolerance evaluation, and any actual as-found flaws larger than this depth would need to be evaluated based on fracture mechanics. The other FCG postulated flaw size cases considered in WCAP-13045 were provided as sensitivity studies to demonstrate that the flaws do not grow significantly over time.

Based on the fatigue crack growth analyses in Tables 12-1 and 12-3 of WCAP-13045, the flaw depth of 0.3" grows to a maximum value of [ ]a.c,e over 40 years of service. All other inputs for the FCG analysis (stress intensity factors, stress, transient cycles and transient definitions) are bounded by, or are similar to, the inputs of WCAP-13045, then the fatigue crack growths for 80 years are similar to or less than the crack growth in Table 12-1 and Table 12-3 of WCAP-13045 for 40 years. Additionally, if the number of transient cycles for 40 years are doubled for 80 years to account for any large differences in the transient cycles, then the final flaw size would continue to be less than the minimum stability flaw size of 1/4T flaw depth ([ ]a,c,e in Table 11-6 of WCAP-13045), which is associated with the location of the highest stressed region. Therefore, the FCG analysis provided in Section 12 of WCAP-13045 remains valid for 80 years of operation.

PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 CONCLUSIONS The objective of this report is to review the RCP casing fracture mechanics integrity evaluations and the fatigue crack growth analysis contained in WCAP-13045 [1], and to confirm that they remain valid for 80 years of service.

Section 3.1 discussed the most limiting pump casing CF8M fracture toughness and the tearing modulus used in the WCAP-13045 J-integral stability evaluations (EPFM). The WCAP-13045 fracture toughness determinations are applicable to 80 years of service because the fracture toughness parameters are at full-aged saturated conditions, therefore, any additional aging past 40 years does not have an impact on the fracture toughness parameters. Section 3.2 compared the minimum fracture toughness in WCAP-13045 with the latest industry toughness correlations in NUREG/CR-4513, Revision 2 [11]. Based on the conclusions in Section 3.2, it is demonstrated that the fracture toughness values in WCAP-13045 are less than (more conservative and limiting) than the fracture toughness values in NUREG/CR-4513. Therefore, as compared with the current industry standards, the fracture toughness values in WCAP-13045 are limiting and applicable to 80 years of operation. Thus, the EPFM analysis for the pump casings contained in WCAP-13045 is valid for 80 years because the minimum fracture toughness used in the stability analysis is applicable to 80 years of service life.

A qualitative evaluation for the fatigue crack growth evaluation was performed in Section 4 of this report for the pump casings. It was determined that the current FCG rate for stainless steel in a water environment based on the 2013 Edition of the ASME Code Section XI, when compared to the rates used in WCAP-13045, are comparable and there will be an insignificant impact on the crack growth analysis. Furthermore, the stresses used in the FCG analysis are generic and envelop the various pump designs. Additionally, these stresses do not change for 80 years of operation, as they are not time dependent. The stress intensity factors used in the FCG analysis are consistent with the current industry standards for similar FCG evaluations.

The transient definitions in the FCG analysis are also not expected to change over the design life, and the cycles used in the WCAP-13045 are assumed to bound the predicted 80 year transient cycles. Finally, there is sufficient margin between the final crack growth and the flaw size used for stability; therefore, even if the 40 year transient cycles are doubled for 80 years of operation, the final flaw size after FCG will be less than the stability flaw size, 1/4T flaw depth, for the stability analysis in WCAP-13045.

In conclusion, the fracture mechanics integrity evaluation in WCAP-13045 [1] for pump casings is applicable to 80 years of design life. The fracture mechanics evaluation for SLR in this TR justifies continuing to perform visual inspections, in lieu of volumetric inspections, for pump casings as incorporated in ASME Code Section XI.

PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6 REFERENCES

1) Westinghouse Document, WCAP-13045, "Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply System,"

September 1991.

2) ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components." * *

3) Code Case N-481, "Alternative Examination Requirements for Cast Austenitic Pump Casings,"Section XI, Division 1, Approval Date: March 5, 1990.

4) PWROG Project Authorization, PA-MSC-1498, Revision 0, "Update for Subsequent License Renewal: WCAP-13045, 'Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems'."

5) Westinghouse Document, WCAP-10931, Revision 1, "Toughness Criteria for Thermally Aged Cast Stainless Steel," July 1986.

6) Slama, G., Petrequin, P., Masson, S. H. and Mager, T. R., "Effect of Aging on Mechanical Properties of Austenitic Stainless Steel Casting Welds," presented at SMIRT 7 Post Conference Seminar 6 - Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, August 29/30, 1983, Monterey, CA.

7) Westinghouse Document, WCAP-10456, "The Effects of Thermal Aging on Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems,"

November 1983.

8) "Evaluation of Flaws in Austenitic Steel Piping," Trans. ASME, Journal of Pressure Vessel Technology, Vol. 108, pp. 352-366, 1986.

9) Bamford, W. H., "Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment," Trans ASME, Journal of Pressure Vessel Technology, February 1979.

10) Kumar, V., German, M. D. and Shih, C. P., "An Engineering Approach for Elastic-Plastic Fracture Analysis," EPRI Report NP-1931, Project 1237-1, Electric Power Research Institute, July 1981.

11) 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," NUREG/CR-4513, Revision 2, U.S. Nuclear Regulatory Commission, Washington, D.C., May 2016.

12) BMS-LGL-84, Revision 0.00, "Protection of Proprietary Information Regarding Submittals to the USNRC including Safety Analysis Reports for Commercial Nuclear Power Plants,"

Effective Date: April 15, 2017.

13) U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Salem Nuclear Generating Station," Docket Numbers 50-272 and 50-311, March 2011. (NRC ADAMS Accession No. ML110900295)

14) U.S. Nuclear Regulatory Commission, NUREG-1831, "Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2," Docket Nos. 50-315 and 50-316, July 2005. (NRC ADAMS Accession No. ML052230442)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2

15) U.S Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Diablo Canyon Nuclear Power Plant, Units 1 and 2," Docket Nos. 50-275 and 50-323, June 2011. (NRC ADAMS Accession No. ML11153A103)

16) U.S. Nuclear Regulatory Commission, NUREG-2181, "Safety Evaluation Report Related to the License Renewal of Sequoyah Nuclear Plant, Units 1 and 2," Docket Nos. 50-327 and 50-328, July 2015. (NRC ADAMS Accession No. ML15187A206)

17) Westinghouse Document, WCAP-16957-P, "A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Salem Generating Station Units 1 and 2 for the License Renewal Program," March 2009.

18) Westinghouse Document, WCAP-13128, "A Demonstration of Compliance of the Primary Loop Pump Casings of the D. C. Cook Units 1 and 2 to ASME Code Case N-481," March 1992.

19)Westinghouse Document, WCAP-13895, "A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings of the Diablo Canyon Nuclear Power Plants Units 1 and 2," October 1993.

PWROG-17033-NP June 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A : ASME CODE SECTION XI CODE CASE N-481 PWROG-17033-NP June 2018 Revision 1

1-Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 Attachment 4 PWROG-17011-NP, REVISION 1 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Update for Subsequent License Renewal: WCAP-14535A, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination" and WCAP-15666-A, "Extension of Reactor Coolant Pump Motor Flywheel Examination"

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWROG-17011-N P Revision 1 Update for Subsequent License Renewal: WCAP-14535A, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination" and WCAP-15666-A, "Extension of Reactor Coolant Pump Motor Flywheel Examination" PA-MSC-1500 Gordon Z. Hall*

Structural Design and Analysis - I Raymond E. Schneider*

Risk Applications and Methods - II May 2018 Reviewer: Earnest S. Shen*

Structural Design and Analysis - I Reviewer: John White*

Risk Applications and Methods - 11 Approved: John McFadden*, Manager Aging Management & License Renewal Approved: James P. Molkenthin*, Program Director PWR Owners Group PMO

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii ACKNOWLEDGEMENTS This report was developed and funded by the PWR Owners Group under the leadership of the participating utility representatives of the Materials Committee.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv WESTINGHOUSE ELECTRIC COMPANY LLC PROPRIETARY LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

1. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

2. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. Information in this report, is the property of, and contains copyright material owned by, Westinghouse Electric Company LLC and /or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document and the material contained therein in strict accordance with the terms and conditions of the agreement under which it was provided to you.

DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non-Proprietary reports to third parties that are supporting implementation at their plant, or for submittals to the USN RC.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V PWR Owners Group United States Member Participation* for PA-MSC-1500 Participant Utility Member Plant Site(s)

Yes No Ameren Missouri Callaway (W) X American Electric Power D.C. Cook 1 & 2 (W) X Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE) X Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X North Anna 1 & 2 (W) X Dominion VA Surry 1 & 2 (W) X Catawba 1 & 2 (W) X Duke Energy Carolinas McGuire 1 & 2 (W) X Oconee 1, 2, & 3 (B&W) X Robinson 2 (W) X Duke Energy Progress Shearon Harris (W) X Entergy Palisades Palisades (CE) X Entergy Nuclear Northeast Indian Point 2 & 3 (W) X Arkansas 1 (B&W) X Entergy Operations South Arkansas 2 (CE) X Waterford 3 (CE) X Braidwood 1 & 2 (W) X Byron 1 & 2 (W) X Exelon Generation Co. LLC TMI 1 (B&W) X Calvert Cliffs 1 & 2 (CE) X Ginna (W) X Beaver Valley 1 & 2 (W) X FirstEnergy Nuclear Operating Co.

Davis-Besse (B&W) X St. Lucie 1 & 2 (CE) X Turkey Point3 & 4 (W) X Florida Power & Light \ NextEra Seabrook (W) X Pt. Beach 1 & 2 (W) X Luminant Power Comanche Peak 1 & 2 (W) X Omaha Public Power District Fort Calhoun (CE)

X Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X PSEG - Nuclear Salem 1 & 2 (W) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Farley 1 & 2 (W) X Southern Nuclear Operating Co.

Vogtle 1 & 2 (W) X Sequoyah 1 & 2 (W) X Tennessee Valley Authority Watts Bar 1 & 2 (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Xcel Energy Prairie Island 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi PWR Owners Group I n ternaf1ona I Mem ber Pa rfICIPS. f IOn* for PA -MSC -1500 Participant Utilitv Member Plant Site(s) Yes No Asco 1 & 2 (W) X Asociaci6n Nuclear Asc6-Vandell6s Vandellos 2 (W) X AxpoAG Beznau 1 & 2 (W) X Centrales Nucleares Almaraz-Trillo Almaraz 1 & 2 (W) X EDF Enerav Sizewell B (W) X Dael 1, 2 & 4 (W) X Electrabel Tihange 1 & 3 (W) X Electricite de France 58 Units X Eletronuclear-Eletrobras Angra 1 (W) X Emirates Nuclear Enerqy Corporation Barakah 1 & 2 X EPZ Borssele X Eskom Koeberg 1 & 2 (W) X Hokkaido Tomari 1, 2 & 3 (MHI) X Japan Atomic Power Company Tsuruga 2 (MHI) X Mihama 3 (W) X Kansai Electric Co., LTD Ohi 1, 2, 3 & 4 (W & MHI) X Takahama 1, 2, 3 & 4 (W & MHI) X Kori 1, 2, 3 & 4 (W) X Hanbit 1 & 2 (W) X Korea Hydro & Nuclear Power Corp.

Hanbit 3, 4, 5 & 6 (CE) X Hanul 3, 4 , 5 & 6 (CE) X Genkai 2, 3 & 4 (MHI) X Kyushu Sendai 1 & 2 (MHI) X Nuklearna Electrarna KRSKO Krsko (W) X RinghalsAB Ringhals 2, 3 & 4 (W) X Shikoku lkata 1, 2 & 3 (MHI) X Taiwan Power Co. Maanshan 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii TABLE OF CONTENTS 1 INTRODUCTION ................................................................................................................................ 1-1 2 BACKGROUND .................................................................................................................................2-1 2.1 DESIGN AND FABRICATION ........................................................................................................ 2-1 2.2 INSPECTION .................................................................................................................................2-3 2.3 STRESS AND FRACTURE EVALUATION .................................................................................... 2-5 2.3.1 Selection of Flywheel Groups for Evaluation ......................................................................... 2-5 2.3.2 Ductile Failure Analysis ..........................................................................................................2-5 2.3.3 Non-ductile Failure Analysis ................................................................................................... 2-5 2.3.4 Fatigue Crack Growth ............................................................................................................2-6 2.3.5 Excessive Deformation Analysis ............................................................................................ 2-6 2.4

SUMMARY

OF STRESS AND FRACTURE RESULTS ................................................................ 2-7 3 RISK ASSESSMENT ......................................................................................................................... 3-1 3.1 RISK-INFORMED REGULATORY GUIDE 1.174 METHODOLOGY ............................................ 3-1 3.2 FAILURE MODES AND EFFECTS ANALYSIS ............................................................................. 3-7 3.3 FLYWHEEL FAILURE PROBABILITY ........................................................................................... 3-9 3.3.1 Method of Calculation Failure Probabilities .......................................................................... 3-1 O 3.3.2 Sensitivity Study ................................................................................................................... 3-14 3.3.3 Failure Probability Assessment Conclusions ....................................................................... 3-16 3.4 CORE DAMAGE EVALUATION .................................................................................................. 3-21 3.4.1 What is the Likelihood of the Event... ................................................................................... 3-22 3.4.2 What are the Consequences? .............................................................................................. 3-23 3.4.3 Risk Calculation .................................................................................................................... 3-23 3.5 CONSIDERATION OF UNCERTAINTY ...................................................................................... 3-29 3.5.1 Initiating Event Frequency .................................................................................................... 3-30 3.5.2 Conditional Flywheel Failure Probability .............................................................................. 3-30 3.5.3 Conditional Core Damage/Large Early Release Probability Associated with a Flywheel Failure Event ........................................................................................................................................ 3-30 3.5.4 Conclusion Regarding Treatment of Uncertainty ................................................................. 3-31 3.6 RISK RESULTS AND CONCLUSIONS ....................................................................................... 3-31 4 CONCLUSIONS .................................................................................................................................4-1 5 REFERENCES ...................................................................................................................................5-1 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii APPENDIX A: CALVERT CLIFFS UNIT 1 & 2 RCP MOTOR FLYWHEEL EVALUATIONS FOR EXTENSION OF ISi INTERVAL ............................................................................................................. A-1 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix List of Tables Table 2-1: RCP Flywheel Inspection Data .. :............................................................................... 2-3 Table 2-2: Flywheel Inspection Data Recordable lndications ..................................................... 2-4 Table 2-3: Flywheel Groups Evaluated for Program MUHP-5043 [2]. ........................................ 2-5 Table 2-4: Ductile Failure Limiting Speed ................................................................................... 2-5 Table 2-5: Critical Crack Lengths for Flywheel Overspeed of 1500 rpm (Considering LBB) ...... 2-6 Table 2-6: Fatigue Crack Growth Assuming 6000 RCP Starts and Stops .................................. 2-6 Table 2-7: Flywheel Deformation at 1500 rpm ........................................................................... 2-7 Table 3-1: Variables for RCP Motor Flywheel Failure Probability Model. ................................. 3-11 Table 3-2: Input Values for RCP Motor Flywheel Failure Probability Model. ............................ 3-12 Table 3-3: Cumulative Probability of Failure over 40, 60 and 80 Years with and without lnservice Inspection <1l ..................................................................................................... 3-14 Table 3-4: Effect of Flywheel Risk Parameter on Failure Probability ....................................... 3-15 Table 3-5: Summary of Flywheel Analysis Parameters ............................................................ 3-21 Table 3-6: Estimated RCP Motor Flywheel Failure Probabilities .............................................. 3-22 Table 3-7: Westinghouse RCP Motor Flywheel Evaluation Group 1 ........................................ 3-26 Table 3-8: RCP Motor Flywheel Evaluation Group 2 ................................................................ 3-27 Table 3-9: Calvert Cliffs Units 1 and 2 RCP Motor Flywheel Evaluation .................................. 3-28 Table 3-10: GDF Sensitivity to Variations in PRA evaluation assumptions for RCP Flywheel Failure Risk Assessment for Extending 10-year inspection intervals to 80 years -

(Flywheel Group 1) ........................................................................................... 3-31 Table 3-11: Evaluation with Respect to Regulatory Guide 1.174 (Key Principles) ................... 3-32 Table A-1: Critical Crack Length in Inches and% Through Flywheel. ....................................... A-2 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 X List of Figures Figure 2-1: Example of a Typical Westinghouse RCP Motor Flywheel. ..................................... 2-2 Figure 3-1: NRC Regulatory Guide 1.174 Basic Steps ............................................................. 3-2 Figure 3-2: Principles of Risk-Informed Regulation [5] ............................................................. 3-4 Figure 3-3: Westinghouse PROF Program Flow Chart for Calculating Failure Probability ...... 3-17 Figure 3-4: Probability of Failure for Flywheel Evaluation Group 1 .......................................... 3-18 Figure 3-5: Probability of Failure for Flywheel Evaluation Group 2 .......................................... 3-19 Figure 3-6: Probability of Failure for Calvert Cliffs Units 1 and 2 ............................................ 3-20 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 xi List of Acronyms B&W Babcock and Wilcox CCDP conditional core damage probability CCL critical crack length CCNPP Calvert Cliffs Nuclear Power Plant CDF core damage frequency CE Combustion Engineering DEGB double ended guillotine break OLE design limiting events FCG fatigue crack growth FSAR final safety analysis report FSAR final safety analysis report GQA graded quality assurance ISi inservice inspection 1ST inservice testing LBB leak-before-break LERF large early release frequency LOCA loss of coolant accident LOOP loss offsite power MT magnetic particle testing NOE non-destructive examination NRC Nuclear Regulatory Commission OD outer diameter PMSC Pump & Motor Services PRA probabilistic risk assessment PROF probability of failure PT penetrant testing PWROG Pressurized Water Reactor Owners Group RCP reactor coolant pump RCPM reactor coolant pump motor RCS reactor coolant system RG regulatory guide rpm revolutions per minute RTNDT reference nil-ductility transition temperature SER safety evaluation report SLR subsequent license renewal SRP standard review plan SRRA structural reliability and risk assessment SSCs systems, structures and components USAR updated safety analysis report UT ultrasonic examination/ultrasonic test w Westinghouse WOG Westinghouse Owners Group W-PROF Westinghouse PROF Software Library PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1 INTRODUCTION The purpose of this topical report (TR) is to extend the applicability of WCAP-14535A [1]

and WCAP-15666-A [2] to subsequent license renewal (SLR), i.e., 80 years of operation.

Westinghouse provided the technical basis in WCAP-14535A [1] for the elimination of inspection requirements for the reactor coolant pump (RCP) motor flywheels for all operating domestic Westinghouse and several B&W plants. The NRC issued a Safety Evaluation Report (SER) in September 12, 1996, accepting the technical arguments but did not allow for total elimination of examinations as WCAP-14535A [1] requested. The SER provided partial relief from the reactor coolant pump (RCP) motor flywheels examination requirements in NRC RG 1.14 [3], by allowing an extension in the examination frequency from 40 months to 10 years. It further relaxed the RG 1.14 examination guidance by recommending an in-place ultrasonic examination (UT) over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or an alternative surface examination, i.e., magnetic particle testing (MT) and/or liquid penetrant testing (PT), of the exposed surfaces defined by the volume of disassembled flywheel. As Section 3.6 of the SER for [1] stated, NRC staff relied solely on the deterministic methodology to review the submittal. The risk assessment was not included in [1] and was not reviewed. WCAP-14535A [1] is applicable to the RCP motor flywheels in all domestic Westinghouse nuclear steam supply system (NSSS) plants, and Oconee Units 1, 2, and 3, Davis Besse, and Three Mile Island Unit 1, which are Babcock and Wilcox (B&W) NSSS plants.

WCAP-15666-A [2] is a follow-up TR that justified extending the 10-year inspection frequency that was approved by the NRC in WCAP-14535A [1] to 20 years. WCAP-15666-A [2] demonstrated that the deterministic results in WCAP-14535A [1] remain valid, and also performed a failure probability analysis to show that the change in risk for a 20-year inspection frequency meet the RG 1.174 [5] acceptance guidelines. The NRC SER for WCAP-15666-A [2] concluded that both the deterministic and probabilistic calculations contained in [2] were acceptable, and approved the 20-year inspection frequency.

WCAP-15666-A [2] is applicable to plants with Westinghouse-designed NSSS plants.

Although it included some data for B&W NSSS plants, however, the TR and the NRC SER did not specifically address the applicability of the risk assessments and other evaluations to the three B&W NSSS plants that WCAP-14535-A [1] was applicable to.

The following is a quote from the NRC SER for [2].

"The NRG staff acknowledges that some of the supporting material for TSTF-421 may also help to support plant-specific applications for the B&W units included in portions of WGAP-15666. The NRG staff will work with licensees for the applicable B& W units to ensure that our processes work as efficiently as possible for those applying for license amendments similar to that described in TSTF-421. The affected licensees are encouraged to discuss this matter with the NRG staff before submitting an application."

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 This same applicability is carried over for the TR presented herein. This TR is not applicable to Combustion Engineering (CE) NSSS plants, with the exception of Calvert Cliffs Units 1 and 2. This TR is applicable to Calvert Cliff Units 1 and 2 as these plants have Westinghouse RCP motors and flywheels. However, these flywheels and motor operating speeds are different than those evaluated in WCAP-15666-A [2].

Westinghouse performed a plant-specific evaluation for Calvert Cliffs Unit 1 and 2, that applied the using the same methods detailed in WCAP-15666-A [2] for 60 years of operation. This 60-year evaluation is extended to 80 years of operation in this TR.

Revision 1 of this TR removes unnecessary contents that are duplicates in WCAP-14535A [1] and WCAP-15666-A [2]. Change bars are not used. All evaluation results and conclusions are unchanged.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 BACKGROUND 2.1 DESIGN AND FABRICATION Westinghouse RCP motor flywheels consist of two large steel discs that are shrunk fit directly to the RCP motor shaft. The individual flywheel discs are bolted together to form an integral flywheel assembly, which is located above the RCP rotor core. Typically, each flywheel disc is keyed to the motor shaft by means of three vertical keyways, positioned at 120° intervals. The bottom disc usually has a circumferential notch along the outside diameter bottom surface for placement of anti-rotation pawls. See Figure 2-1 for the configuration of a typical Westinghouse flywheel.

Westinghouse has manufactured the RCP motors for all operating Westinghouse plants.

All of the RCP motor flywheels for Westinghouse plants are made of SA-533 Grade 8 Class 1 steel. As in WCAP-15666-A [2], a range of RTNoT values from 0°F to 60°F was assumed in the integrity evaluations of [1], which are discussed later in this report.

Westinghouse designed flywheels are also used for Calvert Cliffs Units 1 and 2. They will be addressed separately in Section 3 for the risk assessment, and in APPENDIX A:

for the deterministic evaluations.

Consistent with the evaluations performed in [1 ], larger flywheel outside diameter for the flywheel assembly is used in this TR, because it is conservative with respect to stress and fracture.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 I

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Figure 2-1: Example of a Typical Westinghouse RCP Motor Flywheel PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 2.2 INSPECTION Flywheels are inspected at the plant or during motor refurbishment at an offsite facility.

Inspections are conducted under the ASME Boiler and Pressure Vessel Code,Section XI [4], which identifies the standard practice for control of instrumentation and personnel qualification. Ultrasonic test (UT) level II and Ill examiners conduct the inspections.

WCAP-15666-A [2] discussed the examination volume, approach, access and exposure in detail. This discussion remains applicable for SLR.

Inspection History The flywheel inspection results and the summary of recordable indications from the MUHP-5042 study are presented in Table 2-3 and Table 2-4 of WCAP-15666-A [2]

Inspection History Update A summary of all Westinghouse RCP flywheels that were inspected by Framatome (formerly AREVA) is summarized in Table 2-1.

Four RCP flywheels where determined to have recordable indications. All four indications were determined to be non-relevant; no repairs were required to be performed on any of those RCP flywheels.

The four recordable indications are discussed in Table 2-2.

Table 2-1: RCP Flywheel Inspection Data Total Number Total Number of Total of Inspection Number of Indications Number of Number of Plant with No Indications Inspection With Affecting Flywheels Flywheel or Non-recordable Recordable Flywheel Inspections Indications Indications Integrity A 9 9 8 1 0 8 9 9 8 1 0 C 2 2 2 0 0 D 9 14 14 0 0 E 2 2 2 0 0 F 3 3 3 0 0 G 7 7 7 0 0 H 13 13 13 0 0 I 1 1 1 0 0 J 9 9 8 1 0 K 1 1 1 0 0 L 8 9 8 1 0 M 1 1 1 0 0 N 1 1 1 0 0 Total 75 81 77 4 0 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4 Table 2-2: Flywheel Inspection Data Recordable Indications Plant Year Description of Recordable Indications A Recordable UT indication - Accepted. Lamination with 50% of back wall A 2015 loss 1" x 4".

Procedurally recordable UT indications were identified in the bottom B 2006 flywheel plate during the 45 degree shear wave examination. - Accepted per NB-2530.

Indications were identified in two of three keyways in the lower thickness.

J 2005 The indications were dispositioned as acceptable because they are considered to be "non-relevant due to the machining process."

These were determined to be non-relevant indications. There were several low amplitude responses that were identified during the radial L 2012 examinations. These responses were indicative of small machine grooves or marks that extend 360° around the flywheel.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-5 2.3 STRESS AND FRACTURE EVALUATION Section 2.3 of WCAP-15666-A [2] summarized the stress and fracture evaluation. The ductile and brittle failure mechanisms were considered in flywheel evaluation. The methodology is unchanged for this TR. The evaluation requirements are per RG 1.14

[3].

2.3.1 Selection of Flywheel Groups for Evaluation As discussed in [2], stresses in the flywheel are a strong function of the outer diameter (approximately proportional to the square of the OD dimension). Therefore, the two groups shown in Table 2-3 with the largest flywheel outer diameter (Groups 1 and 2) bound all other groups defined in WCAP-15666-A [2], and were selected for the deterministic and probabilistic evaluations.

Table 2-3: Flywheel Groups Evaluated for Program MUHP-5043 [2]

Flywheel Outer Bore Keyway Radial Comments Evaluation Group Diameter (inch) (inch) Length (inch) 1 76.50 9.375 0.937 Maximum OD.

2 75.75 8.375 0.906 Large OD, minimum bore.

2.3.2 Ductile Failure Analysis The flywheel stresses are dependent on dimensions and rotation speed. Extending the operating period to 80 years does not affect the stress calculation. Therefore, the ductile failure analysis in [2] remains valid for 80 years of operation.

These results from [2] are summarized in Table 2-4. The RG 1.14 acceptance criteria for ductile failure of the flywheels are satisfied.

Table 2-4: Ductile Failure Limiting Speed Crack Length (as measured from the Assuming No Cracks maximum radial location of the keyway)

Flywheel Evaluation Neglecting Considering Group Keyway Keyway 1" Crack 2" Crack 5" Crack 10" Crack Radial Radial Length Length 1 3487 3430 3378 3333 3240 3012 2 3553 3493 3435 3386 3281 3060 2.3.3 Non-ductile Failure Analysis The flywheel stress intensity factor, K1, is dependent on geometry, postulated flaw dimensions and stress condition (due to rotation speed). Extending the operating period to 80 years does not affect the K1 calculations. Furthermore, the flywheel is not local or adjacent to the reactor core; therefore, the effect of irradiated embrittlement is negligible, and the fracture toughness, K1c does not change due to the 80-year extension.

Therefore, the non-ductile failure analysis in [2] remains valid for 80 years of operati9n.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-6 The results from [2] are shown in Table 2-5. The ambient temperature of 70°F was conservatively used as the operating temperature, while the typical containment ambient temperature is 100°F to 120°F. At the maximum flywheel overspeed condition of 1500 rpm (considering LBB), the critical crack lengths were calculated for cracks emanating radially from the keyway. The crack length is defined as radially from the keyway. The percentage through the flywheel is defined as the crack length divided by the radial length from the maximum radial keyway location to the flywheel outer radius, i.e.,

percentage through-wall. The critical crack lengths are quite large, even when considering higher values of RTNDT and a lower than expected operating temperature.

Table 2-5: Critical Crack Lengths for Flywheel Overspeed of 1500 rpm (Considering LBB)

Flywheel Critical Crack Length in Inches and % through Flywheel Evaluation Group RTNDT = 0°F RTNDT = 30°F RTNDT = 60°F 16.6" 7.7" 3.1" 1

(50%) (24%) (9%)

17.5" 8.5" 3.6" 2

(53%) (26%) (11%)

2.3.4 Fatigue Crack Growth FCG is dependent on the flywheel K1at operating and rest states (~K,), and the number of start and shutdown cycles. As discussed previously, the 80-year extension has no impact on the K1 calculations. The 6000 cycles used in the FCG calculation of [2] was determined to be bounding for 80 years of operation. The 6000 cycles for 80 years of operation must be confirmed to be applicable on a plant-specific basis.

The FCG calculations assumed the 6000 cycles of RCP start and shutdown for the 80-year plant life. The FCG results from [2] are applicable and are shown in Table 2-6. The crack growth is negligible over an 80-year life of the flywheel, even when assuming a conservative initial crack length as shown in Table 2-6.

Table 2-6: Fatigue Crack Growth Assuming 6000 RCP Starts and Stops Assumed Crack Keyway Length Flywheel Flywheel Flywheel Initial Growth Radial From ~K1 Evaluation OD Bore Crack after6000 Length Keyway to (ksi'1in)

Group (inch) (inch) Length cycles (inch) OD (inch)

(inch) (inch) 1 76.50 9.375 0.937 32.63 3.26 38 0.08 2 75.75 8.375 0.906 32.78 3.28 37 0.08 2.3.5 Excessive Deformation Analysis The deformation of the flywheel is only dependent on the rotation speed and physical attributes of the flywheel. The 80-year extension has no impact on the excessive deformation analysis of the flywheel. The results in [2] remain applicable to 80 years of operation.

At the flywheel over speed condition of 1500 rpm (157.08 radians/second), the change in the bore radius and outer radius is shown in Table 2-7. A maximum deformation of 0.006 inch is anticipated for the flywheel over speed condition. As deformation is proportional PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-7 to the square of angular speed, al, this represents an increase of 56% over the normal operating deformation of 0.004 inch. This increase would not result in any adverse conditions such as excessive vibrational stress leading to crack propagation, since the flywheel assemblies are typically shrunk fit to the flywheel shaft, and the deformations calculated are negligible.

Table 2-7: Flywheel Deformation at 1500 rpm Flywheel Evaluation Group Change in Bore Radius (inch) Change in Outer Radius (inch) 1 0.003 0.006 2 0.003 0.006

, 2.4

SUMMARY

OF STRESS AND FRACTURE RESULTS The deterministic integrity evaluations in WCAP-15666-A [2] remain applicable for 80 years of operation. The evaluations concluded that the RCP motor flywheels have a very high tolerance for the presence of flaws, especially with the 1500 rpm overspeed due to the application of LBB [2]. As noted in [2], the probabilistic assessment evaluates all credible flywheel speeds. This TR uses the same probabilistic assessment methodology as [2], which is discussed in Section 3.

There are no significant mechanisms for inservice degradation of the flywheels, since they are isolated from the primary coolant environment. The evaluations presented in this section have shown there is no significant deformation of the flywheels, even at maximum overspeed conditions. FCG calculations have shown that even with a large assumed flaw, the crack growth for 80 years of operation is negligible. Therefore, based on these deterministic evaluations, the flywheel inspections completed following manufacture and prior to service are sufficient to ensure their integrity during 80 years of service. As discussed in Section 2.2 and [1 and 2], the most likely source of inservice degradation is damage to the keyway region that could occur during disassembly or reassembly for refurbishment and inspection.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 RISK ASSESSMENT The quantitative risk assessment discussed below provides the justification for applying the WCAP-15666-A [2] 20-year flywheel inspection interval for 80 years of operation.

Specifically, the risk analyses confirms that applying the inspection extension to flywheels in operation up to 80 years has a negligible impact on risk (CDF and LERF),

i.e., it is within the risk acceptance criteria of RG 1.174 [5]. This section provides a discussion on the requirements of [5], and extends the previous flywheel failure probability assessment in [2] to 80 years of operation.

3.1 RISK-INFORMED REGULATORY GUIDE 1.174 METHODOLOGY The NRC risk-informed regulatory framework for modifying a plant's licensing basis is contained in RG 1.174, Revision 2 [5]. The intent of this risk-informed process is to allow insights derived from probabilistic risk assessments to be used in combination with traditional engineering analysis to focus licensee and regulatory attention on issues commensurate with their importance to safety. Additional regulatory guidance is contained in [6].

The approach described in RG-1.17 4 is used in each of the application-specific RGs/SRPs, and has four basic steps as shown in Figure 3-1. The four (4) basic steps are discussed below.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 Traditional Analysis -- - ~ PRA I ,

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and Monitoring ~ Proposed Program Change Principal Elements of Risk-Informed, Plant-Specific Oecisionmaking (from NRC Regulatory Guide RG-1.174)

Figure 3-1: NRC Regulatory Guide 1.174 Basic Steps Step 1: Define the proposed change This element includes identifying:

1. Those aspects of the plant's licensing bases that may be affected by the change

2. All systems, structures, and components (SSCs), procedures, and activities that are covered by the change and consider the original reasons for inclusion of each program requirement

3. Any engineering studies, methods, codes, applicable plant-specific and industry data and operational experience, PRA findings, and research and analysis results relevant to the proposed change. '

Step 2: Perform engineering analysis This element includes performing the evaluation to show that the fundamental safety principles on which the plant design was based are not compromised (defense-in-depth attributes are maintained) and that sufficient safety margins are maintained. The engineering analysis includes both traditional deterministic analysis and probabilistic risk assessment. The evaluation of risk impact should also assess the expected change in CDF and LERF, including a treatment of uncertainties. The results from the traditional

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 analysis and the probabilistic risk assessment must be considered in an integrated manner when making a decision.

Step 3: Define implementation and monitoring program This. element's goal is to assess SSC performance under the proposed change by establishing performance monitoring strategies to confirm assumptions and analyses that were conducted to justify the change.

This is to ensure that no unexpected adverse safety degradation occurs because of the changes. Decisions concerning implementation of changes should be made in light of the uncertainty associated with the results of the evaluation. A monitoring program should have measurable parameters, objective criteria, and parameters that provide an early indication of problems before becoming a safety concern. In addition, the monitoring program should include a cause determination and corrective action plan.

Step 4: Submit proposed change This element includes:

1. Carefully reviewing the proposed change in order to determine the appropriate form of the change request

2. Assuring that information required by the relevant regulation(s) in support of the request is developed

3. Preparing and submitting the request in accordance with relevant procedural requirements.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Five (5) fundamental safety principles are described which should be met for each application for a modification. These are shown in Figure 3-2 and are discussed below.

Change is consistent with defem&-in-depth OlarlQe meets rurrent philosophy. IVlaintain sufficient regulations unless tt is safety margins.

explicitly related 1o a requested exerrption or rule <.flange.

Integrated Decision making Proposed increases in Use performance- COF or risk are small rreasurement and are oonsistent with strategies to rronltor 1he O:mmission's safety Iha change. Goal Policy Statement.

Figure 3-2: Principles of Risk-Informed Regulation [5]

Principle 1: Change meets current regulations unless it is explicitly related to a requested exemption or rule change The proposed change is evaluated against the current regulations (including the general design criteria) to either identify where changes are proposed to the current regulations (e.g., technical specification, license conditions, and FSAR), or where additional information may be required to meet the current regulations.

Principle 2: Change is consistent with defense-in-depth philosophy Defense-in-depth has traditionally been applied in reactor design and operation to provide a multiple means to accomplish safety functions and prevent the release of radioactive material. As defined in RG-1.174, defense-in-depth is maintained by assuring that:

A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5

System redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences to the system (e.g., no risk outliers)

Defenses against potential common cause failures are preserved and the potential for introduction of new common cause failure mechanisms is assessed.

Independence of barriers is not degraded (the barriers are identified as the fuel cladding, reactor coolant pressure boundary, and containment structure)

Defenses against human errors are preserved Defense-in-depth philosophy is not expected to change unless:

A significant increase in the existing challenges to the integrity of the barriers occurs

The probability of failure of each barrier changes significantly,

New or additional failure dependencies are introduced that increase the likelihood of failure compared to the existing conditions, or

The overall redundancy and diversity in the barriers changes.

Principle 3: Maintain sufficient safety margins Safety margins must also be maintained. As described in RG-1.17 4, sufficient safety margins are maintained by assuring that:

Codes and standards, or alternatives proposed for use by the NRC, are met, and

Safety analysis acceptance criteria in the licensing basis (e.g., FSARs, supporting analyses) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainty.

Principle 4: Proposed increases in GDF or risk are small and are consistent with the Commissions Safety Goal Policy Statement To evaluate the proposed change with regard to a possible increase in risk, the risk assessment should be of sufficient quality to evaluate the change. The expected change in CDF and LERF are evaluated to address this principle. An assessment of the uncertainties associated with the evaluation is conducted. Additional qualitative assessments are also performed.

There are two acceptance guidelines, one for CDF and one for LERF, both of which should be used.

The guidelines for CDF are:

If the application can be clearly shown to result in a decrease in CDF, the change will be considered to have satisfied the relevant principle of risk-informed regulation with respect to CDF.

When the calculated increase in CDF is very small, which is taken as less than 1o-6 per reactor year, the change will be considered regardless of whether there is a calculation of the total CDF.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6

When the calculated increase in CDF is in the range of 10- 6 per reactor year to 10-5 per reactor year, applications will be considered only if it can be reasonably shown that the total CDF is less than 104 per reactor year.

Applications which result in increases to CDF above 10*5 per reactor year would not normally be considered.

AND The guidelines for LERF are:

If the application can be clearly shown to result in a decrease in LERF, the change will be considered to have satisfied the relevant principle of risk-informed regulation with respect to LERF

When the calculated increase in LERF is very small, which is taken as being less than 10-7 per reactor year, the change will be considered regardless of whether there is a calculation of the total LERF.

When the calculated increase in LERF is in the range of 10*7 per reactor year to 10-6 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 10-5 per reactor year.

Applications which result in increases to LERF above 10*5 per reactor year would not normally be considered.

These guidelines are intended to provide assurance that proposed increases in CDF and LERF are small and are consistent with the intent of the Commission's Safety Goal Policy Statement.

Principle 5: The impact of the proposed change should be monitored using performance-measurement strategies to monitor the change Performance-based implementation and monitoring strategies are also addressed as part of the key elements of the evaluation as described previously.

The following sections address the principle elements of the RG-1.17 4 process and the principles of risk-informed regulation to RCP motor flywheel examination frequency reduction. '

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 3.2 FAILURE MODES AND EFFECTS ANALYSIS A failure modes and effects analysis is used to identify the potential failure modes of a RCP motor flywheel and the effect that each failure mode would have on the plant SSCs in relation to overall plant safety.

Failure Modes The primary failure mode of the RCP motor flywheel is growth of an undetected fabrication induced flaw in the keyway of the flywheel that emanates radially from that location to a point such that it reaches a critical flaw size during normal or accident conditions. Once the critical flaw size is reached during plant operation, the flywheel has the potential to catastrophically fail, resulting in flywheel fragments, which are essentially high energy missiles that could impact other SSCs important to plant safety. The growth of a flaw is primarily related to stresses generated from changes in the flywheel speed.

The flywheel inspection process, which itself has the potential to introduce flywheel damage as discussed in [1 ], is not considered in the assessment. This is because the purpose of the assessment is to support interval extension, which will reduce unnecessary occurrences for introducing potential damage.

As discussed in [1 ], the normal operating speed of the RCP motor flywheel for Westinghouse RCPs is 1189 revolutions per minute (rpm), with a synchronous speed of 1200 rpm. It is designed for an overspeed of 1500 rpm, which is 125% of the synchronous speed. The flywheel speed can also vary as a result of plant events, including accidents such as a double ended guillotine break (DEGB) in the main reactor coolant loop piping.

Westinghouse designed flywheels are also used for Calvert Cliffs Units 1 and 2. These plants include Byron-Jackson designed pumps and motors and therefore have different normal flywheel operating speeds and different flywheel accident responses. The normal operating speed of the RCP motor flywheel for these RCPs is 900 rpm, with a design limiting speed of 1125 rpm. The maximum overspeed following a design basis LOCA is limited to 1368 rpm as stated in the Calvert Cliffs UFSAR and WCAP-15666-A

[2].

When operating as a motor, the rotor of a polyphase inquction machine rotates in the direction of, but slightly lower than, the rotating magnetic flux provided by the stator.

This slight speed difference is typically expressed in percent and designated slip. If the shaft of the machine is driven above synchronous speed by a prime mover (with line voltage maintained on the stator) the rotor conductors rotate faster than the magnetic flux and the slip becomes negative. The rotor current and consequently the stator current reverse under the condition of negative slip and the machine operates as an induction, or asynchronous, generator. The RCP motor functions as an efficient torque producer under normal conditions. In the unlikely event that a hydraulic torque is applied to the motor shaft in the direction of increasing shaft speed (thus acting as a prime mover), the slip would become negative and, with the stator connected to the grid, the motor would function as a dynamic brake.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 If the power supply to the motor is interrupted (zero voltage), the motor torque would be reduced to a negligible value, since torque is proportional to the supplied voltage.

However, a design feature of Westinghouse NSSS plants ensures that the electrical power supply to the RCP will be maintained for at least 30 seconds after a turbine trip following a LOCA. This design feature is also maintained following a loss of offsite power (LOOP); for the expected case of available off-site power, power to the RCP would continue through the LOCA transient. As a result, reverse torque is provided.

Westinghouse also performed several sensitivity studies to evaluate the effect of the break opening area on the RCP flywheel speed for typical Westinghouse NSSS plants.

Specifically, break sizes equal to a DEGB of the main coolant piping, 60% of that DEGB, and a 3 ft2 have been analyzed. A 3 ft2 break size corresponds to a pipe of approximately 23 inches in inside diameter; the only RCS piping greater than this, is the main coolant loop piping. The first two breaks have blowdown times equal to or less than the RCP trip time; therefore, the applied voltage prevents overspeed. The latter break has an extended blowdown time, but the RCP flow at the time of RCP trip is reduced such that the speed decreases. Smaller breaks are not limiting even though the voltage is maintained for only 30 seconds. Results of these studies were discussed in [1 ].

To investigate the consequences of RCP overspeed, [2] analyzed a spectrum of LOCA events resulting in a range of flywheel transients. Results of that analysis indicated that the limiting event was the DEGB with an instantaneous loss of power, this led to a peak flywheel speed of 3321 rpm. It was also noted that the 3 ft2 break area case showed a decrease in speed such that the normal operating speed is riot exceeded.

Based on the WCAP-15666-A assessments, the following scenarios are associated with the primary mode of potential failure in the Westinghouse RCP motor flywheel that are-related to operating speed and potential overspeed during various conditions:

Failure during normal plant operation resulting in a plant trip (1200 rpm peak speed)

Failure of the RCP motor flywheel associated with a plant transient or LOCA event with no loss of electrical power to the RCP (1200 rpm peak speed)

Failure of the RCP motor flywheel associated with a plant transient or LOCA event (up to 3 ft2 with an instantaneous loss of electrical power to the RCP (1200 rpm peak speed)

Failure of the RCP motor flywheel associated with a DEGB coincident with an instantaneous loss of electrical power, such as LOOP (3321 rpm peak speed).

This case bounds and is conservatively applied to all flywheel transients for LOCA break areas.

WCAP-15666-A [2] was limited in scope to RCPs with Westinghouse supplied pumps and flywheels. It is also the intent of this topical report to extend the applicability of the flywheel inspection extension to Calvert Cliffs Units 1 and 2 which contain Byron Jackson RCPs but use Westinghouse supplied flywheels. It is important to note that as a result of significant design differences between the Westinghouse and Calvert Cliffs PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-9 units, the Calvert Cliffs RCP operational and transient conditions are different.

Specifically, Calvert Cliffs pumps normally operate at 900 rpm with a design speed of 1125 rpm. Furthermore, the peak RCP post LOCA speed is limited to 1368 rpm.

Therefore, the Calvert Cliffs Units 1 and 2 Pump/Flywheel Combinations analyses were based on the following:

Failure during normal plant operation resulting in a plant trip (1125 rpm peak speed)

Failure of the RCP motor flywheel associated with a plant transient or LOCA event with no loss of electrical power to the RCP ( 1125 rpm peak speed)

Failure of the RCP motor flywheel associated with a plant transient or LOCA event (up to 3 ft2) with an instantaneous loss of electrical power to the RCP (1125 rpm peak speed)

Failure of the RCP motor flywheel associated with a DEGB coincident with *an instantaneous loss of electrical power, such as, loss of offsite power (LOOP)

(1368 rpm peak speed). As for the Westinghouse flywheel analysis, this case is conservatively assumed to bound all flywheel transients for LOCA break areas resulting from equivalent reactor coolant pipe breaks greater than a 3.0 ft2 break and less than a double ended break.

Failure Effects The failure of the RCP motor flywheel during normal plant operation would directly result in a reactor trip. However, the potential indirect or spatial effects associated with a postulated flywheel failure present a greater challenge in terms of failure effects or consequences. As discussed previously, the flywheel has the potential to catastrophically fail, resulting in flywheel fragments, which are essentially high energy missiles, which could impact other SSCs important to plant safety. Failure of these other SSCs could potentially impact the overall plant safety in terms of core damage (e.g., as a result of the loss of safety injection) or large, early release (as a result of potential impacts on containment structures or systems).

In order to address plant specific design differences on a generic basis, it is conservatively assumed that failure of the RCP motor flywheel results in core damage and a large early release, i.e., the flywheel failure frequency is equal to GDF and LERF.

Section 3.3 discusses the process for estimating the likelihood of the primary failure mode of the RCP motor flywheel. Section 3.4 then combines this failure probability estimation with the likelihood of various plant events and consequences to estimate the change in risk for extending the flywheel examination interval from 10 years to 20 years, for RCP/Flywheels in service up to 80 years.

3.3 FLYWHEEL FAILURE PROBABILITY To investigate the effect of flywheel inspections on the risk of failure, a structural reliability and risk assessment is performed, a 40-year plant life including up to 80 years PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-10 of operation. Twelve (12) month operating (or fuel) cycles are conservatively assumed for the evaluation. This section discusses the methodology used and summarizes the results from this assessment.

As described in Section 3.2, the Westinghouse RCP has a normal operating speed of 1189 rpm, a synchronous speed of 1200 rpm, and a design speed of 1500 rpm, per [2].

Therefore, a peak speed of 1500 rpm is conservatively used in the evaluation of RCP motor flywheel integrity to represent all conditions except a DEGB coincident with an instantaneous loss of electrical power. For this more limiting event, a peak speed of ,

3321 rpm is used.

The structural reliability evaluation for a Westinghouse RCP utilizes the work previously performed and summarized in [1 ], where the 1500 rpm design speed had been assumed. In addition, this evaluation builds upon the initial analysis discussed in [2].

The structural reliability evaluation for the Calvert Cliffs RCP is based on plant-specific analyses and flywheel failure probabilities which are based on nominal and transient flywheel operation at 1125 rpm and a post design basis LOCA flywheel transient overspeed of 1368 rpm.

3.3.1 Method of Calculation Failure Probabilities The method for calculating flywheel failure probabilities is based on the method in WCAP-15666-A [2].

The probability of failure of the RCP motor flywheel as a function of operating time t, Pr(t < t1 ), is calculated directly for each set of input values using Monte-Carlo simulation with importance sampling. The Monte-Carlo simulation does not force the calculated distribution of time to failure to be of a fixed type (e.g., Weibull, Log-normal or Extreme Value). The actual failure distribution is estimated based upon the distributions of the uncertainties in the key structural reliability model parameters and plant specific input parameters. Importance sampling, as described by Witt [7], is a variance reduction technique to greatly reduce the number of trials required for calculating small failure probabilities. In this technique, random values are selected from the more severe regions to increase the probability of an observable failure occurring. However, when a failure is calculated, the count is corrected to account for the lower probability of simultaneously obtaining all of the more severe random values.

The application of the probability of failure methodology is described based on the Westinghouse RPFWPROF program which is generally described in WCAP-14535A [1]

and WCAP-15666-A [2].

The description of the key input parameters and associated data used in the RPFWPROF program is presented in Table 3-1 and Table 3-2. Table 3-1 includes the key parameters needed for failure probability calculation. Its usage in the program is specified as shown in the last column of Table 3-1 and schematically in the flow chart of Figure 3-3. "Initial" conditions do not change with time, "Steady-State" is not needed for RPFWPROF, "Transient" calculates fatigue crack growth and "Failure" checks to see if the accumulated crack length exceeds the critical length. In addition, parameter RPM-OLE is included in the model to address the impact of design limiting events (OLE).

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-11 Table 3-1: Variables for RCP Motor Flywheel Failure Probability Model No. Name Description of Input Variable Usage Type 1 ORADIUS Outer Flywheel Radius (inch) Initial 2 !RADIUS Inner Flywheel Radius (inch) Initial 3 PFE-PSI Probability of Flaw Existing (PFE) after Preservice ISi Initial 4 !LENGTH Initial Radial Flaw Length (inch) Initial 5 CY1-ISI Operating Cycle for First lnservice Inspection Inspection 6 DCY-ISI Operating Cycle between lnservice Inspections Inspection 7 POD-ISi Flaw Detection Probability per lnservice Inspection Inspection 8 DFP-ISI Fraction PFE Increases per lnservice Inspection Inspection 9 NOTR/CY Number of Transients per Operating Cycle Transient 10 DRPM-TR Speed Change per Transient (RPM) Transient 11 RATE-FCG Fatigue Crack Growth Rate (Inch/Transient) Transient 12 KEXP-FCG Fatigue Crack Growth Rate SIF Exponent Transient 13 RPM-OLE Speed for Design Limiting Event (RPM) Failure 14 TEMP-F Temperature for Design Limiting Event (F) Failure 15 RT-NOT Reference Nil Ductility Transition Temperature (F) Failure 16 F-KIC Crack Initiation Toughness Factor Failure 17 DLENGTH Flywheel Keyway Radial Length (Inch) Failure Variables 5 to 8 are available to calculate the effects of an ISi in the RPFWPROF program. The effect of ISi calculated using these equations, which are used in the SRRA model for the effect of ISi, are consistent with those described in the pc-PRAISE Code User's Manual [1 O]. The parameters needed to describe the selected ISi program are the time of the first inspection, the frequency of subsequent inspections (expressed as the number of fuel or operating cycles between inspections) and the probability of non-detection as a function of crack length. For the RCP motor flywheel, the non-detection probability, which is independent of crack length, is simply one minus a constant value of detection probability, variable 7 (POD-ISi) in Table 3-1. An increase in failure probability due to RCP inspection (chance of incorrect disassembly and reassembly) is included in the ISi model but conservatively not used (variable 8 set to zero) in this evaluation.

The median input values and their uncertainties for each of the parameters of Table 3-1 are shown in Table 3-2. The median is the value at 50% probability (half above and half below this value); it is also the mean (average) value for symmetric distributions, like the normal (bell-shaped curve) distribution.

Uncertainties are based upon expert engineering judgment and previous structural reliability modeling experience. For example, the fracture toughness for initiation as a function of the RTNDT and the uncertainties on these* parameters are based upon prior probabilistic fracture mechanics analyses of the reactor pressure vessel (RPV) [11 ].

Also note that the stress intensity factor calculation for crack growth and failure used the flywheel keyway radial length in addition to the calculated flaw length.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-12 Table 3-2: Input Values for RCP Motor Flywheel Failure Probability Model No. Name Median Distribution Uncertainty*

1 ORADIUS Per Flywheel Group Constant -----

2 !RADIUS Per Flywheel Group Constant -----

3 PFE-PSI 1.000E-01 Constant -----

4 !LENGTH 1.000E-01 Log-Normal 2.153E+OO 5 CY1-ISI 3.000E+OO Constant ----

6 DCY-ISI 4.000E+OO Constant -----

7 POD-ISi 5.000E-01 Constant -----

8 DFP-ISI O.OOOE+OO Constant ----

9 NOTR-CY 1.000E+02 Normal 1.000E+01 9.00E+02 (CCNPP) 9.00E+01 (CCNPP) 10 DR PM-TR**** Normal 1.200E+03 (W) 1.200E+02 (W) 11 RATE-FCG 9.950E-11 Log-Normal 1.414E+OO 12 KEXP-FCG 3.070E+OO Constant -----

13 RPM-OLE 1.500E+03** Normal 1.500E+02**

7.0E+01 (CCNPP) 14 TEMP-F*** Normal 1.250E+01 9.500E+01 (W) 15 RT-NOT 3.000E+01 Normal 1.700E+01 16 F-KIC 1.000E+OO Normal 1.000E-01 17 DLENGTH Per Flywheel Group Constant -----

The uncertainty is a normal standard deviation, the range (median to maximum) for uniform distributions or the corresponding factor for logarithmic distributions.

- RPM-OLE is modified in each case to allow for risk analysis of various plant conditions and their associated flywheel speeds for both Westinghouse Plants and Calvert Cliffs Units 1 and 2. The values used for this variable are discussed in Table 3-3.

- - TEMP-F is set to 95°F for Westinghouse Plants (W) and 70°F for Calvert Cliffs Units 1 and 2 (CCNPP).

- - - DRPM-TR is set to 1200 RPM for Westinghouse plants (W) and 900 RPM for Calvert Cliffs Units 1 and 2 (CCNPP).

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WESTINGHOUSE NON-PROPRIETARY CLASS 3. 3-13 Group specific input variables used in the probability of failure calculations are summarized below:

Flywheel Group ORadius (inch) !Radius (inch) Dlength (inch)

Group1 38.25 4.6875 0.937 Group 2 38.875 4.1875 0.906 Calvert Cliffs 41.00 4.719 0.937 Evaluations were performed to determine the effect on the probability of flywheel failure for continuing the previously approved current inservice inspections in accordance with Reference [2] over the life of the plant through 80 years of , operation and for discontinuing the inspections. The ~valuation also calculated the effects of the inspections being discontinued after ten years. This calculation bounds the effects of any subsequent inservice inspections at 10- to 20-year intervals.

The probability of failure determined by these evaluations is a conservatively calculated parameter because the evaluation conservatively assumes that the probability of a flaw existing after the preservice inspection is 10%, and that the ISi flaw detection probability is only 50%. In reality, most preservice inspection and ISi flaws would be detected, especially for the larger flaw depths which could result in failure. Therefore, the calculated values are very conservative. (The effects of some important parameters on the calculated probability of failure are discussed later in this section). The most important result of the evaluation is the change in calculated probability of failure from continuing versus discontinuing the ISi after 10 years of plant life.

As shown in Figure 3-4, Figure 3-5 and Figure 3-6 and Table 3-3, the ISi provides a negligible benefit for minimizing the potential of failure of the flywheel, The results of this assessment are summarized as follows for a plant life of 40, 60, and 80 years.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-14 Table 3-3: Cumulative Probability of Failure over 40, 60 and 80 Years with and without lnservice Inspection 111 Cumulative Probability of Cumulative Probability of

% Increase in Cumulative Design Flywheel Flywheel Failure with ISi at 4-Failure Probability for Limiting Failure Year Intervals Prior to 10 Years Flywheel Eliminating Inspections Speed with ISi at and without ISi after 10 Years Group (rpm) 4-Year Intervals Over Over 40, 60 Over40 Over60 Over 80 Over 60 Over 80 40

& 80 Years Years Years Years Years Years Years 1 1500 2.39E-7 2.44E-7 2.49E-7 2.52E-7 2.2% 4.5% 5.4%

2 1500 1.29E-7 1.32E-7 1.34E-7 1.35E-7 2.4% 3.5% 4.5%

1 3321 9.87E-3 9.89E-3 9.89E-3 9.99E-3 0.1% 0.2% 1.2%

2 3321 9.01E-3 9.09E-3 9.09E-3 9.09E-3 0.1% 0.1% 0.1%

Calvert 1125 1.50E-8 1.51 E-8 1.51 E-8 1.52E-8 0.5% 0.6% 0.8%

Cliffs Calvert 1368 5.56E-6 5.56E-6 5.62E-6 5.63E-6 0.3% 1.4% 1.5%

Cliffs Note: (1) Results are slightly different from Table 3-8 in [2]. Differences are on the order of 2% and do not impact any conclusions in this analysis or in [2].

As can be seen in Table 3-3, continuing inspection after 10 years has a very minimal impact on the failure probabilities.

Note that for the limiting speed of 3321 rpm, the flywheel failure probability is -1.0E-2 from the first through the 80th year of operation for Flywheel Group 1. It is approximately constant at 9.09E-3 from the first through the 80th year of operation for Flywheel Group 2. The post LOCA limiting speed of Calvert Cliffs Units 1 and 2 of 1368 RPM resulted in reduced failure probabilities, with the equivalent post-LOCA failure probability for a Westinghouse RCP (with a peak speed of 3321 rpm). The Calvert Cliffs cumulative flywheel failure probability for the limiting DEGB LOCA overspeed event was only -5.6E-6.

3.3.2 Sensitivity Study A sensitivity study was performed to determine the effect of select flywheel risk assessment parameters on the probability of failure, as done in [2]. Consistent with [2],

sensitivity studies were performed on a Westinghouse Group 10 flywheel, as this flywheel is representative of average Westinghouse and Byron Jackson flywheel dimensions and configuration. The intent of the sensitivity studies was to illustrate the impact of relatively significant changes to model input parameters on probability of failure predictions. The specific parameters evaluated in this sensitivity study were the PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-15 probability of detection and the initial flaw length. The results of this study are summarized in the Table 3-4 and sections 3.3.2.1 and 3.3.2.2.

Table 3-4: Effect of Flywheel Risk Parameter on Failure Probability (Flywheel Group 10)

Description of Flywheel Risk Probability of Flywheel Probability of Flywheel Parameter Varied Failure after 40 years Failure after 40 years with with ISi prior to and ISi prior to 10 years and after 10 years without ISi after 10 years Base Case (Group 10 of[1]) 1.00E-07 1.04E-07 Probability of Detection of 10% 1.02E-07 1.04E-07 Probability of Detection of 80% 1.00E-07 1.04E-07 Initial flaw length of 0.05 inches 4.57E-08 4.71E-08 Initial flaw length of 0.20 inches 2.97E-07 3.00E-07 The values for the base case were for:

10% probability of a flaw existing after preservice inspection

an initial flaw length of 0.10 inch (1.006 inch with keyway)

an initial ISi at 3 years of plant life, and subsequent inspections at 4-year intervals

probability of detection of 50% per ISi (see [1 ], Table 5-5, flywheel Group 10)

A discussion of the results of the sensitivity studies are summarized below.

3.3.2.1 Sensitivity to Change in Flaw Detection Probability The flaw detection probability was varied from the base case 50% to 10% and 80%. The failure probability increased approximately 3% for a decrease in flaw detection probability from 50% to 10%. The failure probability did not change for an increase in flaw detection probability from 50% to 80%. Therefore, the flaw detection probability, which is a measure of how well the inspections are performed, has essentially no effect on the flywheel failure probability.

3.3.2.2 Sensitivity to Initial Flaw Length The initial flaw length was varied from the base case value of 0.10 inch to 0.05 inch, and 0.20 inch. The failure probability decreased by 54% for a decrease in initial flaw length from 0.10 inch to 0.05 inch, and the failure probability tripled for an increase in initial flaw length from 0.10 inch to 0.20 inch. Therefore, the initial flaw length does affect the flywheel failure probability, but the failure probability is small, even for larger initial flaw lengths. Moreover, it is expected that the probability of the larger flaw being missed during preservice inspection is smaller than the assumed 10% based on reviews of pre-service inspection records in [2].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-16 3.3.3 Failure Probability Assessment Conclusions An evaluation of flywheel structural reliability was performed for each of the flywheel groups selected for evaluation following the process outlined in WCAP-15666-A. Using conservative input values for; preservice flaw existence, initial flaw length, inservice flaw detection capability and RCP start/stop transients, it was shown that flywheel inspections beyond ten years of plant life have no significant benefit relative to the probability of flywheel failure. The reasons are that most flaws that could lead to failure would be detected during the preservice inspection or early in the plant life, and the crack growth is negligible over the plant life. It should be noted that the effect on potential flywheel failure from damage through disassembly and reassembly for inspection has not been evaluated. This is because the purpose of the assessment is to support an inspection interval extension, which will reduce unn~cessary occurrences for introducing potential damage.

Sensitivity studies showed that improved flaw detection capability and more inspections result in a small relative change in the calculated failure probability. The failure probability is most affected by the initial flaw length and its uncertainty. These parameters are determined by the accuracy of the preservice inspection. The uncertainty could be reduced using the results from the first inservice inspection, but would probably not change much during subsequent inspections.

The failure probability estimates identified in [2] show that inspections after 10 years have a very minimal impact on the failure probabilities. These results bound the effects of any subsequent ISi at 10 to 20 year intervals. No credit has been taken for other indications of potential degradation such as pump vibration monitoring and pump maintenance PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-17

_,..., STEA.CY-STATE

~

REAO IN INITIALIZE -....

UNCERTAINTIES PARAMETERS V

CHANGES

/1

'\'I TRANSIENT CHANGES V

CHe:CK 1F" NO NEXT FAILURE ,C TIME OCCUF!S? STEP YES YES

'\] u NO EFFECTS OF PRlNT OUT NEXT CALCULATE PROBABILITY

.......... RANDOM <J, FAILURE ISi OR TRIAL? PR0BA61Ll1Y MONITORING WITH TIME Figure 3-3: Westinghouse PROF Program Flow Chart for Calculating Failure Probability PWROG-17011-NP May 2018 Revision 1

l WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-18 Probability of Failure:

Westinghouse Group 1 Flywheel Probability of Failure 2.60E-07 1 -------:---~&M--~~~~

~Group 1: w/o ISi

,....._Group 1: w/ ISi 0 20 40 60 80 Year Figure 3-4: Probability of Failure for Flywheel Evaluation Group 1 PWROG-17011-NP rviay 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-19 Probability of Failure:

Westinghouse Group 2 Flywheel 2.00E-07 ~ - - - - - - - - - - - - - - -

t----

I 1.80E07 1

1.60E-07 1- - - - -

1 Probability of Failure 1.40E-07 r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..

....m,-Group 2: w/ ISi UOE-07 l* **********-* *---** -****** *****-

1.00E 07 ..1............................................. ,.............................................,.........................................,.........................................1 0 20 40 60 80 Vear Figure 3-5: Probability of Failure for Flywheel Evaluation Group 2 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-20 Probability of Failure:

Calvert Cliffs Flywheel 1.65E . - - - - - - - - - - - - - - -

1.60E-08 + - - - - - - - - - - - - - -

i 1.SSE-08 **;***************************************************************************************************************************************************************************

I -

Proba~ility of 1.SOE-OS II II * .* gL::JJ££.Jt;;;!lll Failure I =+=, cc w/ ISi 1.45E-08 I .....,.. cc w/o ISi 1.40E-08 I 1.35E-08 **,******************************************,*****************************************,***************************************--r****************************************,

0 20 40 60 80 Year Figure 3-6: Probability of Failure for Calvert Cliffs Units 1 and 2 PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-21 3.4 CORE DAMAGE EVALUATION The objective of the risk assessment is to evaluate the core damage risk from the extension of the examination of the RCP motor flywheel, over an extended 80 year in-service duration, relative to other plant risk contributors through a qualitative and quantitative evaluation.

RG 1.174, Revision 2 [5] provides the basis for this evaluation and also provides the acceptance guidelines to make a change to the current licensing basis.

Risk is defined as the combination of likelihood of an event and severity of consequences of an event. Therefore, the following two questions are addressed:

What is the likelihood of the event?

What are the consequences?

The following sections discuss the likelihood and postulated consequences. The likelihood and consequences are then combined in the risk calculation and the results of the evaluation are presented.

Several different scenarios have been identified for potential RCP motor flywheel failures that are related to its operating speed and potential overspeed under certain conditions.

These scenarios are summarized in Table 3-5.

Table 3-5: Summary of Flywheel Analysis Parameters Westinghouse Calvert Cliffs RCP/Flywheel RCP/Flywheel (rpm) (rpm)

Failure during normal plant operation resulting in a 1500* 1125*

plant trip Failure of the RCP motor flywheel associated with 1500* 1125*

a plant transient or LOCA event with NO loss bf electrical power to the RCP Failure of the RCP motor flywheel associated with 1500* 1125*

a plant transient or LOCA event (up to a three square foot break in the main loop) with loss of electrical power to the RCP Failure of the RCP motor flywheel associated with 3321 1368 a large LOCA (from a greater than 3 ft2 break up to the DEGB of the RC loop piping) coincident with an instantaneous electrical power loss (e.g., loss of offsite power (LOOP) or loss of electrical power to the RCP) and therefore no electrical braking to the RCP

RPM are based on the maximum design speed.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-22 3.4.1 What is the Likelihood of the Event The likelihood is addressed by identifying a plaht transient or LOCA event combined with the postulated failure of the flywheel and estimating the probability/frequency of these events. The likelihood of the flywheel failure is discussed in Section 3.3 and the results are provided in Table 3-3 for the two flywheel evaluation groups that bound the other flywheel groups and for the Calvert Cliffs Units 1 and 2 flywheels. The estimated failure probabilities for the different conditions for the various flywheel types and event combinations are shown in Table 3-6.

Table 3-6: Estimated RCP Motor Flywheel Failure Probabilities Cumulative Probabilities of Flywheel Cumulative Probabilities of Failure over 60 Years* Flywheel Failure over 80 Years*

With ISi at 4-Flywheel Group With ISi at 4-Year Year Intervals and Conditions* Intervals Prior to With ISi at 4-Year With ISi at 4- Prior to 10 10 Years and Intervals Year Intervals Years, and without ISi after without ISi after 10 Years 10 Years Group 1 - 2.39E-07 2.49E-07 2.39E-07 2.52E-07 Normal/Accident*

Group 1 - 9.87E-03 9.89E-03 9.87E-03 9.99E-03***

LOCA/LOOP*

Group 2- 1.29E-07 1.34E-07 1.29E-07 1.35E-07 Normal/Accident*

Group 2- 9.09E-03 9.09E-03 9.09E-03 9.09E-03 LOCA/LOOP*

Calvert Cliffs Units 1.SOE-08 1.51 E-08 1.SOE-08 1.52E-08 1&2-Normal/Accident**

Calvert Cliffs Units 5.54E-06 5.62E-06 5.54E-06 5.63E-06 1&2-LOCA/LOOP**

For the failure probability calculations the mean flywheel speed for normal/accident conditions is 1500 rpm; forLOCA/LOOP is it 3321 rpm.

- For the failure probability calculations the mean flywheel speed for normal/accident conditions is 1125 rpm; for LOCA/LOOP is it 1368 rpm.

- - Selected as bounding value PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-23 3.4.2 What are the Consequences?

The consequence evaluation is performed to identify the potential consequences from the failure of the RCP motor flywheel from an integrity standpoint. The consequences are briefly discussed in Section 3.2.

The consequence evaluation includes both direct effects and indirect effects of a flywheel failure. Direct effects are those effects associated directly with the component being evaluated, such as loss of process fluid flow. Indirect effects are those effects on surrounding equipment that may be impacted by mechanisms such as jet impingement, pipe whip, missiles, and flooding.

The direct consequences are defined as failure of the RCP motor flywheel resulting in a failure of the RCP. If a failure of the RCP occurs, a reactor trip would result.

The potential indirect or spatial effects associated with the postulated flywheel failure are associated with the potential missiles generated from the fragmented portions of the flywheel associated with a significant flywheel crack.

For this evaluation, the conditional core damage probability associated with the failure of the flywheel will be assumed to be 1.0 (no credit for safety system actuation to mitigate the consequences of the failure).

3.4.3 Risk Calculation This methodology is described in detail in WCAP-14572, Revision 1-NP-A, Supplement 1 [9]. For failures that cause only an initiating event, the portion of the PRA model that is impacted is the initiating event and its frequency. The core damage frequency from the failure is calculated by:

CDF =IE* CCDP1E Where:

CDF = Core Damage Frequency from a failure (events per year)

CCDP 1E = Conditional Core Damage Probability for the Initiator IE = Initiating Event Frequency (in events per year)

The initiating event frequency (in events per year) is obtained differently for the different conditions. For the normal operating mode, the initiating event frequency is determined from the RCP motor flywheel failure probability model as described in Section 3.3.

Because the model generates a probability, the probability must be transformed into a failure rate. The cumulative probability at a given time is divided by the number of years to end of operating license. In other words, IE= FP/EOL where:

FP = Failure probability from failure probability model (dimensionless)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-24 EOL = Number of years used in. the failure probability model (80 years used to cover an extended plant life). Between 40 and 80 years, the failure probability is relatively constant. -

For the RCP motor flywheel failure following an overspeed event, the core damage frequency of associated with that event (initiating event with flywheel failure) is defined as:

CDF =(IE* CFP)

CCDP where:

CDF = Core Damage Frequency from a failure (events per year)

CCDP = Conditional Core Damage Probability for the initiator and flywheel failure IE = Initiating Event frequency (in events per year)

CFP = Conditional Failure Probability of the flywheel by initiating event The frequencies of the initiating events for the different conditions were identified as follows:

The initiating event frequency for a plant trip or non-LOCA transient is estimated as 1 event/year (plants on average experience 1 plant trip per year).

The probability of a loss of offsite power or loss of power to the RCP following a plant trip was conservatively established from NUREG/CR-6890 [14] as 0.01. This value was based on the observation that the conditional LOOP probability had increased from 0.003 in the 1986-1996 time frame to 0.0053 based on 1997 to 2004 data. Furthermore, the authors noted that the conditional probability in the summer months increase to 0.0091. The LOOP conditional on a LOCA event was estimated from Table 4.2 of NUREG/CR-6538 [13] as 1.4E-02 for PWR plants. LOCAs < 3 ft2 and other plant transient events were conservatively combined, and the probability of a plant transient, concurrent with a LOOP, was conservatively represented by 0.014 and was used.

The frequency of a large break LOCA events with break areas in excess of 3 ft 2 (-23 inches in diameter) was estimated from NUREG-1829 [12]. Mean failure rates of piping are presented in Table 7.19 of that reference. Using 25 and 40 year failure rates, failure rates provided in that table were linearly extrapolated to 60 years and 80 years and then interpolated to obtain a mean frequency of exceeding 3.0 ft2 , Using this process the LOCA exceedance frequency for break areas > 3 ft2 was estimated to be approximately 3.8E-07 per year. For this analysis, the LOCA IE was assigned a bounding value of 1E-06 per year.

Table 3-7, Table 3-8, and Table 3-9 show the calculations that were used to estimate the frequency of the initiating event combined with the probability of the RCP motor flywheel failure. These calculations are also estimates of the core damage frequency given that the assumption of the CCDP is set to 1.0 (no credit taken for any safety systems).

The resulting calculations show that the change in CDF for flywheel Evaluation Group 1 is 1.33E-08/year/RCP, the change in the CDF for flywheel Evaluation Group 2 is 6.00E-09/year/RCP and the change in the CDF for the Calvert Cliffs flywheel is PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-25 1.30E-10/year/RCP. The RG-1.174 criteria for an acceptable change in risk for GDF are 1E-06/year and for LERF is 1E-07/year. These calculations show the change in risk from extending the inspection interval for the RCP motor flywheel continues to remain less than the acceptance criteria.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-26 Table 3-7: Westinghouse RCP Motor Flywheel Evaluation Group 1 Event with RCP Motor Flywheel Initiating Likelihood of RCP Motor Flywheel Failure (and Core Damage Condition Event Failure (@80 years) Frequency, CCDP = 1.0) (per Frequency year)

With ISi after Without ISi after With IS After Without ISi after (per year) 10 Years 10 Years 10 Years 10 Years

1. Normal Operating Condition N/A 2.39E-07 2.52E-07 2.98E-09 3.15E-09

2. Failure of the RCP motor flywheel associated with a plant with NO loss 2.39E-07 2.52E-07 2.39E-07 2.52E-07 of electrical power to the RCP (1200 rpm peak speed)**

3. Failure of the RCP motor flywheel associated with a plant transient 1.40E-02 2.39E-07 2.52E-07 3.34E-09 3.52E-09 (including LOCA event (up to a 3 ft2 break in the RCS loop piping)) with loss of electrical power to the RCP (1200 rpm peak speed)** 1.0 x (1.4E-02)

4. Failure of the RCP motor flywheel associated with a large LOCA (from 1.40E-08 9.87E-03 9.99E-03 1.38E-10 1.4E-10 a greater than 3 ft' break up to a DEGB of the RCS loop piping) coincident with an instantaneous power loss (e.g., loss of offsite power (LOOP) or loss of electrical power to the RCP) and therefore no electrical braking effects (3321 rpm peak speed)

Totals 2.45E-07 2.588E-07 Change in GDF for one Flywheel (per 1.33E-08 RCP risk)

Change in GDF for 4 RCPs (4 Flywheels) 5.32E-08

- The peak speed is 1200 rpm, however, 1500 rpm is used for the failure probability calculations.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-27 Table 3-8: RCP Motor Flywheel Evaluation Group 2 Event with RCP Motor Flywheel Initiating Likelihood of RCP Motor Flywheel Failure (and Core Damage Condition Event Failure (@80 years) Frequency, CCDP = 1.0) (per Frequency year)

With ISi after Without ISi after With IS After Without ISi after (per year) 10 Years 10 Years 10 Years 10 Years

1. Normal Operating Condition N/A 1.29E-07 1.35E-07 1.61 E-09 1.69E-09

2. Failure of the RCP . motor flywheel associated with a plant with NO loss of 1.29E-07 1.35E-07 1.29E-07 1.35E-07 electrical power to the RCP (1200 rpm peak speed)**

3. Failure of the RCP motor flywheel associated with a plant transient 1.40E-02 1.29E-07 1.35E-07 1.81E-09 1.89E-09 (including LOCA event (up to a 3 ft2 break in the RCS loop piping)) with loss of electrical power to the RCP (1200 rpm peak speed)** 1.0 x (1.4E-02)

4. Failure of the RCP motor flywheel associated with a large LOCA (from a 1.40E-08 9.09E-03 9.09E-03 1.27E-10 1.273E-10 greater than 3 ft' break up to a DEGB of the RCS loop piping) coincident with an instantaneous power loss (e.g., loss of offsite power (LOOP) or loss of electrical power to the RCP) and therefore no electrical braking effects (3321 rpm peak speed)

Totals 1.33E-07 1.39E-07.

Change in CDF for one Flywheel (per RCP 6.00E-09 risk)

Change in CDF for 4 RCPs (4 Flywheels) 2.40E-08

- The peak speed is 1200 rpm, however, 1500 rpm is used for the failure probability calculations.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-28 Table 3-9: Calvert Cliffs Units 1 and 2 RCP Motor Flywheel Evaluation Event with RCP Motor Flywheel Initiating Likelihood of RCP Motor Flywheel Failure (and Core Damage Condition Event Failure (@80 years) Frequency, CCDP = 1.0) (per Frequency year)

With ISi after Without ISi after With IS After Without ISi after (per year) 10 Years 10 Years 10 Years 10 Years

1. Normal Operating Condition N/A 1.50E-08 1.52E-08 1.88E-10 1.90E-10

2. Failure of the RCP motor flywheel associated with a plant with NO loss of 1 1.50E-08 1.52E-08 1.SOE-08 1.52E-08 electrical power to the RCP (900 rpm peak speed)**

3. Failure of the RCP motor flywheel associated with a plant transient 1.40E-02 1.50E-08 1.52E-08 (including LOCA event (up to a 3 ft2 2.11 E-10 2.12E-10 break in the RCS loop piping)) with loss of electrical power to the RCP (900 rpm peak speed)** 1.0 x (1.4E-02)

4. Failure of the RCP motor flywheel associated with a large LOCA (from a 1.40E-08 5.54E-06 5.63E-06 greater than 3 ft' break up to a DEGB of the RCS loop piping) coincident with an instantaneous power loss (e.g., loss 7.76E-14 7.87E-14 of offsite power (LOOP) or loss of electrical power to

the RCP) and therefore no electrical braking effects (3321 rpm peak speed)

Totals 1.54E-08 1.557E-08 Change in CDF for one Flywheel (per RCP 1.30E-10 risk)

Change in CDF for 4 RCPs (4 Flywheels) 5.20E-10

- The peak speed is 900 rpm, however, 1125 rpm is used for the failure probability calculations.

PWROG-17011-NP May 2018 Revision 1 j

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-29 3.5 CONSIDERATION OF UNCERTAINTY This section provides a discussion of uncertainties associated with the core damage risk assessment. The discussion follows the general guidance of NUREG-1855 [15] in that the potential key model assumptions and uncertainties are identified and their impact is evaluated with respect to the current application .

. The baseline risk assessment discussed in Section 3.4 includes several significant conservatisms which are intended to bias the results of the analysis in a conservative direction. Specifically, these assumptions include:

1. All flywheel failure events result in both core damage and a large early release.

This tacitly assumes that the missiles generated by the flywheel will result in both an unrecoverable LOCA and a loss in containment integrity sufficient to support a large release of radionuclides. This is a highly unlikely sequence as events resulting from a reactor trip would have control rods inserted prior to the failure and the potential for flywheel fragments to render all safety injection flow paths unavailable is unlikely. Furthermore, there is virtually no likelihood that flywheel fragments could significantly impact the ability of the containment to perform its function or prevent containment isolation.

2. The flywheel failure probability is based on a bounding selection of rotational flywheel speeds. This assumption is intended to simplify the event grouping while upwardly biasing the flywheel failure probabilities. The flywheel failure probability model used to assess the failure probability has been developed as a realistic model. Details of that model are provided in [2] and a sensitivity study to typical input assumptions is provided in Section 3.2.

3. Non-LOCA plant events that could result in a LOOP were assigned a LOOP probability of 0.014. This value is representative of the conditional LOCA/LOOP failure probability and as previously discussed overstates the LOOP potential for the more likely events.

4. The failure probabilities of flywheels are based on the cumulative failure probability over the lifetime of the flywheel. This is conservative because the failure rates are observed to stabilize during the later years of operation.

While these assumptions are intended to provide a bounding estimate of risk, uncertainty associated with other parameters may be of interest in understanding the potential risks of the risk evaluation. As discussed in Section 3.4, the risk of the.

inspection interval extension to 80 years has three elements: the frequency of the initiating event, the probability of flywheel failure associated with an event, and the conditional probability of core damage associated with the failure. Sensitivity studies were performed to investigate the potential impact in changes to the risk assessment modeling assumptions. The results of these studies are included in Table 3-10. The uncertainty associated with each of these factors is discussed below.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-30 3.5.1 Initiating Event Frequency As discussed in Table 3-7 through Table 3-9 the flywheel failure risks are assigned to four bins: normal operation, plant transient events without a loss of off-site power, plant transient events (non-large LOCA) with a loss of off-site power and large LOCA events with a loss of offsite power. Normal operational events (for example RCP start-ups and shutdowns) are based on the flywheel operating life and a bounding number of start-up and shutdown cycles. This is a low contributor to flywheel failure risks. Transient events are assumed to result in acceleration of the flywheel to design speeds. The risk assessment assumed that the plant will experience one transient event per year. A review of plant operation in the United States between 1988 and 2015 demonstrates that overall plant operation has improved and more typical plant failure probabilities are less than 0.80 per year1 . The impact of the reduction in the plant trip frequency results in a 2.6E-09 per year per RCP reduction in CDF from the baseline value. The conditional LOOP probability contributes to the event frequencies for transient and LOCA events.

Increasing the conditional LOOP probability from 0.014 to 0.05 only increases the incremental CDF by 5E-10 per year per RCP. Finally, the frequency of a large LOCA has a significant uncertainty attached to its mean value. In this study the large LOCA frequency (for breaks greater than 3 ft 2) was increased. an order of magnitude from 1E-06 per year to 1E-05 per year with no observable impact on plant risk.

3.5.2 Conditional Flywheel Failure Probability To simplify analyses flywheel failure probabilities were based on 80 year end of life failure assumption and, with the exception of the large LOCA event, the assumption that the flywheel failure condition occurs at the plant design flywheel speed. For Westinghouse plants this was 1500 rpm. However, many plant transients are expected to result in events with lower flywheel speeds closer to that of nominal operation.

Assuming flywheel failure probabilities associated with 1200 rpm operation, per .RCP core damage frequency would reduce to 1E-11 per year.

3.5.3 Conditional Core Damage/Large Early Release Probability Associated with a Flywheel Failure Event The baseline analysis assumes a conditional core damage probability and a conditional large early release probability of 1.0. As discussed above, this is a limiting assumption and the actual values are expected to be much lower; therefore, the conditional LERP probabilities would be negligible.

1 IN/EXT-16-39534, Initiating Event Rates at U.S. Nuclear Power Plants: 1988-2015, INL, May 2016.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-31 Table 3-1 O: CDF Sensitivity to Variations in PRA evaluation assumptions for RCP Flywheel Failure Risk Assessment for Extending 10-year inspection intervals to 80 years -(Flywheel Group 1)

Incremental Change in CDF (per Year)

Risk Impact of Single Risk impact of Flywheel Fl heel Failure Failure 4 RCP Plant Baseline Change in CDF 1.33E-08 5.32E-08 PWR general and other transient 1.07E-08 4.28E-08 reduction to 0.8 per year Increase the Conditional LOOP 1.38E-08 5.51E-08 probability to 0.05 Increase the LOCA frequency for 2 1.33E-08 5.33E-08 breaks >3 ft to 1 E-05/year Flywheel Failure probability reduced for normal operation and the non-large 1.00E-11 4.01 E-11 LOCA transient based on 1200 rpm 3.5.4 Conclusion Regarding Treatment of Uncertainty The above sensitivity studies confirm that even for a relatively large increase in modeling parameters, the incremental CDF would continue to remain below the 1.0E-06 per year core damage and 1.0E-07 per year LERF criteria in [5] supporting the conclusion that this is a very small risk increase. This report assumes the incremental LERF and incremental CDF are equal. This is an extremely conservative assumption.

3.6 RISK RESULTS AND CONCLUSIONS Given the extremely low failure probabilities for the RCP motor flywheel during normal/accident conditions and the extremely low probability of LOCA/LOOP, and assuming a CCDP of 1.0 (complete failure of the safety systems), the CDF and change in risk would still not exceed the risk criteria in [5] (~CDF<1.0E-6 per year and ~LERF

<1.0E-07 per year).

Even considering the uncertainties associated with this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even when all -

four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

Because of the evaluation results for core damage frequency and the conservative assumption that failure of the RCP motor flywheel results in core damage and a large early release, the calculations were not performed for the LERF. If detailed LERF analyses were performed, it is expected that the relative LERF contribution associated with these events would be significantly less than 20%. Regardless, this assessment assumes the calculated CDF is equal to LERF and that results are less than the LERF acceptance criterion (1 E-07/reactor year).

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-32 The key principles identified in RG-1.174 were also reviewed and the responses based on the evaluation are provided in Table 3-11.

This evaluation, in conjunction with the previous deterministic calculations described throughout the report, concludes that the extension of the RCP motor flywheel examination from 10 to 20 years for RCP flywheels in operation up to 80 years would not be expected to result in a significant increase in risk; therefore, the proposed change is acceptable.

Table 3-11: Evaluation with Respectto Regulatory Guide 1.174 (Key Principles)

Key Principles Evaluation Response Change meets current regulations unless it is No exemption or rule change is requested.

explicitly related to a requested exemption or This TR documents applicability of current ISi rule change inspection intervals through 80 years of operation.

Change is consistent with defense-in-depth The potential for failure of the RCP motor philosophy flywheel is negligible during normal accident conditions, and does not impact any plant structures, svstems or components (SSCs).

Maintain sufficient safetv marqins No safetv analysis marqins are chanqed.

Proposed increases in CDF or risk are small The proposed increase in risk is estimated to and are consistent with the Commission's be negligible.

Safety Goal Policy Statement The RCS leakage exists prior to a LOCA (no core damage consequences are associated with the RCS leakage). No credit taken is taken for RCS leakaqe detection.

Use performance-measurement to monitor the NOE examinations are performed on a 20-change year frequency for up to 80 years.

Other indications of potential degradation of the RCP motor flywheel are available (e.g., pump vibration monitoring, and pump maintenance).

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 CONCLUSIONS The results and conclusions as summarized in WCAP-14535A [1] remain valid and are reiterated below:

1. RCP flywheels are carefully designed and manufactured from excellent quality steel, which has a high fracture toughness.

2. The RCP flywheel overspeed is the critical loading; however, LBB has limited the maximum speed to 1500 rpm. (Note, however, that LBB for LBLOCA was not considered in the risk assessment performed in WCAP-15666-A [2], which does consider the overspeed due to the LBLOCA.)

3. RCP flywheel inspections have been performed for over 20 years, with no service-induced flaws.

4. The RCP flywheel integrity evaluations determined a very high flaw tolerance for the RCP flywheels.

5. Crack growth during service is negligible.

6. The structural reliability studies concluded that eliminating inspections will not change the probability of failure.

7. The inspections result in man-rem exposure and the potential for flywheel damage during assembly and reassembly.

The deterministic results as summarized in WCAP-15666-A [2] remain applicable for 80 years of operation. The risk assessments are updated and presented in Section 3 of this report.

1. The failure probabilities for the RCP motor flywheels are small.

2. The change in risk is less than the Regulatory Guide 1.174 CDF and LERF acceptance criteria.

3. The 20-year ISi frequency for the RCP motor flywheel, approved by the NRC in

[2], remains applicable for 80 years of operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 REFERENCES

1. Westinghouse Report, WCAP-14535A, Rev. 0, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination," November 1996.

2. Westinghouse Report, WCAP-15666-A, Rev. 1, "Extension of Reactor Coolant Pump Motor Flywheel Examination," October 2003.

3. United States Nuclear Regulatory Commission, Office of Standards Development, Regulatory Guide 1.14, Rev. 1, "Reactor Coolant Pump Flywheel Integrity," August 1975.

4. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition with 2008 Addenda.

5. United States Nuclear Regulatory Commission, Regulatory Guide 1.174, Rev. 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011.

6. United States Nuclear Regulatory Commission NUREG-0800, Standard Review Plan 19.0, "Use of Probabilistic Risk Assessment in Plant Specific Risk- Informed Decision Making: General Guidance."

7. F. J. Witt, "Development and Applications of Probabilistic Fracture Mechanics for Critical Nuclear Reactor Components," pages 55-70, Advances in Probabilistic Fracture Mechanics, ASME PVP-Vol. 92, 1984.

8. @RISK, Risk Analysis and Simulation add-In for Lotus 1-2-3, Version 2.01 Users Guide, Palisade Corporation, Newfield, NY, February 6, 1992.

9. Westinghouse Report, WCAP-14572, Supplement 1, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed lnservice Inspection," Revision 1-NP-A, February 1999.

10. United States Nuclear Regulatory Commission NUREG/CR-5864, Theoretical and User's Manual for pc-PRAISE, A Probabilistic Fracture Mechanics Computer Code for Piping Reliability Analysis, July 1992.

11. Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loading: Parts 1 and 2. Electric Power Research Institute, Palo Alto, CA: June 1995. TR-105001.

12. United States Nuclear Regulatory Commission NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process," April 2008.

13. United States Nuclear Regulatory Commission NUREG/CR-6538, "Evaluation of LOCA With Delayed Loop and Loop With Delayed LOCAAccident Scenarios," July 1997.

14. United States Nuclear Regulatory Commission NUREG/CR-6890, "Reevaluation of Station Blackout Risk at Nuclear Power Plants: Analysis of Loss of Offsite Power Events," 1986-2004," December 2005.

15. United States Nuclear Regulatory Commission NUREG-1855, Rev. "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking Final Report," July, 2016.

PWROG-17011-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A: CALVERT CLIFFS UNIT 1 & 2 RCP MOTOR FLYWHEEL EVALUATIONS FOR EXTENSION OF ISi INTERVAL Background and Purpose WCAP-15666-A [2] extended the ISi intervals for Westinghouse RCP motors from 10 to 20 years. Although Calvert Cliffs Units 1 and 2 are Combustion Engineering NSSS plants, they have Westinghouse RCP motors and flywheels; however, the motor operating speeds are different than those evaluated in WCAP-15666-A [2]. A Calvert Cliffs plant-specific deterministic calculation and a probabilistic evaluation were performed using the methodology of [2] to justify 20-year ISi interval for 60 years of plant operation.

The probabilistic evaluations for Calvert Cliffs were updated in Section 3 of this report for 80 years of operation. The purpose of this Appendix is to evaluate and extend the applicability of [2] to 80-year plant operation for Calvert Cliffs Units 1 and 2.

Ductile Failure Analysis As discussed in Section 2.3.2 of this report, the flywheel stresses are dependent on dimensions and rotation speed. Extending the operating period to 80 years does not affect the stress calculation. Therefore, the current ductile failure analysis for 60 years remains valid for 80 years of operation.

The ductile failure limiting speed was determined for the flywheel for two cases. Case 1 considered that no cracks were present but accounted for the reduced cross sectional area resulting from the keyway. Case 2 considered that a 10-inch radial crack existed emanating from the center of the keyway through the full thickness of the flywheel.

The calculated limiting speeds are:

Case 1: 3219 rpm (considering the keyway only, no crack)

Case 2: 2856 rpm (considering the keyway and a 1O" crack)

Given the nominal operating speed of 900 rpm for Calvert Cliffs plants, criterion item f [3]

is satisfied since this is lower than one half of the lowest calculated critical speed of 2856/2 = 1428 rpm, considering both no cracks present and a large crack (10") present.

Given the LOCA over speed of 1368 rpm for the Calvert Cliffs plants, criterion item f [3] is satisfied because it is less than any calculated critical speeds considering both no cracks present and a large crack present.

Non-ductile Failure Analysis As discussed in Section 2.3.3 of this report, extending the operating period to 80 years does not affect the K1 calculations, and the flywheel fracture toughness, Kie would not change due to the 80 year extension. Therefore, the current non-ductile failure analysis for 60 years remains valid for 80 years of operation. As in discussed in Section 2.3.3, Table 2-5, RTNDT values of 0°F, 30°F and 60°F were used to calculate the critical flaw sizes shown in Table A-1.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 l '

Table A-1: Critical Crack Length in Inches and % Through Flywheel RTNDT 0°F 30°F 60°F Critical Crack Length 18.5" 8.8" 3.7"

% Through the Flywheel 52% 25% 10%

Note: The % through the flywheel is calculated as CCL in the table divided by the radial length from the maximum radial keyway location to the flywheel outer radius [CCL/ (41.0" - 4. 7188" - 0.937")].

Fatigue Crack Growth As discussed in Section 2.3.4 of th.is report, extending the operating period to 80 years does not affect the K1 and .6.K1 calculations. The 6000 design cycles of start and shutdown used for the FCG was determined to be bounding for 80 years of operation.

However, the 6000 cycles for 80 years of operation must be confirmed for this TR to be applicable. The FCG of 0.025 inch after 80 years or 6000 cycles is negligible even when assuming a large initial crack length of 3. 7 inches.

Excessive Deformation Analysis As discussed in Section 2.3.5 of this report, the 80-year extension has no impact on the excessive deformation analysis of the flywheel. The current deformation results for 60 years remain applicable to 80 years of operation.

The change in the RCP flywheel bore radius and outer diameter at overspeed condition of 1368 rpm are:

.6.a = the change in the flywheel bore radius at overspeed = 0.003 inch

.6.b = the change in the flywheel outside radius at overspeed = 0.006 inch Since .6. is proportional to al, this represents a 231 % increase [(co 05 ;con)2 = (1368 / 900) 2

2.31 231 %] over the deformation at the normal operating speed.

This increase would not result in any adverse conditions, such as excessive flywheel vibrational stresses that would result in crack propagation since the flywheel assemblies are interference fit to the flywheel shaft and the calculated deformations are small and insignificant. It is noted that the deformation for Calvert Cliffs flywheels is less than the that of Westinghouse flywheels reported in Section 2.3.5 of this report.

Conclusion The current Calvert Cliffs evaluation and results for 60 years are applicable for 80 years of operation. The stress and fracture evaluation results for Calvert Cliffs flywheels are consistent with the flywheels evaluated in [3]. The probabilistic risk evaluation, in conjunction with the deterministic calculations described above, concluded that extension of the RCP motor flywheel ISi from 1O to 20 years for flywheels in service up to 80 years is acceptable.

PWROG-17011-NP May 2018 Revision 1

Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 Attachment 5 PWROG-17031-NP, REVISION 1 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

!ZED WATER C:TOR PWROG-17031-NP Revision 1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Update for Subsequent License Renewal: WCAP-15338-A, "A Review of Cracking Associated with Weld

Deposited Cladding in Operating PWR Plants" sc . .1497 May 2018

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWROG-17031-NP Revision 1 Update for Subsequent License Renewal: WCAP-15338-A, "A Review of Cracking.Associated with Weld Deposited Cladding in Operating PWR Plants" PA-MSC-1497 Gordon Z. Hall*

Structural Design and Analysis - I May 2018 Reviewer:Earnest S. Shen*

Structural Design and Analysis - I Approved: Stephen P. Rigby*, Manager Structural Design and Analysis - I Approved: James P. Molkenthin*, Program Director PWR Owners Group PMO

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii ACKNOWLEDGEMENTS This report was developed and funded by the PWR Owners Group under the leadership of the participating utility representatives of the Materials Committee.

PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv WESTINGHOUSE ELECTRIC COMPANY LLC PROPRIETARY LEGAL NOTICE:

This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

1. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

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PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V PWR Owners Group United States Member Participation* for PA-MSC-1497 Participant Utility Member Plant Site(s) Yes No Ameren Missouri Callaway (W) X American Electric Power D.C. Cook 1 & 2 (W) X Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE) X Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X North Anna 1 & 2 (W) X Dominion VA Surry 1 & 2 (W) X Catawba 1 & 2 (W) X McGuire 1 & 2 (W) X Duke Energy Carolinas Oconee 1 (B&W) X Oconee 2, & 3 (B&W) X Robinson 2 (W) X Duke Energy Progress Shearon Harris (W) X Entergy Palisades Palisades (CE) X Entergy Nuclear Northeast , Indian Point 2 & 3 (W) X Arkansas 1 (B&W) X Entergy Operations South Arkansas 2 (CE) X Waterford 3 (CE) X Braidwood 1 & 2 (W) X Byron 1 & 2 (W) X Exelon Generation Co. LLC

TMI 1 (B&W) X Calvert Cliffs 1 & 2 (CE) X Ginna (W) X Beaver Valley 1 & 2 (W) X FirstEnergy Nuclear Operating Co.

Davis-Besse (B&W) X St. Lucie 1 & 2 (CE) X Turkey Point 3 & 4 (W) X Florida Power & Light \ NextEra Seabrook (W) X Pt. Beach 1 & 2 (W) X PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi PWR Owners Group United States Member Participation* for PA-MSC-1497 Participant Utility Member Plant Site(s) Yes No Luminant Power Comanche Peak 1 & 2 (W) X Omaha Public Power District . Fort Calhoun (CE) X Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X PSEG - Nuclear Salem 1 & 2 (W) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Farley 1 & 2 (W) X Southern Nuclear Operating Co.

Vogtle 1 & 2 (W) X Sequoyah 1 & 2 (W) X Tennessee Valley Authority Watts Bar 1 & 2 (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Xcel Energy Prairie Island 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii PWR Owners Group international Member Participation* for PA-MSC-1497 Participant Utility Member Plant Site(s) Yes No Asco 1 & 2 (W) X Asociaci6n Nuclear Asc6-Vandell6s Vandellos 2 (W) X Axpo AG Beznau 1 & 2 (W) X Centrales Nucleares Almaraz-Trillo Almaraz 1 & 2 (W) X EDF Energy Sizewell B (W) X Dael 1, 2 & 4 (W) X Electrabel Tihange 1 & 3 (W) X Electricite de France 58 Units X Eletronuclear-Eletrobras Angra 1 (W) X Emirates Nuclear Energy Corporation Barakah 1 & 2 X EPZ Borssele X Eskom Koeberg 1 & 2 (W) X Hokkaido Tomari 1, 2 & 3 (MHI) X Japan Atomic Power Company Tsuruga 2 (MHI) X Mihama 3 (W) X Kansai Electric Co., LTD Ohi 1, 2, 3 & 4 (W & MHI) X Takahama 1, 2, 3 & 4 (W & MHI) X Kori 1, 2, 3 & 4 (W) X Hanbit 1 & 2 (W) X Korea Hydro & Nuclear Power Corp.

Hanbit 3, 4, 5 & 6 (CE) X Hanul 3, 4 , 5 & 6 (CE) X Genkai 2, 3 & 4 (MHI) X Kyushu Sendai 1 & 2 (MHI) X Nuklearna Electrarna KRSKO Krsko (W) X Ringhals AB Ringhals 2, 3 & 4 (W) X Shikoku lkata 1, 2 & 3 (MHI) X Taiwan Power Co. Maanshan 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii TABLE OF CONTENTS 1 Background and Introduction ......................................................................................... 1-1 2 Mechanisms of Cracking Associated with Weld Deposited Cladding ............................. 2-1 3 Plant Experience with Defects in and under the Weld-deposited Cladding ................... 3-1 3.1 PWR Service Experience Since 1999 ............................................................................ 3-1 4 Effects of Cladding on Fracture Analysis ....................................................................... .4-1 5 Vessel Integrity Assessment .......................................................................................... 5-1 5.1 Potential for Inservice Exposure of the Vessel Base Metal To Reactor Coolant Water.. 5-1 5.2 Fatigue Usage ................................................................................................................ 5-1 5.3 Acceptance Criteria ........................................................................................................ 5-2 5.3.1 ASME Section XI - IWB-3500 ............................................................................. 5-2

,5.3.2 ASME Section XI - IWB-3600 ............................................................................ 5-2 5.4 Fatigue crack growth ...................................................................................................... 5-3 5.5 Allowable Flaw Size - Normal, Upset & Test Conditions ............................................... 5-8 5.6 Allowable Flaw Size - Emergency & Faulted Conditions ............................................... 5-9 6 Summary and Conclusions ............................................................................................ 6-1 7 References ..................................................................................................................... 7-1 PWROG-17031-NP May 2018 Revision 1

I WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1 Background and Introduction As discussed in WCAP-15338-A [1], underclad cracking was initially detected at the Rotterdam Dockyard Manufacturing (ROM) Company during magnetic particle inspections of a reactor vessel in January 1971. These inspections were performed as part of an investigation initiated by ROM as a result of industry observations reported in December 1970. Subsequent evaluations by Westinghouse in the 1970s concluded that these underclad cracks would not have an impact on the integrity of reactor vessels for a full 40 years of operation. The evaluation was submitted to the Atomic Energy Commission in 1972, and the AEC review concurred. This type of underclad cracking is now commonly referred to as reheat cracking.

In late 1979, underclad cracking in reactor vessels resurfaced in the form of "cold cracking". Supplemental inspections confirmed that such cracking existed in a select group of reactor vessels. Fracture evaluations of the detected flaw indications confirmed their acceptability for a 60 year design life [1].

The purpose of this Topical Report (TR) is to update the 60 year fatigue crack growth analysis in [1] and confirm that the analysis is applicable to subsequent license renewal (SLR), up to 80 years of operation. The fracture toughness values used in Appendix A of

[1] will be confirmed for 80 years of operation. Operating experience that is contained in Sections 2 and 3 of [1] will also be updated.

This TR is applicable to all Westinghouse Nuclear Steam Supply System (NSSS) plants.

Revision 1 of this TR removes unnecessary contents that are duplicates in WCAP-15338-A [1]. All evaluation results and conclusions are unchanged from Revision O of this TR.

PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 Mechanisms of Cracking Associated with Weld Deposited Cladding As discussed in WCAP-15338-A [1] and repeated here, underclad cracking was initially detected in 1970, and has been extensively investigated by Westinghouse and others over the past 30 years. This type of cracking in reactor vessels has also been identified in France and Japan, in addition to the United States.

The cracking has occurred in the low alloy steel base metal heat-affected zone (HAZ) beneath the austenitic stainless steel weld overlay that is deposited to protect the ferritic material from corrosion. Two types of underclad cracking have been identified.

Reheat cracking has occurred as a result of post weld heat treatment of single-layer austenitic stainless steel cladding applied using high-heat-input welding processes on ASME SA-508, Class 2 forgings. The high-heat-input welding processes effecting reheat cracking, based upon tests of both laboratory samples and clad nozzle cutouts, include:

strip clad, six-wire clad and manual inert gas (MIG) cladding processes. Testing also confirmed that reheat cracking did not occur with one-wire and two-wire submerged arc cladding processes. The cracks are often numerous and are located in the base metal region directly beneath the cladding. They are confined to a region approximately 0.125 inch deep and 0.4 inch long.

Cold cracking has occurred in ASME SA-508, Class 3 forgings after deposition of the second and third layers of cladding, where no pre-heating or post-heating was applied during the cladding procedure. The cold cracking was determined to be attributable to residual stresses near the yield strength in the weld metal and base metal interface after cladding deposition, combined with a crack-sensitive microstructure in the HAZ and high levels of diffusible hydrogen in the austenitic stainless steel or lnconel weld metals. The hydrogen diffused into the HAZ and caused cold (hydrogen-induced) cracking as the HAZ cooled. Destructive analyses have demonstrated that these cracks vary in depth from 0.007 inch to 0.295 inch and in length from 0.078 inch to 2.0 inches. Typical cold crack dimensions were 0.078 inch to 0.157 inch in depth, and 0.196 inch to 0.59 inch in length. As with the reheat cracks, these cracks initiate at or near the clad/base metal fusion line and penetrate into the base metal.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 Plant Experience with Defects in and under the Weld-deposited Cladding In Section 3 of WCAP-15338-A [1], the historic operating experiences were discussed in detail. Additional operating experiences since the publication of [1] are discussed in this section.

3.1 PWR Service Experience Since 1999 A review of the recent service experience resulting from degraded cladding was performed and very few new instances were identified. The three cases discussed below are the only known new cases [3] and [4]. Plants cited in WCAP-15338-A [1]

which are still in operation continue to experience no detrimental effects of the missing cladding. Therefore, it has been shown to be acceptable even if underclad cracks become a surface crack exposing the base metal to reactor coolant system (RCS) fluid.

1. Callaway Reactor Vessel Bottom Head Region.

An indication in the cladding region at the bottom of the reactor vessel was identified visually, due to a rust stain that was indicative of exposed low alloy steel. The indication was determined to encompass an area of 1.5 inch x 0.75 inch. The location was characterized as 302.94 degrees from the vessel "O" location, and 384.89 degrees from the flange surface. The plant has operated since 2004 with no issues, as verified by three separate inspections, each of which involved removing the core barrel.

2. Diablo Canyon Unit 1 Reactor Vessel Inlet Nozzle During the 2005 inspection of the Diablo Canyon Unit 1 inlet nozzle inner radius, a visual examination identified an area of approximately 1.025 inch x 0.53 inch of clad scraping (spa II) at 1O degrees from the bottom dead center of the nozzle. This particular region was re-examined visually in 2014, and it was determined that there was no noticeable change in the past 9 years. No degradation was identified, nor was it expected, as the PWR RCS is de-oxygenated by the hydrogen overpressure which is present during operation.

3. Qinshan Reactor Vessel Bottom Head Region Indications were discovered in the bottom head region of Qinshan Phase '1 reactor vessel-when-it was examined in 1"999. As discnssed in *[4], it was unclear *whether the base metal was exposed. Due to the irregularity of the surface in the vicinity of the indication, a replication was made of the area and the shape of the degradation scar was determined by a laser scan. Since the original examination, the region has been examined three times, and no change has been observed.

The evaluation in [4] concluded that Qinshan is safe to operate until 2041 as requested, a total of 50 years (end of design life).

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 Effects of Cladding on Fracture Analysis The effects of cladding on the fracture analysis were discussed in detail in Section 4 of WCAP-15338-A [1]. Experiments were performed and measurements were taken.

Fracture analyses of reactor pressure vessels subjected to thermal shock have included various assumptions regarding the behavior of the cladding and its influence on the fracture resistance of the vessel. The effect of cladding is also important because of its relevance to underclad cracks. For the most part, it was assumed that the welded clad layer, being lower in strength and higher in ductility than the low-alloy pressure vessel steel, would produce no observable effect on the strength or apparent fracture toughness of the pressure vessel. The clad layer is assumed to have a sufficient strength to reduce the stress intensity factor, or crack driving force.

As discussed in Section 4 of [1 ], bend bar tests were conducted to study the effect of cladding on the structural behavior in the operating reactor vessels. The residual stress measurements were discussed in [1] in detail. The residual stress measurement confirmed the bend bar test results. It was concluded in [1] that the effects of cladding will be more important at lower temperatures, where the stresses are higher. At temperatures greater than 180°F (82°C) the cladding has virtually no impact on fracture behavior, and this is the very lower end of the temperature range of plant operation. The effects of the cladding are considered for flaws that penetrate the cladding into the base metal. The actual impact of the cladding residual stress on the fracture evaluation is negligible, even for irradiated materials.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 Vessel Integrity Assessment This section discusses the reactor vessel integrity evaluation and assessment.

5.1 Potential for lnservice Exposure of the Vessel Base Metal To Reactor Coolant Water As discussed in Section 5.1 of WCAP-15338-A [1 ], the occurrence of wastage or wall thinning of the carbon steel vessel base metal requires the breaching of the complete thickness of the cladding so that the base metal is exposed to the RCS environment.

This process consists of two sequential stages:

1. Cracking and separation of a portion of the clad weld metal resulting in the exposure of the base metal to the primary water, and

2. Corrosive attack and wastage of the carbon steel base metal due to its exposure to the RCS water Delamination and separation of the complete clad thickness can occur either by mechanical distress or by micro-cracking induced by metallurgical degradation mechanisms. Examples of mechanical distress are denting and overload (overloads can result in metal plasticity and cracking) cracking caused by mechanical impact loads such as those caused by a loose part. Metallurgical mechanisms include intergranular stress corrosion cracking (IGSCC) and transgranular stress corrosion cracking (TGSCC) mechanisms.

IGSCC of the clad metal can occur if the weld is sensitized (chromium depleted grain boundaries) and is exposed to oxygenated water. TGSCC can occur in the cladding only in the presence of a chloride environment. The typical PWR operating and shut down RCS chemistry contains oxygen and chloride levels that are significantly below the threshold levels required to initiate either IGSCC or TGSCC.

Thus there is no degradation mechanism that can contribute to additional breaching of the clad thickness and result in any exposure of the vessel base metal. Even if the base metal were exposed, the degree of corrosive attack and wastage due to operation is insignificant based on operating experience and analyses based on corrosion tests.

5.2 Fatigue Usage As repo~rted iii WCAP-15338-A [1], the maximum cumu-lative fatigue usage factor for the reactor vessel is 0.04 or less for 60 years of operation. Assuming transient cycles linearly scale from 60 to 80 years, the maximum usage factor would be 0.053. This shows that the likelihood of fatigue cracks initiating during service is very low for 80 years of operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-2 5.3 Acceptance Criteria 5.3.1 ASME Section XI - IWB-3500 The underclad cracks which have been identified over the years are very shallow, with a maximum depth of 0.295 inch (7.5mm). The flaw indications indicative of underclad cracks that have been identified during pre-service and inservice inspections are all within the flaw acceptance standard of the ASME Code Section XI, Paragraph IWB-3500. However, the USNRC RAI [1, Section 8] stated that the ASME Section XI IWB-3600 criteria should be used as evaluation criteria. Westinghouse provided a response to this RAI question and the USNRC accepted the response in a Safety Evaluation Report (SER) issued on September 25, 2002. The accepted response is included in Append ix A of WCAP-15338-A [1].

5.3.2 ASME Section XI - IWB-3600 There are two alternative sets of flaw acceptance criteria for ferritic components, for continued service without repair in paragraph IWB-3600 of ASME Code Section XI.

Either of the criteria below can be used as discussed in Appendix A of WCAP-15338-A

[1].

(1) Acceptance criteria based on flaw size (IWB-3611)

(2) Acceptance criteria based on stress intensity factor, K1 (IWB-3612)

Both criteria are comparable for thick sections, and the acceptance criteria based on the stress intensity factor have been determined by past experience to be less restrictive for thin sections, and for outside surface flaws in many cases. In all cases, the most beneficial criteria have been used in the evaluation discussed below.

5.3.2.1 Criteria Based on Flaw Size The code acceptance criteria stated in IWB-3611 of Section XI for ferritic steel components 4 inches and greater in wall thickness are:

81 < 0.1 ac for normal conditions (including upset and test conditions) and, a1 < 0.5 ai for faulted conditions (including emergency conditions) where, a1 = The maximum size to which the detected flaw is calculated to grow until the next

- inspection. An-80 year period is considered in the calculation *herein*.

ac = The minimum critical flaw size under normal operating conditions.

ai = The minimum critical flaw size for initiation of non-arresting growth under postulated faulted conditions.

5.3.2.2 Criteria Based on Applied Stress Intensity Factors Alternatively, the code acceptance criteria stated in IWB-3612 of Section XI for ferritic steel components criteria based on applied stress intensity factors can be used:

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 K < Kia for normal conditions (including upset and test conditions)

I ..f[o K < Kie for faulted conditions (including emergency conditions)

/ ../2 where, K1 = the maximum applied stress intensity factor for the final flaw size after crack growth.

Kia = Fracture toughness based on crack arrest for the corresponding crack tip temperature.

Kie = Fracture toughness based on fracture initiation for the corresponding crack tip temperature.

5.4 Fatigue crack growth A series of fatigue crack growth (FCG) calculations were performed to provide a prediction of future growth of underclad cracks for service periods up to 60 years in [1].

The 60-year FCG calculation was revised and updated for the 80-year SLR application in this TR.

To complete the fatigue crack growth analysis, the methodology of Section XI of the ASME Code was used with the entire set of design transients applied over an 80 year period. The cycles applicable to 40 years of operation were conservatively multiplied by a factor of 2 to account for 80 years of operation. The analysis assumes a flaw of a specified size and shape, considers each design transient, and calculates the crack growth, adding the crack growth increment to the original flaw size, and then repeating the process until all transient cycles have been accounted for.

The crack growth was conservatively calculated using the ASME Section XI, Appendix A, A-4300, crack growth rate for water environments [2]. This is the most current crack growth rate and is comparable to the rate used in the original analysis in [1], which dates back to the ASME Code in 1979. This crack growth rate is shown in Figure 5-1. Even though the underclad cracks are not exposed to the PWR water environment, the water crack growth rate was used for conservatism.

A series of flaw types were postulated to address the various possible shapes for the underclad cracks. Specifically, the postulated flaw depths ranged from 0.05 inch

- (1.3mm)*to 0.30 inch *(7.6mm); which is beyond th*e 0.295 inch (7.5mm) m-aximum* depth of an underclad cold crack. The shape of the flaws analyzed (flaw depth/flaw length) ranged from 0.01 through 0.5. The results are shown in Table 5-1 through Table 5-3.

The maximum flaw size of 0.4267 inch at the end of 80 years is less than the minimum allowable flaw size of 0.67 inch, presented in Section 5.5.

Therefore, it can be concluded that the crack growth is insignificant for any type of flaw which might exist at the clad/base metal interface and into the base metal for both nozzle bore and vessel shell regions.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-4 Table 5-1: Fatigue Crack Growth Result for Beltline Region, Axial Flaw (Water Environment)

Initial Depth after Depth after Depth after Depth after Flaw Depth 20 years 40 years 60 years 80 years Flaw Shape AR = I/a = 2 0.050 0.0504 0.0504 0.0504 0.0504 0.125 0.1256 0.1263 0.1263 0.1271 0.200 0.2023 0.2038 0.2054 0.2077 0.250 0.2534 0.2573 0.2612 0.2651 0.300 0.3046 0.3092 0.3147 0.3193 Flaw Shape AR = I/a = 6 0.050 0.0504 0.0512 0.0512 0.0519 0.125 0.1302 0.1349 0.1403 0.1465 0.200 0.2108 0.2224 0.2341 0.2472 0.250 0.2643 0.2790 0.2945 0.3116 0.300 0.3178 0.3364 0.3557 0.3767 Continuous Flaw {I/a = 100) 0.050 0.0507 0.0513 0.0520 0.0527 0.125 0.1323 0.1399 0.1481 0.1578 0.200 0.2156 0.2318 0.2495 0.2693 0.250 0.2713 0.2937 0.3187 0.3469 0.300 0.3277 0.3569 0.3895 0.4267 Note: Aspect Ratio I/a = flaw length / flaw depth. Depths are in inches.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-5 Table 5-2: FCG Results for Beltline Region, Circumferential Flaw in Water Initial Depth after Depth after Depth after Depth after Flaw Depth 20 years 40 years 60 years 80 years Flaw Shape AR =I/a =2 0.050 0.0504 0.0504 0.0504 o.b5o4 0.125 0.1250 0.1256 0.1256 0.1,256 0.200 0.2000 0.2007 0.2007 0.2015 0.250 0.2503 0.2511 0.2519 0.2519 0.300 0.3007 0.3015 0.3023 0.30,30 i'

Flaw Shape AR =I/a =6 0.050 0.0504 0.0504 0.0504 0.0504 0.125 0.1263 0.1271 0.1279 0.1287 0.200 0.2031 0.2062 0.2093 0.2124 0.250 0.2550 0.2604 0.2658 0.2720 0.300 0.3077 0.3147 0.3216 0.3294 Continuous Flaw (I/a =100) 0.050 0.0501 0.0502 0.0503 0.0504 0.125 0.1265 0.1278 0.1291 0.1305 0.200 0.2043 0.2083 0.2124 0.2167 0.250 0.2573 0.2646 0.2721 0.2801 0.300 0.3106 0.3208 0.3315 0.3429 Note: Aspect Ratio I/a = flaw length / flaw depth. Depths are in inches.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-6 Table 5-3: FCG Results for Inlet Nozzle to Shell Weld, Axial Flaw in Water Initial Depth after Depth after Depth after Depth after Flaw Depth 20 years 40 years 60 years 80 years Flaw Shape AR =I/a =2 0.050 0.0500 0.0500 0.0500 0.0505 0.125 0.1253 0.1253 0.1253 0.1253 0.200 0.2001 0.2011 0.2011 0.2011 0.250 0.2506 0.2506 0.2517 0.2517 0.300 0.3012 0.3022 0.3022 0.3033 Flaw Shape AR =I/a =6 0.050 0.0505 0.0505 0.0505 0.0505 0.125 0.1264 0.1274 0.1274 0.1285 0.200 0.2032 0.2064 0.2095 0.2127 0.250 0.2559 0.2611 0.2664 0.2717 0.300 0.3085 0.3159 0.3243 0.3327 Continuous Flaw (I/a =100) 0.0500 0.0502 0.0503 0.0505 0.0506 0.1250 0.1271 0.1287 0.1303 0.1321 0.2000 0.2059 0.2111 0.2164 0.2222 0.2500 0.2597 0.2693 0.2796 0.2908 0.3000 0.3141 0.3276 0.3419 0.3576 Note: Aspect Ratio I/a = flaw length / flaw depth. Depths are in inches.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-7 For other R ratios see subparagraph A-4300(b)(2)

,o-7 L----------L----....L.---J..-..&......L.....11...L-..1...ii.....---------1....----.1.---..a........._...._........_""'-'

100 Figure 5-1: Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels Exposed to Water Environment [2, Fig. A-4300-2]

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-8 5.5 Allowable Flaw Size - Normal, Upset & Test Conditions The allowable flaw size for normal, upset and test conditions was calculated and documented in Appendix A of WCAP-15338-A [1], using the criteria in Section 5.3.2.2.

The fracture toughness for ferritic steels has been taken directly from the reference curves of Appendix A, ASME Section XI. In the transition temperature region, these curves can be represented by the following equations:

Kie= 33.2 + 20.734 exp [0.02 (T- RTNDT)]

K1a = 26.8 + 12.445 exp. [0.0145 (T- RTNoT)]

where K1e and Kia are in ksi-vin.

While these equations are the simplified form in the current ASME Section XI, they are mathematically identical to those presented in [1 ]; therefore, there is no impact on the results.

The upper shelf temperature regime requires utilization of a shelf toughness, which is not specified in the ASME Code. A value of 200 ksi-vin was used for upper shelf fracture toughness, as test data shows this to be a conservative value as discussed in WCAP-15338-A [1 ]. As shown in Table 5-4, the limiting transients are in the upper temperature range. Fracture toughness* K1c per ASME Section XI, A-4200 would yield values higher than 200 ksi-vin. Lower temperature transients are protected by the pressure-temperature (P-T) limits per ASME Section XI, Appendix G, assuming a 1/4T flaw, which is much larger than those flaws evaluated in this TR. This remains applicable for extension of plant operations from 60 to 80 years.

The upper shelf toughness of 200 ksi-vin is used to evaluate the normal operating, upset, and test condition transients. Portions of the heatup and cooldown transients that drop to temperatures below the upper shelf region are governed by plant-specific P-T limit curves, which provide adequate margins of safety to prevent brittle fracture concerns of the reactor vessel. Therefore, the allowable flaw size determined in Appendix A of [1]

remains applicable for the 80-year SLR application.

The allowable flaw size results for normal, upset and test conditions are provided in Table A-4.1 of WCAP-15338-A [1] and repeated in Table 5-4. The minimum allowable flaw size is 0.67 inch.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-9 Table 5-4: Allowable Flaw Size Summary for Beltline Region - Normal, Upset & Test Conditions Flaw Shape Governing Transient Allowable Flaw Size inches (a/t)

Aspect Ratio 2: 1 Inadvertent Safety Injection 4.07 (0.525)

Aspect Ratio 6: 1 Reactor Trip with Cooldown and S.I. 1.34 (0.173)

Continuous Flaw Excessive Feedwater Flow 0.67 (0.086)

Note: A wall thickness of 7.75 inches was used.

5.6 Allowable Flaw Size - Emergency & Faulted Conditions The allowable flaw sizes for emergency and faulted conditions were also documented in Section A-5 of WCAP-15338-A [1] and shown in Table 5-5.

Table 5-5: Allowable Axial Flaw Sizes for Beltline Region - Emergency and Faulted Conditions Allowable Flaw Size Flaw Shape Depth (inches) Through-wall Ratio (a/t)

Aspect Ratio 2: 1 3.88 0.501 Aspect Ratio 6: 1 1.70 0.219 Continuous Flaw 1.25 0.162 Note: A wall thickness of 7.75 inches was used.

As discussed in Section A-1 of WCAP-15338-A [1], the emergency and faulted conditions are ultimately governed by plant-specific treatment of pressurized thermal shock (PTS). The PTS events are covered through each plant's compliance with the screening criteria of 10CFR50.61. This screening criteria is independent of the plant operating period (whether 60 or 80 years).

The assumed upper shelf value of 200 ksi-Vin was used to determine the allowables, and the temperatures of the emergency and faulted transients considered correspond to the upper shelf for the material. The RT Nor is not expected to change significantly from 60 to 80 years as the rate of material embrittlement from sustained exposure decreases at higher fluence levels, and it does not impact the evaluations summarized herein since the normal operating, upset, and test condition transients result in the most limiting allowable flaw size (0.67 inch) using a conservative upper shelf toughness of 200 ksi-vin.

There are also several conservatisms included in the analysis. Underclad cracks are assumed to be surface cracks, which results in a conservative K1* Conservatively assuming the flaw is exposed to water, the crack growth rate for a water environment is PWROG-17031-NP May 2018 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-10 used. This results in a higher growth rate than assuming an air environment. The full flaw depth is assumed to be in the base material, and linear elastic fracture mechanics is used to determine the allowable flaw sizes. Therefore, the calculation of allowable flaw size for 60 years in [1] remains applicable for 80 years. Note that the largest flaw size of 0.4267 inch at the end of 80 years shown in Table 5-1 is less than the minimum allowable flaw size of 0.67 inch Table 5-4.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6 Summary and Conclusions The purpose of this report is to update the 60 year FCG analysis in WCAP-15338-A [1]

and confirm that the rest of the evaluation in [1] remains applicable to 80 years of operation.

As summarized in [1], there are many levels of defense in depth relative to the underclad cracks. There is no known mechanism for the creation of additional flaws in this region; therefore, the only potential concern is the potential propagation of the existing flaws.

Flaw indications indicative of underclad cracks have been evaluated in accordance with the acceptance criteria in the ASME Code,Section XI. These indications have been identified during pre-service and inservice inspections in those plants that were considered to have cladding conditions which have the potential for underclad cracking.

These flaw indications were dispositioned as being acceptable for further service without repair or detailed evaluation, because they meet the conservative requirements of the ASME Code Section XI, Paragraph IWB-3500. Fracture evaluations have also been performed to evaluate underclad cracks, and the results also concluded that the flaws are acceptable.

A number of previous operation experience summarized in [1] involved cladding cracks, as well as exposure of the base metal due to cladding removal. These cladding cracks were postulated to extend into the base metal in the analysis. In these cases the cracks were postulated to be exposed to the water environment, and successive monitoring inspections were performed. No changes of the indications were identified due to propagation or further deterioration of any type. Based on these observations, these inspections were terminated.

Finally, underclad cracks identified during pre-service .and inservice inspections have been evaluated in accordance with the acceptance criteria in the ASME Code,Section XI. The observed underclad cracks are very shallow, confined in depth to less than 0.295 inch and have lengths up to 2.0 inches. The FCG assessment for these small cracks concluded that there would be very little growth for 80 years of operation, even if they were exposed to the RCS water and with a crack tip pressure of 2,500 psi. For the worst case scenario, a 0.30-inch deep continuous axial flaw in the beltline region would grow to 0.43 inch after 80 years. The minimum allowable axial flaw size for normal, upset, and test conditions is 0.67 inch and for emergency and faulted conditions is 1.25 inches. Since the maximum flaw depth of 0.4267 inch after 80 years of FCG is less than the minimum allowable flaw size of 0.67 inch, underclad cracks of any shape are acceptable for 80 years, regardless of the size or orientation of the flaws. Therefore, it can be concluded that underclad cracks are acceptable relative to the structural integrity of the reactor vessel for 80 years.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 References

1. Westinghouse Report, WCAP-15338-A, Rev. 0, "A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants," October 2002.

2. ASME Boiler & Pressure Vessel Code Section XI, 2001 Edition through 2003 Addenda.

3. Westinghouse Document, LTR-PSDR-TAM-14-003, Rev. 0, "Reactor Vessel Inlet Nozzle Cladding Damage Assessment for Diablo Canyon Unit 1," February 2014.

4. Westinghouse Report, WCAP-18158-P, Rev. 0, "Qinshan Phase I Reactor Vessel Cladding Wear Evaluation for Operating Life Extension up to 50 years," October 2016.

PWROG-17031-NP May 2018 Revision 1

PWROG-17031-NP Revision 1 Proprietary Class 3

- This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Author Approval Hall Gordon Z May-23-2018 14:54:02 Reviewer Approval Shen Earnest S May-23-2018 15:07:09 Manager Approval Rigby Stephen May-23-2018 15:53:25 Manager Approval Molkenthin James May-23-2018 15:56:52 Files approved on May-23-2018

Serial No.: 18-340 Docket Nos.: 50-280/281 Enclosure 4 Attachment 6 PWROG-17090-NP, REVISION 0 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

i I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWROG-17090-NP Revision 0 Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination PA-MSC-1367, Task 3 D. Brett Lynch*

Structural Design and Analysis 111 March 2018 Reviewer: Benjamin E. Mays*

Structural Design and Analysis Ill Approved: Lynn Patterson*, Manager Structural Design and Analysis Ill Approved: James P. Molkenthin*, Program Director PWR Owners Group PMO

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii ACKNOWLEDGEMENTS This report was developed and funded by the PWR Owners Group under the leadership of the participating utility representatives of the Materials Committee.

PWRO~ 17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

1. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

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COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. As a member of the PWR Owners Group, you are permitted to copy and redistribute all or portions of the report within your organization; however all copies made by you must include the copyright notice in all instances.

DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non-Proprietary reports to third parties that are supporting implementation at their plant, or for submittals to the USN RC.

PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V PWR Owners Group United States Member i:>articipation* for PA-MSC-1367, Task 3 Participant Utility Member Plant Site(s)

Yes No Ameren Missouri Callaway (W) X American Electric Power D.C. Cook 1 & 2 (W) X Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE) X Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X North Anna 1 & 2 (W) X Dominion VA Surry 1 & 2 (W) X Catawba 1 & 2 (W) X Duke Energy Carolinas McGuire 1 & 2 (W) X Oconee 1, 2, & 3 (B&W) X Robinson 2 (W) X Duke Energy Progress Shearon Harris (W) X Entergy Palisades Palisades (CE) X Entergy Nuclear Northeast Indian Point 2 & 3 (W) X Arkansas 1 (B&W) X Entergy Operations South Arkansas 2 (CE) X Waterford 3 (CE) X Braidwood 1 & 2 (W) X Byron 1 & 2 (W) X Exelon Generation Co. LLC TMl 1 (B&W) X

Calvert Cliffs 1 & 2 (CE) X Ginna (W) X Beaver Valley 1 & 2 (W) X FirstEnergy Nuclear Operating Co.

Davis-Besse (B&W) X St. Lucie 1 & 2 (CE) X Turkey Point 3 & 4 (W) X Florida Power & Light \ NextEra Seabrook (W) X Pt. Beach 1 & 2 (W) X Luminant Power Comanche Peak 1 & 2 (W) X Omaha Public Power District Fort Calhoun (CE) X Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X PSEG - Nuclear Salem 1 & 2 (W) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Farley 1 & 2 (W) X Southern Nuclear Operating Co.

Vogtle 1 & 2 (W) X Sequoyah 1 & 2 (W) X Tennessee Valley Authority Watts Bar 1 & 2 (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Xcel Energy Prairie Island 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWR Owners Group

. f10n* f or PA -MSC -1367 Tas k 3 nternat1ona I M em ber Pa rt*1c1pa

' Participant Utility Member Plant Site(s) Yes No Asco 1 &2 (W) X Asociaci6n Nuclear Asc6-Vandellos Vandellos 2 (W) X Axpo AG Beznau 1 & 2 (W) X Centrales Nucleares Almaraz-Trillo Almaraz 1 & 2 (W) X EDF Energy Sizewell B (W) X Dael 1, 2 & 4 (W) X Electrabel Tlhange 1 & 3 (W) X Electricite de France 58 Units X Eletronuclear-Eletrobras Angra 1 (W) X Emirates Nuclear Energy Corporation Barakah 1 &2 X EPZ Borssele X Eskom Koeberg 1 & 2 (W) X Hokkaido Tomari 1, 2 & 3 (MHI) X Japan Atomic Power Company Tsuruga 2 (MHI) X Mihama 3 (W) X Kansai Electric Co., LTD Ohi 1, 2, 3&4 (W& MHI) X Takahama 1, 2, 3 & 4 (W & MHI) X Kori 1, 2, 3 & 4 (W) X Hanbit 1 & 2 (W) X Korea Hydro & Nuclear Power Corp.

Hanbit 3, 4, 5 & 6 (CE) X Hanul 3, 4, 5 &6 (CE) X Genkai 2, 3 & 4 (MHI) X Kyushu Sendai 1 & 2 (MHI) X Nuklearna Electrarna KRSKO Krsko (W) X Ringhals AB Ringhals 2, 3 & 4 (W) X Shikoku lkata 1, 2 & 3 (MHI) X Taiwan Power Co. Maanshan 1 & 2 (W) X

Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................................................................................ 1 2.0

SUMMARY

OF RES ULTS ............................................................................................................. 1 3.0 METHODOLOGY ........................................................................................................................... 2 3.1 Evaluation of Rotterdam Forging Material Upper-Shelf Energy .......................................... .4 3.2 Evaluation of Rotterdam Weld Materials Upper-Shelf Energy and Chemistry .................... 5 4.0 GENERIC ROTTERDAM FORGING UPPER-SHELF ENERGY ................................................. 6 4.1 Rheinstahl Huttenwerke AG ................................................................................................. 7 4.2 Fried-Krupp Huttenwerke AG Forgings .............................................................................. 10 4.3 Rotterdam Forgings from Other or Unknown Suppliers ..................................................... 14 4.4 Nozzle Upper-Shelf Energy Value Applicability ................................................................. 15 5.0 GENERIC ROTTERDAM WELD UPPER-SHELF ENERGY AND CHEMISTRY ....................... 17 5.1 Submerged Arc Weld (SAW) .............................................................................................. 17 5.2 Shielded Metal Arc Weld (SMAW) ...................................................................................... 19 5.3 Weld Analysis Summary .................................................................................. :.................. 22

6.0 REFERENCES

............................................................................................................................. 23 Appendix A Supplemental Charpy Impact Energy Data for E8015-G Electrode Welds ....................A-1 PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii List of Tables Table 1 Modern Charpy V-Notch Test Specimen Orientation Terminology .............................. 3 Table 2 Summary of Rotterdam Forging Suppliers ................................................................... 6 Table 3 Summary of Rheinstahl Huttenwerke AG Forgings USE Data .................................... 8 Table 4 Statistical Analysis of Rheinstahl Huttenwerke AG Forgings ..................................... 10 Table 5 Summary of Fried-Krupp Huttenwerke AG Forgings USE Data ................................. 11 Table 6 Statistical Analysis of Fried-Krupp Huttenwerke AG Forgings ................................... 13 Table 7 Summary of Rotterdam Forgings from Other or Unknown Suppliers USE Data ....... 14 Table 8 Statistical Analysis Comparing All Rotterdam Forgings to the Rotterdam Nozzle Forgings ...................................................................................................................... 16 Table 9 Available Rotterdam SAW USE Data ......................................................................... 18 Table 10 Available Rotterdam SAW Chemistry Data ................................................................ 18 Table 11 Statistical Analysis of SAW Welds .............................................................................. 19 Table 12 Available Rotterdam SMAW USE and Chemistry Data ............................................. 20 Table 13 Statistical Analysis of SMAW Weld Nickel Weight Percent ........................................ 22 Table A-1 Supplemental Charpy Impact Energy Data for E8015-G Electrode Welds .............. A-2 List of Figures Figure 1 Comparison of "Weak" Direction and "Strong" Direction Test Specimens .................... 3 PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix List of Acronyms ADAMS Agencywide Documents Access and Management System ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials (ASTM International)

BTP Branch Technical Position CFR Code of Federal Regulations CMTR(s) Certified Material Test Report(s)

Cu Copper CVN Charpy V-notch LT Longitudinal Ni Nickel NRC Nuclear Regulatory Commission PMO Project Management Office PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group RIS Regulatory Issue Summary SAW Submerged Arc Weld SMAW Shielded Metal Arc Weld TL Transverse U.S. United States USE Upper-Shelf Energy cr Standard Deviation PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1

1.0 INTRODUCTION

Licensees have been addressing the embrittlement of additional reactor vessel components and welds not previously within the scope of 10 CFR 50, Appendix G [Ref. 1] due to the effects of aging. 10 CFR 50, Appendix G requires that, in the transverse direction, reactor vessel beltline base metal and weld material Upper-Shelf Energy (USE) values be greater than or equal to 75 ft-lb initially, and remain greater than or equal to 50 ft-lb (68 J) throughout the lifetime of the reactor vessel. In Regulatory Issue Summary (RIS) 2014-11 [Ref. 2], the Nuclear Regulatory Commission (NRC) identified a threshold of 1 x 10 17 n/cm2 (E > 1.0 Me\/) for the projected end of life fluence over which the effects of embrittlement must be considered to meet the 1 O CFR 50, Appendix G requirement. Extended operating durations associated with license renewals can increase the fluence for components outside the traditional beltline beyond this threshold. It should be noted that prior to exceeding 1 x 10 17 n/cm2 (E > 1.0 Me\/), the 10 CFR 50, Appendix G requirements do not apply. The materials outside the reactor vessel beltline and extended reactor vessel beltline, i.e. below the 1 x 1017 n/cm2 (E > 1.0 Me\/) embrittlement threshold during the plant life, were required to meet the American Society of Mechanical Engineers (ASME) Code edition in use at the time of fabrication.

When addressing these additional components for reactor vessels fabricated by the Rotterdam Dockyard Company, some licensees have found it difficult to identify the material information required to establish the initial USE values in accordance with American Society for Testing and Materials (ASTM) E185-82 [Ref. 3], as required by 10 CFR 50, Appendix G. The difficulty in identifying material information stems from significantly less strict testing and reporting requirements at the time of fabrication of the Rotterdam reactor vessels (late 1960's to early 1970's) compared to modern ASME Code requirements.

The objective of this topical report is to provide conservative, generic USE and conservative, generic Copper (Cu) and Nickel (Ni) weight percent values that can be used for Rotterdam reactor vessel welds and forgings when no or limited material information is available. These generic values are developed utilizing data from the surveillance capsule program records and Certified Material Test Reports (CMTRs) available to Westinghouse.

2.0

SUMMARY

OF RESULTS After reviewing all available Charpy data for Rotterdam fabricated reactor vessel welds and forgings the following conservative conclusions have been drawn:

For a forging with insufficient data to determine USE with material supplied by Rheinstahl Huttenwerke AG, a generic lower bound value of 56 ft-lb can be used based on a mean minus 2 standard deviations evaluation. See Section 4.1 for more details.

For a forging with insufficient data to determine USE with material supplied by Fried-Krupp Huttenwerke AG or with an unknown Rotterdam supplier, a generic lower bound value of 52 ft-lb can be used based on a mean minus 2 standard deviations evaluation. See Section 4.2 for more details.

For a Rotterdam Submerged Arc Weld (SAW), the USE can be set to a generic lower bound value of 75 ft-lbs, the Cu weight percent can be set to an upper bound value of 0.23, and the Ni weight percent can be set to an upper bound PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2 value of 0.56. The generic USE value is based on a mean minus 2 standard deviations evaluation. The generic chemistry values are based on a mean plus 1 standard deviation evaluation. See Section 5.1 for more details.

For a Rotterdam Shielded Metal Arc Weld (SMAW), the USE can be set to a lower bound value of 72 ft-lbs, the Cu weight percent can be set to an upper bound value of 0.35, and the Ni weight percent can be set to an upper bound value of 1.13. The Cu value is the generic value from Regulatory Guide 1.99, Revision 2

[Ref. 7]. The Ni value is based on a mean plus 1 standard deviation evaluation.

These values can also be used if the type of Rotterdam weld is unknown. See Section 5.2 and 5.3 for more details.

3.0 METHODOLOGY Herein, generic USE values are determined based on the mean USE of common components minus 2 standard deviations (cr). The mean USE is based on a review of all Charpy impact energy and shear data. When data is available with reported shear greater than or equal to 95%, the USE values are established in accordance with 10 CFR 50, Appendix G [Ref. 1],

which specifies that USE be calculated based on American Society for Testing and Materials (ASTM) E185-82 [Ref. 3]. In this case, USE is calculated based on an interpretation of ASTM E185-82 that is best explained by the most recent version of the ASTM E185 (2016 version).

ASTM E185-16 [Ref. 4], Section 3.1.5, defines the Charpy upper-shelf energy level as the following:

[T]he average energy value for all Charpy specimen tests (preferably three or more) whose test temperature is at or above the Charpy upper-shelf onset; specimens tested at temperatures greater than 83°G (150°F) above the Charpy upper-shelf onset shall not be included, unless no data are available between the onset temperature and onset +83°C (+150°F).

ASTM E185-16 [Ref. 4], Section 3.1.6, defines Charpy upper-shelf onset as the following:

[T]he temperature at which the fracture appearance of all Charpy specimens tested is at or above 95% shear.

Using the above guidelines and in compliance with ASTM E185-82 [Ref. 3], the average of all Charpy data ~ 95% shear is reported as the USE when the shear data is available. In some instances, there may be data that are deemed 'outliers,' which are data points that are uncharacteristically high or low relative to other data at or above 95% shear. These 'outlier' data points are removed from the c::letermination of the USE based on engineering judgment.

When transverse data do not exist, the methodology in NUREG-0800, Branch Technical Position (BTP) 5-3 [Ref. 5] Position 1.2 is used to estimate the USE. The guidance of NUREG-0800, BTP 5-3 [Ref. 5] Position 1.2, states that when estimating Charpy V-notch (CVN) USE:

ff tests were only made on longitudinal specimens, the values should be reduced to 65% of the longitudinal values to estimate the transverse properties.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3 In many cases, the CVN orientation is not reported in CMTRs. For reactor vessel material fabricated before 1973, such as the Rotterdam materials considered herein, the typical industry practice was to perform CVN tests in the strong direction. In instances where the orientation is not reported, the CVN orientation is conservatively assumed to be in the "strong direction".

Note that the terminology for the orientation of specimens used in this report is consistent with the terminology used in the CMTRs. Thus, "longitudinal" and "tangential" are used interchangeably for the "strong direction," and "transverse" and "axial" are used interchangeably to represent the "weak direction." Table 1 provides the most accurate modern terminology for the strong and weak directions, and Figure 1 provides a comparison of strong and weak direction test specimens.

Table 1 Mdo ern Ch arpy V - Notc h T est Sipec1men 0 rientation Terminoloav Reactor Vessel Material Type Weak Direction Strong Direction Plate Transverse (TL) Longitudinal (LT)

Forging Axial Tangential Normal to Major Working Parallel to Major Working General Direction Direction

weAKa PIBECDQH 11 STRONG11 PIRfCJIQN ASME TRANSVERSE ASME lONGITUDINAL ASTMT-t. ASTML*T RPV CIRC. FLAW RPV AXIAL R.AW Figure 1 Comparison of "Weak" Direction and "Strong" Direction Test Specimens In some cases there is not enough information to determine the USE according to ASTM E185-82 in either the weak or strong direction. In these cases, engineering judgment is used to determine how to evaluate and utilize the data. The sections below provide the methodology used to review and evaluate the forgings and welds respectively.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4 3.1 EVALUATION OF ROTTERDAM FORGING MATERIAL UPPER-SHELF ENERGY In order to use the mean USE minus 2 standard deviations (cr) approach, the USE values are determined for all available components. The amount of available CVN data varies depending on the plant and component. In some instances there is enough information to define the USE in the transverse direction using the method of ASTM E185-82. These instances are primarily associated with testing performed to determine the limiting material for inclusion in the reactor vessel surveillance programs. There are other instances where data is only available in the longitudinal direction. In these instances BTP 5-3 [Ref. 5] is utilized to convert the data to transverse data. In other instances, enough data was obtained to develop a full Charpy curve, but shear values were not recorded. Finally, some CVN data does not report the shear and/or stops at approximately room temperature (-70°F). In these cases, insufficient data is available to determine the USE, and only approximations based on the data are available. The following discussion explains how different specific situations are addressed.

If the CVN dataset contains at least one shear data point greater than 95%, but some data points report no shear, all data points with an impact-energy approximately equal to or greater than the impact energy of the shear data points known to be;:: 95% are assumed to have greater than or equal to 95% shear. All non-outlier data points with known or assumed shear at;:: 95% are averaged to determine the USE and incorporated into the calculation of the generic USE.

If the CVN dataset contains limited or no shear data, however, the upper shelf can clearly be determined from the data provided (through visual inspection), the USE is identified and incorporated into the calculation of the generic USE. The USE is identified by an approximately constant energy vs. temperature region.

For example, in some cases, data points at four temperatures over a 50°F range exhibited energy values within a 10 ft-lb scatter or less. The existence of the upper-shelf region is confirmed by plotting the impact energy data and identifying if the plot levels off at higher temperatures. The USE represents an average of all Charpy energy values considered to be in the upper-shelf region.

If the CVN dataset reports shear values, but all data indicates a shear less than 95%, the USE is reported as greater than or equal to the maximum reported CVN impact energy, and is not incorporated into the calculation of the generic USE.

If the CVN dataset included limited shear data or did not include shear data and Charpy impact energies are increasing throughout the temperature range available, it is unknown if the upper shelf has been reached. The USE values are conservatively determined based on the information available and the actual USE values may be higher. The USE is set equal to the highest Charpy impact energy value available or, if the highest data point is determined to be a potential

'outlier' or a non-representative data point, the USE is set equal to a value less than the highest value based on the average of the comparable preceding data points. In these instances, the USE is not incorporated into the calculation of the generic USE.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5 3.2 EVALUATION OF ROTTERDAM WELD MATERIALS UPPER-SHEJ_F ENERGY AND CHEMISTRY Rotterdam-supplied CMTRs which contain data on weld materials used in the fabrication of vessels by Rotterdam do not consistently specify where the materials were used. The CMTRs often contain Charpy impact data at a limited number of temperatures or at a single temperature. The industry practice at the time of fabrication of the Rotterdam reactor vessels (late 1960's to early 1970's) was to test Charpy specimens at 10°F to show 30 ft-lbs or more of absorbed energy, and the test information contained in the CMTRs was considered sufficient to satisfy the fracture toughness requirements of ASME Code at the time. Since this amount of information is not sufficient to determine a USE, and there exist instances where the weld heat is not identified for a specific weld seam, a generic USE is developed herein.

The Rotterdam CMTRs identify two types of welds used in the fabrication of the vessels, shielded metal arc welds (SMAW) and submerged arc welds (SAW). Each weld type is addressed separately.

The generic USE value for the SAWs is the mean minus 2 standard deviations (cr) value of the initial USE of the surveillance welds for all Rotterdam fabricated vessels. Outside of the baseline measurements for the reactor vessel surveillance programs, no USE information is available for Rotterdam SAW materials. As discussed previously, the weld material was typically tested at only one temperature, and insufficient data exists to determine the USE w.ith accuracy. Therefore, only the results relevant to the reactor vessel surveillance capsule program unirradiated testing at Rotterdam and Westinghouse can be used to determine the generic USE for an unirradiated Rotterdam SAW material. The surveillance capsule programs contain weld specimens which represent every Rotterdam SAW heat vendor and every Rotterdam flux type (although not every heat-flux type combination is represented). The core region welds had the same specification requirements as the other reactor vessel welds; however, for the core region welds, Rotterdam was required to "aim for" both a Charpy V-Notch Transition Temperature (T cv) and a Nil-Ductility Transition Temperature (NOTT) less than 10°F and to furnish additional test results relevant to T cv and NOTT. Both the T cv and NOTT do not occur near the upper-shelf region, and thus, the surveillance capsule program test results are generically representative of the SAWs produced at Rotterdam for USE calculations. Note that Linde 80 flux type welds are intentionally excluded from this analysis, as these welds have been analyzed generically previously (e.g., Reference 6), and the use of this flux type is believed to be applicable only to two of the Rotterdam fabricated reactor vessels.

The generic USE value for SMAWs deviates from the mean minus 2 standard deviations approach previously discussed, because the SMAWs Charpy tests typically do not provide enough information to determine a USE. Instead of a true USE value, a lower bound USE value is developed for each component based on the available information. If no shear is available, the lower bound USE is reported as the average of the Charpy impact energies at the test temperature, typically around 10°F. When shear values are reported and each is less than 95%, then the maximum Charpy impact energy value is reported. The generic USE value for SMAWs is then based on analysis of the lower bound USE data arid corresponding shear data, as available. By reporting the USE in this manner, a conservative representation of the USE for SMAWs is provided based on the lowest possible value of USE.

In addition to the generic USE, the generic chemistry, i.e., Cu and Ni weight percentages, is PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6 determined for both SAWs and SMAW based on the mean plus 1cr. This method is based on Regulatory Guide 1.99, Revision 2 [Ref. 7], which states that conservative chemistry estimates are a mean plus one standard deviation. If ~ common heat-flux type combination is shared between multiple welds, the average chemistry value for the heat is considered as one data point when determining the generic weld chemistry values as not to assign undue weight to the material, since it is representative of just one heat-flux type combination. The chemistry data used in the evaluation consists of the measurements from the reactor vessel surveillance program, supplemented with all available chemistry data for heats outside the surveillance programs. The data is limited to deposited weld chemistry results, unless otherwise noted. The chemistry analysis of the bare weld wire is excluded, as the deposition process can affect the chemistry of the weld.

4.0 GENERIC ROTTERDAM FORGING UPPER-SHELF ENERGY This section reviews all Rotterdam reactor vessel forgings, including the supplier responsible for the forging, the USE in the strong direction ("known" and estimated), the estimated USE values in the weak direction determined using BTP 5-3 Position 1.2 [Ref. 5], and the USE values in the weak direction determined from the original CMTR data.

The data herein is evaluated using the guidance of 10 CFR 50.61 [Ref. 8], which states:

Data from reactor vessels fabricated to the same material specification in the same shop as the vessel in question and in the same time period is an example of "generic data."

For the purposes of this evaluation, "same shop" is considered to be the same supplier responsible for the forging. Table 2 breaks down the vessel components according to the responsible supplier. Note that all reactor vessel head materials considered herein are the original plant materials.

Table 2 summary of Rotterdam Forginq Suppliers Number of Number of Materials Number of Materials with Number Supplier with "Known" Components "Known" Strong- Weak-Direction USE(a) of Nozzles Direction USE(a)

Rheinstahl 38 9 10 16 Huttenwerke AG Klockner-Werke AG 8 8 2 0 Fried-Krupp 67 38 5 47 Huttenwerke AG Terni 6 2 0 0 Unknown 2 0 0 1 Marrel-Freres 2 0 0 2 Table 2 note contained on following page.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7 Table 2 Note:

a. "Known" USE values are those which could be positively identified with .:: 95% shear values or visually. These USE values are not marked with a.:: symbol in the following tables in this section. See the following tables and Section 3.1 for more details.

4.1 RHEINSTAHL HUTTENWERKE AG Table 3 contains the forgings procured from Rheinstahl Huttenwerke AG. All forgings were manufactured to the ASTM A508, Class 2 specification and were manufactured in the late 1960s and early 1970s timeframe. Note, for USE values preceded by a">" symbol, the listed USE value is conservatively determined based on the information available, and the actual USE value may be higher.

Table 4 statistically evaluates the USE data for all Rheinstahl Huttenwerke AG supplied forgings. In Table 4, the mean weak-direction USE determined using BTP 5-3 estimates is identical to the USE determined using known weak-direction data. The mean minus 2 standard deviation weak-direction USE value utilizing actual weak-direction data is lower than the corresponding BTP 5-3 value due to a larger standard deviation. A value of 56 ft-lbs, corresponding to the measured weak data mean minus two standard deviations, is conservative for use when USE cannot be determined from available data for a Rheinstahl Huttenwerke AG forging as this value also bounds the BTP 5-3 mean minus two standard deviations value. The results indicate that the generic unirradiated USE in the weak-direction minus two standard deviations is greater than the 10 CFR 50, AppendixG [Ref.1] criterion for a minimum irradiated USE of 50 ft-lbs.

The value of 56 ft-lbs is taken from the "known USE" column. Theses "known" values are identified as the values in Table 3 which do not include a ".::". This value is more appropriate than the values in the "known and estimated" USE column as the estimated data is incomplete and represents'an unnecessary penalty on the generic value. The "known and estimated" USE data is shown for information, and is not recommended for use as it contains a significant amount of data that may not represent the actual USE for Rheinstahl Huttenwerke AG forgings.

The "estimated" USE suppresses the mean and increases the standard deviation as a result of intentional conservatism and incomplete data.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8 Table 3 summary of Rh ems

. t a hi Huttenwerke AG Forgmgs USEDaa t <a>

Uooer-Shelf Enerav (ft-lbs)

Strong Weak Plant Component Direction<bl BTP 5.3<c> Direction Plant A Head Flange 128 83 N/A Head Flange .:: 141 .:: 92 N/A Vessel Flange .:: 163 .:: 106 N/A Upper Shell 143(d) 93 111 (d)

Plant B Intermediate Shell 119(d) 77 75(d)

Lower Shell 116(e) 75 64(e)

Bottom Head Ring .:: 113 .:: 73 N/A Head Flange .:: 142 .:: 92 N/A Vessel Flange .:: 156 .:: 101 N/A Upper Shell .:: 105 .:: 68 N/A Plant C Intermediate Shell 133(e) 86 88(e)

Lower Shell 137(d) 89 98(d>

Bottom Head Ring .:: 113 .:: 73 N/A Upper Shell .:: 87 .:: 57 N/A Intermediate Shell .:: 82 .:: 53 N/A Lower Shell 135(e) 88 77(e)

Inlet Nozzle 09 .:: 72 .:: 47 N/A Plant D Inlet Nozzle 1O .:: 98 .:: 64 N/A Inlet Nozzle 11 .:: 6o<fl .:: 39(f) N/A Outlet Nozzle 12 .:: 89 .:: 58 N/A Outlet Nozzle 13 .:: 98 .:: 64 N/A Outlet Nozzle 14 .:: 6o(fl .:: 39(f) N/A PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9 Table 3 Summary of Rheinstahl Huttenwerke A G F orgmgs USEDaa t (al Upper-Shelf Enerav (ft-lbs)

Strong Weak Plant Component Direction<bl BTP 5_3<cl Direction Upper Shell ~ 68(f) ~44(f) N/A Intermediate Shell ~ 99 ~ 64 91(d)

Lower Shell 135(e) 88 95(e)

Inlet Nozzle 09 ~ 109 ~ 71 N/A Plant E

Inlet Nozzle 10 ~ 89 ~ 58 N/A Inlet Nozzle 11 ~ 80 ~ 52 N/A Outlet Nozzle 12 ~ 101 ~ 66 NIA Outlet Nozzle 13 ~ 90 ~ 59 N/A Outlet Nozzle 14 ~ 90 ~ 59 N/A Upper Shell ~ 87 ~ 57 N/A Intermediate Shell 115(e) 75 72(e)

Lower Shell ~ 112 ~ 73 8o(dJ Plant F Inlet Nozzle 09 ~ 72 ~ 47 N/A Outlet Nozzle 12 ~ 93 ~ 60 N/A Outlet Nozzle 13 ~ 64(f) ~ 42(fl N/A Outlet Nozzle 14 ~ 113 ~ 73 N/A Table 3 Notes:

a. All USE values are determined by averaging available Charpy energy values with reported shear ~

95% (from CMTRs and surveillance program baseline reports, etc., as available) per ASTM E185-82 methods, unless otherwise noted. "NIA" indicates the information is not available.

b. The Charpy data identified with a "~" symbol included limited shear data or did not include shear data, and it is unknown if the upper shelf has been reached, since Charpy impact energies are increasing throughout the temperature range available. For USE values preceded with a"~" symbol the USE is set equal to a value less than or equal to the highest CVN value available. For USE values preceded with a ">" symbol the listed USE value is conservatively determined based on the information available; the actual USE value is likely higher.

c. NRC Branch Technical Position (BTP) 5-3 [Ref. 5] Position 1.2 was utilized to convert strong-direction USE data to weak-direction USE data by reducing the strong-direction energy values to 65% of the reported values.

d. USE determination includes data taken from supplemental Westinghouse test records associated with the surveillance capsule program. *

e. USE is the average of all available Charpy energy values with reported shear~ 95% per ASTM E185-82 methods. Charpy data included data taken from Reactor Vessel Surveillance Programs baseline test reports.

f. This USE value likely does not provide an accurate representation of USE. The actual USE is likely much higher since a Charpy test with a similar absorbed energy has a shear value much less than 95%. Therefore, this data point is excluded from the statistical analysis.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 10 Table 4 Statistical Analvsis of Rheinstahl Huttenwerke AG**Forainas (a)

Known USE Known and Estimated USE Strong Weak Strong Weak Direction BTP 5-3 Direction Direction BTP 5-3 Direction Mean 129 84 84 110 72 84 Standard 10 7 14 24 16 14 Deviation Mean-2a 109 70 56 62 40 56 Maximum 143 93 111 163 106 111 Minimum 115 75 64 72 47 64

of Components 9 10 34 10 Included Table 4 Note:

a. Statistical analysis of Table 3 values, unless the value is noted as excluded. "Estimated" values are identified with a "2:" symbol.

4.2 FRIED-KRUPP HUTTENWERKE AG FORGINGS Table 5 contains the forgings procured from Fried-Krupp Huttenwerke AG. All forgings were manufactured to the ASTM A508, Class 2 specification or the corresponding ASME SA508, Class 2 specification and were manufactured in the late 1960s and early 1970s timeframe.

Note, for USE values preceded by a ">" symbol, the listed USE value is conservatively determined based on the information available, and the actual USE value may be higher.

Table 6 statistically evaluates the USE data for all Fried-Krupp Huttenwerke AG supplied forgings. In Table 6, the mean known weak-direction USE is greater than the weak-direction USE based on BTP 5-3 estimates. The mean minus two standard deviation weak-direction USE value utilizing actual weak-direction data is lower than the corresponding BTP 5-3 value due to a larger standard deviation. A value of 52 ft-lbs, corresponding to the measured weak data mean minus two standard deviations is conservative for use when USE cannot be determined from available data for a Fried-Krupp Huttenwerke AG forging as this value also bounds the BTP 5-3 mean minus two standard deviations value. The results indicate that the generic unirradiated USE in the weak-direction minus two standard deviations is greater than the 10 CFR 50, Appendix G [Ref. 1] criterion for a minimum irradiated USE of 50 ft-lbs.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 11 The value of 52 ft-lbs is taken from the "known USE" column. These "known" values are identified as the values in Table 5 which do not include a "~". This value is more appropriate than the values in the "known and estimated" USE column as the estimated data is incomplete and represents an unnecessary penalty on the generic value. The "known and estimated" USE data is shown for information, and is not recommended for use as it contains a significant amount of data that may not represent the actual USE for Fried-Krupp Huttenwerke AG forgings. The "estimated" USE suppresses the mean and increases the standard deviation as a result of intentional conservatism and incomplete data.

Table 5 summary of F.rie d-Krupp Huttenwerke AG Forgmgs USEData (al Upper-Shelf Energy (ft-lbs)

Strong Weak Plant Component Directionlbl BTP 5_3(cl Direction Intermediate Shell 128<cti 83 62(d)

Lower Shell 136 88 111 (e)

Bottom Head Ring 162 105 NIA Inlet Nozzle 11 ~ 113<fl ~ 73 NIA Inlet Nozzle 12 126 82 NIA Plant A Inlet Nozzle 13 ~ 125<fl ~ 82 NIA Inlet Nozzle 14 137 89 NIA Outlet Nozzle 15 119(h) 77 NIA Outlet Nozzle 16 121<h) 79 NIA Outlet Nozzle 17 141 92 NIA Outlet Nozzle 18 ~ 104<fl ~ 68 NIA Inlet Nozzle 11 ~ 106 ~ 69 NIA Inlet Nozzle 12 ~ 93 ~ 60 NIA Inlet Nozzle 13 ~ 119 ~ 77 NIA Inlet Nozzle 14 ~ 106 ~ 69 NIA Plant B Outlet Nozzle 15 ~ 92 ~ 60 NIA Outlet Nozzle 16 ~ 84 ~ 55 NIA Outlet Nozzle 17 ~ 109 ~ 71 NIA Outlet Nozzle 18 ~ 127 ~ 83 NIA Inlet Nozzle 11 ~ 79 ~ 51 NIA Inlet Nozzle 12 ~ 109 ~ 71 NIA Inlet Nozzle 13 ~ 113 ~ 73 NIA Inlet Nozzle 14 134(h) 87 NIA Plant C Outlet Nozzle 15 ~ 86 ~ 56 NIA Outlet Nozzle 16 ~ 77 ~ 50 NIA Outlet Nozzle 17 ~ 106 ~ 69 NIA Outlet Nozzle 18 ~ 144 ~ 94 NIA Head Flange ~ 173 ~ 112 NIA Plant D Vessel Flange 152<h) 99 NIA Plant E Vessel Flange ~ 166 ~ 108 NIA PWROG-17090-NP March 2018 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 12 Table 5 Summary of F. rie d -K rupp Huttenwerk e AG Forgmas USEDaa t Cal Upper-Shelf Enerav (ft-lbs)

Strong Weak Plant Component Direction BTP 5.3<c> Direction Head Flange .:: 130 .:: 85 N/A Plant F Vessel Flange .:: 146 .:: 95 N/A Head Flange 146 95 N/A Vessel Flange 211 137 N/A Intermediate Shell 148(d)(e) 96 11 o(d)(e)

Lower Shell 156 101 123(e)

Bottom Head Ring 162 105 NIA Inlet Nozzle 11 121 79 N/A Plant G Inlet Nozzle 12 103 67 N/A Inlet Nozzle 13 94 61 N/A Inlet Nozzle 14 133 86 N/A Outlet Nozzle 15 139 90 N/A Outlet Nozzle 16 110 72 N/A Outlet Nozzle 17 129 84 N/A Outlet Nozzle 18 .:: 129(f) .:: 84 N/A Head Flange 158( 9 ) 103 N/A Lower Shell 153 99 N/A Bottom Head Ring .:: 110 .:: 72 N/A Inlet Nozzle 11 133(h) 86 N/A Inlet Nozzle 12 132(h) 86 N/A Plant H Inlet Nozzle 13 125(h) 81 N/A Inlet Nozzle 14 117(h) 76 N/A Outlet Nozzle 15 130(h) 85 N/A Outlet Nozzle 16 137(h) 89 N/A Outlet Nozzle 17 125(h) 81 N/A Outlet Nozzle 18 .:: 111 .:: 72 N/A Head Flange . 156(h) 101 N/A Vessel Flange 173(h) 112 N/A Intermediate Shell 15o(d) 98 94(d)

Bottom Head Ring .:: 110 .:: 72 N/A Inlet Nozzle 11 144( ) 9 94 N/A Plant I Inlet Nozzle 12 .:: 13o<fl .:: 85 N/A Inlet Nozzle 13 134( 9 ) 87 N/A Outlet Nozzle 15 122(h) 79 N/A Outlet Nozzle 16 .:: 104 .:: 68 N/A Outlet Nozzle 17 118( 9 ) 77 N/A Outlet Nozzle 18 .:: 129 .:: 84 N/A Table 5 notes contained on following page.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 13 Table 5 Notes:

a. All USE values are determined by averaging available Charpy energy values with reported shear o::
95% (from CMTRs and surveillance program baseline reports, etc., as available) per ASTM E185-82 methods, unless otherwise noted. "NIA" indicates the information is not available.
b. Unless otherwise noted, the Charpy data identified with a "o::" symbol included limited shear data or did not include shear data, and it is unknown if the upper shelf has been reached, since Charpy impact energies are increasing throughout the temperature range available. For USE values preceded with a "o::" symbol the USE is set equal to a value less than or equal to the highest CVN value available. For USE values preceded with a "o::" symbol the listed USE value is conservatively determined based on the information available; the actual USE value is likely higher.
c. NRC Branch Technical Position (BTP) 5-3 [Ref. 5] Position 1.2 was utilized to convert strong-direction USE data to weak-direction USE data by reducing the strong-direction energy values to 65% of the reported values.
d. USE is the average of all available Charpy energy values with reported shear o:: 95% per ASTM E185-82 methods. Charpy data included data taken from Reactor Vessel Surveillance Programs baseline test reports. *
e. USE determination includes data taken from supplemental Westinghouse test records associated with the surveillance capsule program.
f. All reported shear values are less than 95% shear. The reported value is less than or equal to the maximum energy value of a specimen with less than 95% shear. As a result, the USE is higher than the CVN data reported.
g. USE includes averaged data points without a reported shear but assumed to be o:: 95% shear based on comparison of the CVN data points known to be at o:: 95% shear for the same material.
h. The dataset contained limited or no shear data; however, the USE could clearly be determined from the data provided. For this material, the upper shelf could be identified as a result of the existence of an approximately constant energy vs. temperature region. This USE represents an average of all Charpy energy values considered to be in the upper shelf region.
Table 6 stat1st1ca IAna1ys1s ofF"r1e dK - rupp Huttenwerke A GF orgmgs (a)
Known USE Known and Estimated USE Strong Weak Strong Weak Direction BTP 5-3 Direction Direction BTP 5-3 Direction Mean 137 89 100 128 83 100 Standard Deviation 21 14 24 25 16 24 Mean-2a 95 61 52 78 51 52 Maximum 211 137 123 211 137 123 Minimum 94 61 62 77 50 62
of Components 38 5 67 5 Included Table 6 Note:
a. Statistical analysis of Table 5 values. "Estimated" values are identified with a "o::" symbol.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 14 4.3 ROTTERDAM FORGINGS FROM OTHER OR UNKNOWN SUPPLIERS Table 7 contains the forgings produced from suppliers other than Rheinstahl Huttenwerke AG or Fried-Krupp Huttenwerke AG. All forgings were manufactured to the ASTM A508, Class 2 specification or the corresponding ASME SA508, Class 2 specification and were manufactured in the late 1960s and early 1970s timeframe. As can be seen from the results, a generic USE is not required for any of these components. All components have a USE determined with ASTM E185-82 or the USE is able to be conservatively estimated to be significantly greater than the 10 CFR 50, Appendix G USE criterion for a minimum irradiated USE of 50 ft-lb for operati!lg plants. Based on the data in Table 7 and the generic values determined for Rheinstahl Huttenwerke AG or Fried-Krupp Huttenwerke AG forgings, a Rotterdam forging with an unknown supplier from the late 1960's or early -1970's timeframe will have a USE value of at least 52 ft-lbs, the minimum generic value determined in this report for Rotterdam forgings.
Table 7 summar
0 fR otterdam Forgmgs f rom Oh t er or unknown supp r1ers USEData (al Uooer-Shelf Energ11 (ft-lbs)
Strong BTP Weak Plant Component Supplier Direction!bl 5_3(c) Direction Closure Head Ring Terni 148 96 N/A 155(d) 101 N/A Plant A Vessel Flange Klockner-Werke AG Upper Shell Klockner-Werke AG 152 99 N/A Plant B Top Head Ring Terni .:: 126 .:: 82 N/A Plant C Top Head Ring Terni .:: 126 .:: 82 N/A Plant E Head Flange Unknown .:: 144 .:: 94 N/A Inlet Nozzle 10 Marrel-Freres .:: 118 0:: 77 N/A Plant F Inlet Nozzle 11 Marrel-Freres .:: 115 0:: 75 N/A Closure Head Ring Terni 148 96 N/A Plant G Upper Shell Klockner-Werke AG 144 94 N/A Closure Head Ring Terni .:: 155 .:: 101 N/A Vessel Flange Klockner-Werke AG 235(e) 153 N/A Plant H Upper Shell Klockner-Werke AG 156 101 N/A Intermediate Shell Klockner-Werke AG 161(f) 105 134(f)
Closure Head Ring Terni .:: 155 .:: 101 N/A Plant I Upper Shell Klockner-Werke AG 156 101 N/A Lower Shell Klockner-Werke AG 151 98 141<9)
Table 7 notes contained on following page.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 15 Table 7 Notes:
a. All USE values are determined by averaging available Charpy energy values with reported shear ~
95% (from CMTRs and surveillance program baseline reports, etc., as available) per ASTM E185-82 methods, unless otherwise noted. "NIA" indicates the information is not available.
b. The Charpy data identified with a"~" symbol included limited shear data or did not include shear data, and it is unknown if the upper shelf has been reached, since Charpy impact energies are increasing throughout the temperature range available. For USE values preceded with a"~" symbol the USE is set equal to a value less than or equal to the highest CVN value available. For USE values preceded with a "~" symbol the listed USE value is conservatively determined based on the information available: the actual USE value is likely higher.
c. NRC Branch Technical Position (BTP) 5-3 [Ref. 5] Position 1.2 was utilized to convert strong-direction USE data to weak-direction USE data by reducing the strong-direction energy values to 65% of the reported values.
d. The dataset contained limited or no shear data; however, the USE could clearly be determined from the data provided. For this material, the upper shelf could be identified as a result of the existence of an approximately constant energy vs. temperature region. This USE represents an average of all Charpy energy values considered to be in the upper shelf region.
e. USE includes averaged data points without a reported shear but assumed to be~ 95% shear based on comparison of the CVN data points known to be at ~ 95% shear for the same material.
f. USE is the average of all available Charpy energy values with reported shear~ 95% per ASTM E185-82 methods. Charpy data included data taken from Reactor Vessel Surveillance Programs baseline test reports.
g. USE determination includes data taken from supplemental Westinghouse test records associated with the surveillance capsule program.
4.4 NOZZLE UPPER-SHELF ENERGY VALUE APPLICABILITY Since the geometry and size of nozzle forgings is different from the beltline forgings, the Rotterdam nozzle forging USE data statistics are compared to USE data statistics for all Rotterdam forgings in Table 8. No evaluation on measured weak-direction data is provided, because no measured weak-direction Charpy data is available for Rotterdam nozzle forgings.
An evaluation of Table 8 indicates that the mean of the nozzle forging USE data tend to be less than the mean of all forging USE data; however, the nozzle USE data have less scatter, which decreases the standard deviation. As a result, the mean minus two standard deviations are in good agreement when comparing the "known" USE for all Rotterdam forgings and the "known" USE for nozzle forgings. It is concluded that the forging USE values calculated herein based on all available Rotterdam forgings are applicable to the Rotterdam nozzle forgings.
Note that most of the nozzle forgings were supplied by Fried-Krupp Huttenwerke AG. A comparison of the results for all Fried-Krupp forgings in Table 6 and those for all Rotterdam forgings in Table 8 shows no difference in the mean minus 2 standard deviations results for the BTP 5-3 analyses. This observation provides further justification of the applicability of the generic USE to the nozzles.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 16 Table 8 Statistical Analysis Comparing All Rotterdam Forgings tothR e otterdam Nozze IFori.mgs All Components, Nozzles, Known USE Known USE Strong BTP Strong BTP Direction 5-3 Direction 5-3 Mean 140 91 126 82 Standard Deviation 23 15 12 8 Mean-2a 94 61 102 66 Maximum 235 153 144 94 Minimum 94 61 94 61
of Components 57 24 Included PWROG-17090-NP March 2018 Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 17 5.0 GENERIC ROTTERDAM WELD UPPER-SHELF ENERGY AND CHEMISTRY The Rotterdam CMTRs identify two types of welds that were used in the fabrication of Rotterdam vessels, shielded metal arc welds (SMAWs) and submerged arc welds (SAWs).
Each weld type is addressed separately in the following sections. Note that the weld type and vendor names are taken as written directly from the original records.
5.1 SUBMERGED ARC WELD (SAW}
This section analyzes the SAW materials available to Westinghouse and utilized by Rotterdam for the manufacturing of reactor vessels with the exception of the welds with Linde 80 flux type.
Linde 80 flux type welds are excluded from this analysis and discussion, because these welds have been thoroughly analyzed previously (e.g., Reference 6). In addition, only one U.S. PWR site is believed to have Rotterdam fabricated reactor vessels which utilized the Linde 80 flux type. Thus, these materials are not considered generically for Rotterdam materials herein.
Table 9 contains all available USE data for the non-Linde 80 surveillance welds produced by Rotterdam. No USE data is available outside of the reactor vessel surveillance program welds.
Table 10 contains the surveillance program chemistry supplemented with all available chemistry data for SAW heats not included in the surveillance programs. The chemistry is based on measurements of the deposited weld chemistry, unless otherwise noted in Table 10.
Table 11 evaluates the USE data for all available Rotterdam SAWs. The results indicate that the average Rotterdam SAW USE minus two standard deviations of 75 ft-lbs is greater than the 10 CFR 50, Appendix G [Ref. 1] minimum irradiated USE screening criterion for operating plants (50 ft-lbs). It is also noted that all measured values are greater than the mean minus two standard deviations value of 75 ft-lbs. This conservative generic value of 75 ft-lbs can be utilized for Rotterdam SAWs when insufficient data is available to determine a weld-specific USE value.
Table 11 provides generic values of 0.23 Cu weight percent and 0.56 Ni weight percent based on a mean plus 1cr. This method is based on Regulatory Guide 1.99, Revision 2, which states that conservative chemistry estimates are a mean plus one standard deviation. If a common heat-flux type combination is shared between multiple welds, the average chemistry value for the heat is considered as one data point when determining the generic weld values as not to assign undue weight to the material, since it is representative of just one heat-flux type combination. The generic weight percent values of 0.23 for Cu and 0.56 for Ni can be utilized for Rotterdam SAWs when insufficient data is available to determine weld-specific chemistry values.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 18 Table 9 Available Rotterdam SAW USE Data USE<a>
Plant(s) Weld Vendor Flux Tvoe (ft-lbs)
Plant A, Plant G Heesch LW320 129 Plant H, & Plant I Plant B Smit-Weld SAF89 112 Plant C Arco Chem SAF89 110 Plant D Westf. U SAF89 128 Plant E Smit-Weld SAF89 95 Plant F Phenix U LW320 109 Plant J Bohler LW320 82 Table 9 Note:
a. The USE value is the average of all available absorbed energy values with a shear ~ 95% per standard ASTM E185-82 methodology. See Section 3.0 for additional details.
Table 10 A va1a*1 bl e Rotterdam SA WCh em1stry Data Cu Cu Ni Weld Flux Lot Averaged<a> Heat Averaged Lot Averaged<a>
Heat Vendor Type Lot (wt.%) (wt.%) (wt.%)
25531 Smit-Weld SAF89 01211 0.10 0.10 0.18 716126 Phenix U LW320 26 0.10 0.10 0.05 01103 0.33 0.17 25295 Smit-Weld SAF89 01135 0.25 0.29 N/A 01170 0.30 N/A Arco 4278 SAF89 01211 0.12 0.12 0.11 Chem 0227 Bohler LW320 14 0.19 0.19 0.56 895075 Heesch LW320 P46 0.04 0.036 0.72 899680 Heesch LW320 P23 0.03 0.03 0.75 1725 Westf. U SAF89 02275 0.11 0.11 0.12 Arco 01180 0.16 N/A 801 SAF89 0.17 Chem 01211 0.18 N/A 01135 0.1 ?'b) N/A 25006 Smit-Weld SAF89 0.17 01985 o.11 N/A 25017 Smit-Weld SAF89 01197 0.33 0.33 N/A Arco 4275 SAF89 02275 0.12 0.12 0.11 Chem Arco 4292 SAF89 02275 0.12 0.12 0.15 Chem Arco 721858 SAF89 01197 0.08 0.08 N/A Chem Table 10 Notes contained on following page.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 19 Table 10 Notes:
a. Lot averaged chemistry values are the average of all available as-deposited weld chemistry measurements for the specific heat and flux lot combination, unless otherwise noted.
b. The chemistry value is based on the weld wire analysis since no chemistry data on the deposited content of the weld is available.
Table 11 Statistical Analysis of SAW Welds Chemistry USE Heat Averaged Ni (ft-lb) Cu (wt.%) (wt.%)
Mean 109 0.14 0.29 Standard Deviation 17 0.09 0.27 Mean-2a 75 - -
Mean+ a - 0.23 0.56 Maximum 129 0.33 0.75 Minimum 82 0.03 0.05
of Welds 7 14 10 5.2 SHIELDED METAL ARC WELD (SMAW)
Table 12 identifies all SMAW heats that were used in reactor vessel fabrication by Rotterdam and that are available in Westinghouse records. Actual USE measurements are only available for three weld materials (Heat #'s 818-025612, 7565.7158, and 7703.7265) in Table 12. For these materials, the actual measured USE is greater than or equal to 116 ft-lbs. The remainder of the available data is based on Charpy tests completed at 10.4°F or lower, and shear values which are either unknown or less than 95%. The USE is typically reached at a temperature greater than 10°F, as demonstrated by the welds with actual measured USE values, which reached the upper-shelf at temperatures of approximately 70°F or higher.
Since insufficient data is available to establish an accurate mean and standard deviation, a conservative lower bound USE value will be determined based on all Charpy impact energies in Table 12. A review of Table 12 indicates that the lowest value based on an unknown shear and Charpy measurements of the specific heat is 72 ft-lbs for Heat# 9092. Heat # 7359.6708 has an impact energy of 63 ft-lbs, which is less than 72 ft-lbs; however, the Charpy test data indicates a maximum shear value of only 52%. The actual USE for Heat # 7359.6708 is therefore expected to be much greater than 63 ft-lbs. Since a USE value of 72 ft-lbs is only 9 ft-lbs greater, it is determined that 72 ft-lbs is a bounding USE value for Heat # 7359.6708.
Therefore, 72 ft-lbs is considered to be a bounding and conservative USE value for a Rotterdam-fabricated SMAW with an unknown USE.
Since only two welds with test results for Cu weight percent exist, insufficient information is available to determine a generic Cu value; thus, the Regulatory Guide 1.99, Revision 2 [Ref. 7],
generic value of 0.35% for Cu can be used for an unknown Rotterdam-fabricated SMAW. The PWROG-17090-NP March 2018 Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 20 limited data available indicates that a value of 0.35% is very conservative. Per Table 13, the chemistry data indicates that a generic value of 1.13% for Ni is acceptable for unknown heats of SMAW used by Rotterdam. This value is based on the Regulatory Guide 1.99, Revision 2 mean plus one standard deviation approach. This value of 1.13% is conservative and greater than the Regulatory Guide 1.99, Revision 2 generic value of 1.0%.
Table 12 "I bl e Rotterd am SMAW USE an d Ch em1stry A va1a . t Daa t USECbl Shear<c> cu<d> Ni<dl Heat Type<a> Vendor<a> (ft-lb) (%) (wt.%) (wt.%)
818- ~ 91(e)
E8015-G B&W 85 N/A 0.87 021736 818-E8015-G B&W ~ 97 N/A N/A 1.11 022108 818- ~ 77(fJ E8015-G B&W N/A N/A 1.07 022778 818- ~ 77(fJ E8015-G B&W N/A 0.01 1.15 023006 818- ~ 77(fJ E8015-G B&W N/A N/A 1.06 024509 818- ~ 77'fJ E8015-G B&W N/A N/A 0.96 024510 818- ~ 77(fJ E8015-G B&W N/A N/A 0.97 024790 818- ~
E8015-G B&W 77(fJ N/A N/A 0.97 024965 818- ~ 77'fJ E8015-G B&W N/A N/A 0.95 025134 818- ~ 77'fJ E8015-G B&W N/A N/A 1.04 025185 818- ~ 77'fJ E8015-G B&W N/A N/A 1.04 025186 818- ~ 77(fJ E8015-G B&W N/A NIA 0.81 025371 818- ~
E8015-G B&W 77(fJ N/A N/A 0.90 025391 818- ~ 77(fJ E8015-G B&W N/A N/A 0.92 025392 818- ~ 77'fJ E8015-G B&W N/A N/A 0.81 025561 818- ~ 77'fJ E8015-G B&W N/A N/A 0.85 025562 818- ~ 77(fJ E8015-G B&W N/A N/A 0.87 025611 818- 130<9 )
E8015-G B&W 100% 0.023 0.91 025612 818- ~ 77'fJ E8015-G B&W N/A N/A 0.76 025655 PWROG-17090-NP March 2018 Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 21 Table 12
1 bl e Ro tterd am SMAW USE an d Ch em1s:rv A va1a . t Daa t USE Shear<cl cu<d> Ni<d>
Heat Type<a) Vendor<a> (ft-lb) (%) (wt.%) (wt.%)
401W9661 E8018-C3 RACO ~ 166 N/A N/A 0.97 KG66ELH, Soudo-5835.3423 ~ 98 N/A N/A 1.15 E9018-G metal KG66ELH, Soudo-5835.3900 ~ 80 N/A N/A 1.25 E9018-G metal KG66ELH, Soudo-6236.4063 ~ 120 N/A N/A 1.24 E9018-G metal KG66ELH, Soudo-6236.4450 ~ 95 N/A N/A 1.13 E9018-G metal KG66ELH, Soudo-6497.4647 ~ 73 N/A N/A 1.04 E9018-G metal KG66ELH, Soudo-6497.4675 ~ 83 N/A N/A 1.04 E9018-G metal KG66ELH, Soudo-6507.4705 ~ 76 N/A N/A 1.36 E9018-G metal KG66ELH, Soudo-6747.5458 ~ 96 N/A N/A 0.90 E9018-G metal KG66ELH, Soudo- ~ 87'e) 7011.6032 68 NIA 0.93 E9018-G metal KG66ELH, Soudo- ~ 108(e) 7011.6143 74 N/A 0.91 E9018-G metal KG66ELH, Soudo- ~ 63(e) 7359.6708 52 N/A 0.83 E9018-G metal KG66ELH, Soudo- 116(9 )
7565.7158 100 N/A N/A E9018-G metal KG66ELH, Soudo- 134(9 )
7703.7265 100 NIA 0.94 E9018-G metal Molyth.,
8640 Secher. ~ 103 N/A N/A N/A E8015-G Molyth.,
8825 Secher. ~ 85 N/A NIA N/A E8015-G Molyth.,
8928 Secher. ~ 96 N/A N/A N/A E8015-G Molyth.,
9004 Secher. ~ 112 N/A N/A N/A E8015-G Molyth.,
9092 Secher. ~ 72 N/A N/A N/A E8015-G Table 12 Notes (continued on following page):
a. The weld type and vendor names are taken directly from the original records.
b. The USE values are the average of all available absorbed energy values from Charpy tests completed at 10.4°F or below with no available shear data or with limited shear data (all available values are less than 95%), unless otherwise noted. The actual USE values are expected to be much greater in many cases.
c. Identifies the shear value corresponding to the lower bound USE. NIA indicates that there is no shear information available. Values of 100 correspond to multiple test specimens showing 100% shear.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 22
d. When multiple measurements are available, the chemistry values are the average of all available measurements for the heat.
e. The USE value is the maximum Charpy value recorded with a shear less than 95%, as no values of shear above 95% are recorded.
f. Mechanical test data is not available for all type E8015-G weld heats. However, a non-conformance review performed by Rotterdam determined the acceptability of the material, i.e. Charpy results greater than or equal to 30 ft-lbs at 10°F, based on Charpy tests results available for 27 different heats of type E8015-G welds. A review of these 27 heats is presented in Appendix A of this report.
The review calculated a mean minus 2cr value of 77 ft-lbs.
g. The USE value is the average of all available absorbed energy values with a shear~ 95% per ASTM E185-82. See Section 3.2 for additional details.
Table 13 Statistical Analysis of SMAW Weld Nickel Weight Percent Ni (wt.%)
Mean 0.99 Standard Deviation 0.14 Mean +a 1.13 Maximum 1.36 Minimum 0.76
of Materials Included 32 5.3 WELD ANALYSIS
SUMMARY
The previous subsections provide generic information that can be used when material-specific Rotterdam SAW or SMAW information is unavailable. If insufficient data exists to determine whether a Rotterdam weld is a SAW or a SMAW, then the generic SMAW properties can be utilized. The generic SMAW USE, Cu weight percent, and Ni weight percent values are all more limiting than the corresponding SAW values. Thus, for an unknown Rotterdam weld, the USE can be set to 72 ft-lbs; the Cu weight percent can be set to 0.35; and the Ni weight percent can be set to 1.13.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 23
6.0 REFERENCES
1. Code of Federal Regulations 10 CFR 50, Appendix G, "Fracture Toughness Requirements,"
Federal Register, Volume 60, No. 243, December 19, 1995.
2. U. S. NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," October 14, 2014. [Agencywide Documents Access and Management System (ADAMS) Accession Number ML14149A165]
3. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.
4. ASTM E185-16, "Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels," ASTM International, December 2016.
5. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position (BTP) 5-3, Revision 2, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, March 2007.
[ADAMS Accession Number ML070850035]
6. AREVA NP, Inc. Report BAW-2313, Revision 7, Supplement 1, Revision 1, "Supplement to B&W Fabricated Reactor Vessel Materials and Surveillance Data Information for Surry Unit 1 and Unit 2," AREVA Document No. 77-2313S-007-001, February 2017.
7. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
[ADAMS Accession Number ML003740284]
8. Code of Federal Regulations 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 Appendix A Supplemental Charpy Impact Energy Data for E8015-G Electrode Welds PWROG-17090-NP March 2018 Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 For some type E8015-G weld heats, mechanical test data is not available. However, a Rotterdam non-conformance review identified Charpy results for 27 separate 8015-type electrodes manufactured in the same shop as those utilized at Rotterdam within the previous 5 years. The results of this non-conformance review are documented and analyzed in Table A-1 to provide a surrogate USE for the materials without specific test data. No shear data is available.
Table A-1 sup p emen ta I Ch arpy -
mpact Enerav Data for E8015 G El ect ro de W e Id s Type 8015 CVN Data at 10°F CVN Data at 10°F CVN Data at 10°F Averaged Material Test#1 Test#2 Test#3 CVN Data Number (ft-lbs) (ft-lbs) (ft-lbs) (ft-lbs) 1 91 95 95 94 2 100 101 149 117 3 65 76 89 77 4 80 83 105 89 5 88 90 100 93 6 99- 105 105 103 7 70 84 75 76 8 102 104 107 104 9 84 92 95 90 10 109 117 120 115 11 118 118 125 120 12 90 91 96 92 13 65 90 91 82 14 91 99 100 97 15 95 100 103 99 16 94 96 111 100 17 87 94 97 93 18 94 103 105 101 19 105 110 118 111 20 95 98 102 98 21 98 98 103 100 22 83 96 102 94 23 105 119 120 115 24 109 110 112 110 25 95 95 110 100 26 95 100 104 100 27 94 97 99 97 Mean 99 Standard Deviation 11 Mean - 2cr 77 PWROG-17090-NP March 2018 Revision 0
_ PWROG-17090-NP Revision 0 Proprietary Class 3
This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**
Author Approval Lynch Donald Mar-27-2018 15:53:08 Reviewer Approval Mays Benjamin E Mar-28-2018 06:53:51 Approver Approval Molkenthin James Mar-28-2018 08:07:23 Approver Approval Patterson Lynn Mar-28-2018 08: 10:28 Final Approval Lynch Donald Mar-28-2018 11:08:19 Files approved on Mar-28-2018

ML18291A837

Text

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

3.0 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

1.0 INTRODUCTION

SUMMARY

6.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

6.0 REFERENCES

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