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{{Adams
#REDIRECT [[L-2018-004, Enclosure 4 - Non-Proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)]]
| number = ML18037A837
| issue date = 01/30/2018
| title = Enclosure 4 - Non-Proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
| author name =
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number =
| contact person =
| case reference number = L-2018-004
| package number = ML18037A812
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 903
}}
 
=Text=
{{#Wiki_filter:Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4
 
Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
(13 Attachments)
(903 Total Pages, including cover sheets)
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4  Document List  - LTR-REA-17-116-NP, Revision 0, Reactor Vessel Neutron Exposure Data in Support of the Turkey Point Unit 3 and Unit 4 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA), December 1, 2017 
- Areva Topical Report ANP-3646NP Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels C & D Service Loads at 80 Years, January 5, 2018 (Non-
 
Proprietary) 
- Areva Topical Report, ANP-3647NP Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels C & D Service Loads at 80 Years, January 5, 2018 (Non-
 
Proprietary) Attachment 4 - SIA Report No. 1700109.402P, Revision 3 - REDACTED, Evaluation of Fatigue of ASME Section III, Class 1 Components for Turkey Point Units 3 and 4 for Subsequent License Renewal, December 2017 
- SIA Report No. 1700109.401P, Revision 3 - REDACTED, Evaluation of Environmentally-Assisted Fatigue for Turk ey Point Units 3 and 4 for Subsequent License Renewal, January 2018  - SIA Environmentally Assisted Fatigue Calculations Pressurizer Lower Head 1700804.316P - REDACTED, Revision 0, 3-D Finite Element Model of Pressurizer Bottom Head, Skirt Assembly and H eater Wells, September 28, 2017 1700804.317, Revision 0, Pressurizer Low er Head Green's Functions and Unit Pressure, October 5, 2017 1700804.318, Revision 0, Pressurize r Lower Head Loads, Fatigue and EAF Analysis, November 7, 2017 Pressurizer Spray Nozzle 1700804.313P - REDACTED, Revision 1, Pressurizer Spray Nozzle Loads, December 7, 2017 1700804.314P - REDACTED, Revision 1, Pressurizer Spray Nozzle Finite Element Model and Stress Analyses, December 7, 2017 1700804.315P - REDACTED, Revision 1, Pressurizer Spray Nozzle Fatigue Analysis, December 7, 2017 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 
- Westinghouse Environmentally Assisted Fatigue Calculations LTR-SDA-II-17-13-NP, Rev. 2, Environmentally Assisted Fatigue Evaluation of the Turkey Point Unit 3 and Unit 4 Pressu rizer Upper Head and Shell and Reactor Vessel Core Support Blocks, November 30, 2017 LTR-CECO-17-025-NP, Rev. 1, Environmentally Assisted Fatigue Evaluation of the Turkey Point Unit 3 and Unit 4 Replacement Steam Generators, November 30, 2017 Attachment 8
- Areva Environmentally Assisted Fa tigue Calculations: Areva Letter No.
AREVA-17-02742, date Decemb er 6, 2017, Final CUF EN Results - Turkey Point 3 & 4 - SLR EAF Analyses I. 32-9280707, Rev. 0, Turkey Point -3
& 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation, December 15, 2017 II. 32-9280708, Rev. 0, Turkey Point 3 & 4 Replacement RVCH J Groove, December 12, 2017 III. 32-9280709, Rev. 0, 12/15/17, TP CRDM Latch Housing Environmentally Assisted Fatigue, December 15, 2017 IV. 32-9280710, Rev. 0, TP Vent Nozzle Environmentally Assisted Fatigue, December 14, 2017 V. 32-9280711, Rev. 0, Turkey Point SLR EAF Analysis for Reactor Vessel Flange, December 14, 2017 VI. 32-9280712, Rev. 0, TP CRDM Lower Joint Environmentally Assisted Fatigue, December 15, 2017 
- PWROG-17031-NP, Rev. 0, Update for Subsequent License Renewal:
WCAP-15338-A, A Review of Cracking A ssociated with Weld Deposited Cladding in Operating PWR Plants, August 2017  0
- PWR Owners Group, PWROG-17011-NP, Rev. 0, Update for Subsequent License Renewal: WCAP-14535A, Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination' and WCAP-15666-A, Exte nsion of Reactor Coolant Pump Motor Flywheel Examination, November 2017  1 - WCAP-15354-NP, Revision 1,  Techni cal Justification for Eliminating Primary Loop Pipe Rupture as a Structural Design Basis for Turkey Point Units 3 and 4 Nuclear Power Plants for the Subsequent License Renewal Time-Limited Aging Analysis Program (80 Years) Leak-Before-Break Evaluation, August 2017  2 - SIA Leak-Before-Break Evaluation for Auxiliary Lines 0901350.401, Revision 3, Leak-Before-Break Evaluation Accumulator, Pressurizer Surge and Residual Heat Removal Lines Tu rkey Point Units 3 and 4, September 2017 0901350.304, Revision 3, Fatigue Crack Growth Evaluation, September 18, 2017 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4  3 - PWROG-17033- NP, Revision 0, Update for Subsequent License Renewal: WCAP-13045, Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Ty pe Nuclear Steam Supply Systems, October 2017
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 1Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
LTR-REA-17-116-NP, Revision 0 Reactor Vessel Neutron Exposure Data in Support of the Turkey Point Unit 3 and Unit 4 Subsequent License Renewal (SLR) Time-Limited Aging Analysis (TLAA), December 1, 2017 (59 Total Pages, including cover sheets)
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 2Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
Areva Topical Report ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80 Years, January 5, 2018 (Non-proprietary)
(67 Total Pages, including cover sheets)
 
ANP-3646NP Revision 0
 
Low Upper-Shelf Toughness Fracture
 
Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report
 
January 2018 AREVA Inc.
(c) 2018 AREVA Inc.
ANP-3646NP Revision 0 Copyright © 2018 AREVA Inc.
All Rights Reserved AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page i    Nature of Changes Item Section(s) or Page(s) Description and Justification 1 All Initial Issue
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page ii    Contents Page
 
==1.0 INTRODUCTION==
............................................................................................... 1-1
 
===1.1 Equivalent===
Margins Analysis-A nalysis of Record .................................. 1-3
 
===2.0 REGULATORY===
REQUIREMENTS .................................................................... 2-1
 
===2.1 Regulatory===
Requ irements ....................................................................... 2-1
 
===2.2 Compliance===
with 10 CFR 50 Ap pendix G and Acceptance Criteria .................................................................................................... 2-2
 
====2.2.1 Acceptance====
Criteria Levels A and B ............................................. 2-3
 
==3.0 DESCRIPTION==
OF TURKEY P OINT REACTOR VESSELS ............................. 3-1
 
===4.0 MATERIAL===
PRO PERTIES................................................................................. 4-1 4.1 J-Integral Resist ance Model ................................................................... 4-1
 
===4.2 Mechanical===
Properties of Weld Metals .................................................... 4-4
 
====4.2.1 Mechanical====
Properties for the Turkey Point Reactor Vessels ........................................................................................ 4-4
 
===5.0 FRACTURE===
MECHANICS ANALYSIS .............................................................. 5-1
 
===5.1 Methodology===
........................................................................................... 5-1
 
===5.2 Procedure===
for Evaluating Levels A and B Service Loadings ................... 5-2
 
===5.3 Evaluation===
for Flaw Extension
................................................................. 5-6
 
====5.3.1 Reactor====
Vessel S hell Welds ......................................................... 5-7
 
====5.3.2 Reactor====
Vessel Transition Welds and RV Nozzle Welds ........................................................................................... 5-7
 
===5.4 Evaluation===
for Flaw Stability .................................................................... 5-8
 
====5.4.1 Reactor====
Vessel S hell Welds ......................................................... 5-9
 
====5.4.2 Reactor====
Vessel Transition Welds and RV Nozzle Welds ........................................................................................... 5-9 6.0
 
==SUMMARY==
AND CONCLUSIONS .................................................................... 6-1
 
===6.1 Reactor===
Vessel Shell Welds
.................................................................... 6-1
 
===6.2 Reactor===
Vessel Transition Weld s and RV Nozzle Welds ........................ 6-1
 
==7.0 REFERENCES==
.................................................................................................. 7-1
 
===8.0 CERTIFICATION===
............................................................................................... 8-1 APPENDIX A B&WOG J-R MODEL-DATA ANALYSIS AND EMPIRICAL MODEL DEVELOPMENT .................................................................................. A-1
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page iii List of Tables Table 3-1 Reactor Vessel Weld Locations--Copper Content and 80-Year Fluence Projections ................................................................................... 3-2 Table 4-1 Parameters in Jd Model 4B ....................................................................... 4-3 Table 4-2 Mechanical Properties of Turk ey Point RV Shell Materials ........................ 4-5 Table 5-1 Reactor Vessel Shell Dimensions and Operating Conditions ................. 5-10 Table 5-2 Plant Specific Flaw Ev aluation Summary for RV Shell Regions .............. 5-11 Table 5-3 Reactor Vessel Nozzle Be lt Dimensions ................................................. 5-12 Table 5-4 Flaw Evaluation Summary of Turk ey Point Upper Transition and RV Nozzle-to-Shell Welds ............................................................................. 5-13 Table 5-5 Flaw Evaluation Summary of Turkey Point Lower Transition Welds
........ 5-13 Table 5-6 Applied J-Integral versus Flaw Ext ensions of Turkey Point Controlling RV Shell Weld (SA-1101) ........................................................................ 5-14 Table 5-7 Mean & Lower Bound J-R Curve Values for Turkey Point Controlling RV Shell Weld  (SA-1101) ....................................................................... 5-15 Table A-1 Model 4B, Range of Test Data .................................................................. A-3 Table A-2 New B&WOG Specimen Data fo r Model Assessment ............................... A-5 Table A-3 Jd Model Coefficients (M odels 4B, 5B, and 6B) ...................................... A-10 Table A-4 EMA Reconciliation for Limiting RV Shell Welds-Models 4B and 6B .... A-18
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page iv    List of Figures Figure 3-1 Reactor Vessel-Turkey Point Unit 3 .................................................... 3-3 Figure 3-2 Reactor Vessel-Turkey Point Unit 4 .................................................... 3-4 Figure 5-1 J-Integral Flaw Extension for Turkey Point Controlling Reactor
 
Vessel Shell weld (SA-1101) ............................................................... 5-16 Figure 5-2 J-Integral versus Flaw Extension for Turkey Point Inlet Nozzle ........... 5-17 Figure A-1 BAW-2251A, Appendix B, Figure 3-1 .................................................... A-4 Figure A-2 Jd (0.1) vs Fluence B&WOG J-R Model 4B and New Test Data (Normalized to Standard Conditions)
..................................................... A-7 Figure A-3 Original and New Data and Model Fit Normalized at Standardized
 
Conditions vs a.................................................................................. A-11 Figure A-4 Original and New Data and Model Fit Normalized at Standard
 
Conditions vs Fluence ......................................................................... A-12 Figure A-5 Model 6B Residuals vs Fitted Values .................................................. A-13 Figure A-6 Model 6B Standardized Residu als vs Fitted Values ............................ A-14 Figure A-7 Normal Q-Q Plot of Standar dized Residuals ....................................... A-15 Figure A-8 Comparison of Models 4B, 5B, and 6B at Standard Conditions .......... A-17
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page v    Nomenclature Acronym Definition B&W Babcock and Wilcox B&WOG Babcock and Wilcox Owners Group CvUSE Charpy Upper Shelf Energy EFPY Effective Full Power Years EMA Equivalent Margins Analysis INF Inlet Nozzle Forging Jd J deformation J-R J-integral Resistance LAR License Amendment Request ONF Outlet Nozzle Forging PTN Turkey Point Plant PWROG Pressurized Water Reactor Owners Group RV Reactor Vessel RVWG Reactor Vessel Working Group SLR Subsequent License Renewal SRP Standard Review Plan Sy Yield Strength TSs Technical Specifications USE Upper Shelf Energy
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page vi    ABSTRACT This topical report  presents the result s of an equivalent margins analysis (EMA) considering Levels A and B service loads fo r high copper Linde 80 weld metals using fluence values expected at 80-years (subsequent license renewal--SLR). This report applies to the following Westinghouse-desig ned reactor vessels fabricated by B&W:
Turkey Point Plant (PTN) Units 3 and 4.
Note that the Turkey Point EMA reported herein is technically identical to the Turkey Point EMA reported in BAW-2192P, Supplement 1, Revision 0, which wa s submitted to the NRC by the PWROG on December 15, 2017. That is, Sections 1.
0 through 7.0 of ANP-3646P were generated by extracting Turkey Point 3 and 4-specific results from Sections 1.0 through 7.0 of BAW-2192, Supplement 1, Revision 0.
Appendix A to ANP-3646 P, B&WOG J-R Model-Data Analysis and Empirical Model Development, is identical to Appendix A of BAW-
 
2192, Supplement 1, Revision 0, with the exception that refe rences to plants other than Turkey Point 3 and 4 were removed from Sections A.1, A.2, and A.4.
The analytical procedure used in this topica l report is in accordance with ASME Section XI, Appendix K, Subarticle K-1200. EMA resu lts are reported for all reactor vessel weld locations with 80-year fluence proj ections that exceed 1.0 E+17 n/cm 2 (E> 1.0 MeV).
The ASME Section XI, acceptance criteria for Levels A & B Service Loads for all reactor vessel shell welds are satisfied. The accept ance criteria for Levels A & B Service Loads for RV transition welds and RV nozzle welds are also satisfied. Consistent with BAW-2192PA, Revision 00, the B&WOG J-R Model 4B is used for Linde 80 welds.
Model 4B was developed based on fracture toughness test data obtained through approximately 1990, with specimen fluence t hat ranges from 0.
0 to 8.45E+18 n/cm
: 2. Eighty-year fluence estimates for Turkey Point Units 3 and 4 exceeds 8.45E+18 n/cm 2  (e.g., maximum 80-year 1/4 T fluenc e is estimated at 6.53E+19 n/cm
: 2) and use of Model 4B to estimate J-integral resistance values, including the associated model uncertainty, for 80-years is made by extrapol ation of the model. To assess the model extrapolation uncertainty, Model 4B is compared to new fracture toughness test data AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page vii (1990 to 2017) irradiated to fluence r anging from approximately 8.0E+18 n/cm 2 to 5.8E+19 n/cm
: 2. The majority of test data fell above the Model 4B mean and all of the test data fell above the Model 4B mean minus 2 standard error band. Therefore, use of Model 4B and associated uncertainty to extr apolate J-integral resistance for 80-year fluence applications was determined to be appropriate. This assessment is reported in Appendix A herein.
To further substantiate the us e of Model 4B, all of the or iginal fracture toughness data used to develop Model 4B was combined with new fracture toughness data, using the same model form, and a new Model 6B was generated. Model 6B was found to be
 
essentially equivalent to Model 4B with res pect to model mean and 2 standard errors.
The EMA results reported herein using Model 4B were reconciled to Model 6B, with little or no change to the EMA results. Model 6B development and the EM A reconciliation to Model 4B are reported in Appendix A.
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 1-1 
 
==1.0 INTRODUCTION==
The purpose of this topical report  is to pr esent  the results of an equivalent margins analysis (EMA) considering Levels A and B service loads for high copper Linde 80 weld metals and applicable non-Linde 80 welds using fluence values expected at 80-years (subsequent license renewal--SLR). This topical report applies to the following Westinghouse-designed reactor vessels:  Turkey Point Plant (PTN) Units 3 and 4, also referred to as Turkey Point Units 3 and 4. Note that the Turkey Point EMA reported herein is technically identical to the Turkey Point EMA reported in BAW-2192P, Supplement 1, Revision 0 [3], which was submitted to the NRC by the PWROG on December 15, 2017. That is, Sections 1.
0 through 7.0 of ANP-3646P were generated by extracting Turkey Point 3 and 4-specific results from Sections 1.0 through 7.0 of BAW-2192, Supplement 1, Revision 0.
Appendix A to ANP-3646 P, B&WOG J-R Model-Data Analysis and Empirical Model Developmen t,  is identical to Appendix A of BAW-2192, Supplement 1, Revision 0, with the exception that refe rences to plants other than Turkey Point 3 and 4 were removed from Sections A.1, A.2, and A.4.
Equivalent margins analyses fo r Turkey Point Units 3 and 4 are reported for all reactor vessel weld locations with 80-year fluenc e projections that exceed 1.0 E+17 n/cm 2  (E> 1.0 MeV) [2]. Upper shelf energy evaluations at reactor vessel base metal locations with 80-year fluence projections greater than 1.0 E+17 n/cm 2, if needed,  will be addressed separately in the Turkey Point Units 3 and 4 subsequent license renewal application. The EMA utilizes the B&WOG J-integral resistance (J-R) Model 4B reported in BAW-2192PA, Appendix B, [1]. The followi ng groups are used for the welds within the scope of this report:
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 1-2 Reactor Vessel Shell Welds-circumferential welds for  Turkey Point Units 3 and 4 (also referred to as Turkey Point 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. Transition Welds and RV Nozzle Welds-welds that are located above and below the reactor vessel shell welds that may experience 80-year fluence greater than 1.0 E+17 n/cm 2 [2] and, must consider the effects of neutron irradiation embrittlement. In addition, the transition welds are located at geometric discontinuities (e.g., lower shell to lower head and upper shell to nozzle belt forging).
These locations may or may not have been included as par t of the 10 CFR 50 Appendix G [4] "beltline" def inition 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.
The EMA evaluations in this report are for a ll weld locations expected to receive fluence
> 1.0E+17 n/cm 2 [2] at 80 years. The use of the terms "beltline" and/or "extended beltline" are not used in this report.
The 60-year EMA summary report for Turk ey Point Units 3 and 4 are reported in Section 1.1. Section 2.0 provides the current NRC regulatory requirements for the EMA. Section 3.0 provides a description of the reactor vessels within the scope of this report, with illustrations of reactor vessel welds that are evaluated for equivalent margins in Figures 3-1 through 3-2. Section
 
===4.0 provides===
the material properties that are required for the EMA, and Se ction 5.0 presents the results of the EMA. Section 6.0 provides the summary and conclusions, Secti on 7.0 lists all references and Appendix A provides the technical basis for the use of B&WOG J-R Model 4B for the EMA reported
 
herein.
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 1-3 
 
===1.1 Equivalent===
Margins Analysis-Analysis of Record The summary reports for the PTN EMA analyses of record are as follows: Turkey Point Units 3 and 4 The Turkey Point Units 3 and 4 current licensing basis equivalent margins analysis at 48 EFPY is summarized in Section 2.1.2 of NRC document "TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING EXTENDED POWER UPRATE (TAC NOS. ME4907 AND ME4908)," Adam s Accession number ML 11293A365 [11]. NRC acceptance of the Turkey Point EMA at 48 EFPY for EPU is based on the following documentation:  LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE, ATTACHMENT 4, L-2010-113, Attachm ent 4  ADAMS
--ML103560177 [12]  Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) Lic ense Amendment Request (LAR) No. 205 and Equivalent Margin An alysis (EMA), L-2010-303, ADAMS-ML103610321 [13].
Reference [13] contains AREVA docum ent 77-2312-03 (P), LOW UPPER-SHELF TOUGHNESS FRACTURE MECHANICS ANALYSIS OF REACTOR VESSELS OF TURKEY POINT UNITS 3 AND 4 FOR EXTENDED LIFE THROUGH 48 EFFECTIVE FULL POWER YEARS  Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1, L-2011-029, ADAMS--ML110700068 [14]. Reference [14] provides a response to RAI CVIB-1.2 regarding the code year used to perform the equivalent margins analysis (i.e., 1998 Edition vs 2004 Edition). The NRC SER for the Turkey Point Extended Power Uprate, Section 2.1.2, contains the evaluation of upper shelf energy--ADAMS-ML 11293A365 [11]
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B 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 [4
], IV, A, 1, Reacto r 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 R egulation or Director, Office of New Reactors, as appropriate, that lower va lues 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 anal ysis 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 t he 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 y ears prior to the dat e when the predicted Charpy upper-shelf energy will no longer satisfy the require ments 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.
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 2-2 When the reactor vessels within the scope of this report were f abricated, Charpy V-notch testing of the reactor vessel we lds was in accordance with the original construction code, which did not specifically require Charpy V-notch tests on the upper shelf.
The construction code is as follows:  Turkey Point--ASME B&PV Code, Se ction III, 1965 Edition through Summer 1966 Addenda In accordance with NRC Regulatory Guide 1.161 [15], the NRC has determined that the analytical methods described in ASME Sect ion 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 the limiting design transient (i.e., cooldown at 100 F/h) is consistent with BAW-2192PA [1
], Section 5.3. Section 4.1 of this report includes a summa ry of the B&WOG J-integral resistance model, and Section 4.2 provides mechani cal properties of weld metals. The Linde 80 weld locations t hat are included within the scope of this report (i.e., weld locations with 80-year projected fluence > 1.0E+17 n/cm
: 2) 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 t he 2007 Edition with 2008 Addenda [16] of Section XI of the ASME Code, Appendix K.
The current edition of ASME Section XI listed in 10 CFR 50.55a is the 2013 Edition [17]. With regard to Appendix K, there ar e 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.
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 2-3 The material properties used in this anal ysis are based on ASME Section II, Part D, 2007 Edition with 2008 Addenda. The only change in the material properties listed in the 2013 Edition of ASME Section II for the applicable properties is the coefficient of thermal expansion for stainless steel at 600°F; this value was changed from 9.8E-6 in/in/°F to 9.9E-6 in/in/°F. This does not impact the Levels A and B evaluation reported herein. 2.2.1 Acceptance Criteria Levels A and B ASME Section XI [17], Subarticl e K-2200, provides the followin g acceptance criteria for Levels A and B Service Conditions: a) When evaluating adequacy of the upper s helf toughness for the weld material for Levels A and B Service Loadings, an interior semi-elliptical surface flaw with a depth one-quarter of the wall thickness and a length six times the depth shall be postulated, with the flaw's major axis orie nted along the weld of concern, and the flaw plane oriented in the radial dire ction. When evaluating adequacy of the upper shelf toughness for the base mate rial, both interior axial and circumferential flaws with depths one quarter of the wall thickness and lengths six times the depth shall be postulat ed, and toughness properties for the corresponding orientation shall be used. Smaller flaw sizes may be used when justified. Two criteria shall be satisfied: 1. The applied J-integral evaluated at a pressure 1.15 times the accumulation pressure as defined in the plant specif ic Overpressure Protection Report, with a structural factor of 1 on thermal loading for the plant specific heatup and cooldown conditions, shall be less than the J-integral of the material at a ductile flaw extension of 0.1 in. (2.5 mm). 2. Flaw extensions at pressures up to 1.25 times the accumulation pressure of K-2200(a)(1) shall be ductile and stable, using a structural factor of 1 on thermal loading for the plant specific heatup and cooldown conditions.
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 2-4 b) The J-integral resistance versus flaw extension curve shall be a conservative representation for the vesse l material under evaluation.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 3-1 
 
==3.0 DESCRIPTION==
OF TURKEY POINT REACTOR VESSELS The Turkey Point reactor vessels and applicable weld locations are shown in Figures 3-1 through 3-2. All weld locations evaluated for equivalent margins in this report are identified by an asterisk (*) in eac h Figure. Plant-specific weld copper and nickel content and 80-year fluence projections data needed for the equivalent margins analysis are provided in Table 3-1. The fl uence projections are reported for all reactor vessel weld locations that are expected to exceed 1.0E+17 n/cm 2 at 80-years. The 80-year fluence projections are conservati ve estimates based on detailed transport calculations completed by Westinghouse Elec tric Corporation using a methodology that is in compliance with Regulatory Guide 1.190 (i.e., Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heat up and Cooldown Limit Cu rves," May 2004.)
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 c ontent for transition welds and RV nozzle-to-nozzle belt forging welds reported in Table 3-1 were obtained from either the EMA analysis of record or Turkey Point reactor vessel fabrication reports. 
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 3-2 Table 3-1  Reactor Vessel Weld Locations--Copper Content and 80-Year Fluence Projections Reactor Vessel Material Material ID and/or Heat Number  Cu, wt% Ni, wt% (IS) Inside Wetted Surface Fluence or (*) clad/base metal n/cm 2 E> 1.0 MeV) Turkey Point Unit 3, 80-Year Fluence (E > 1.0 MeV)
US Forging to INF Welds Heat 8T17620.190.57(IS) 1.50E+18 Heat 8T1554B0.160.57(IS) 1.50E+18 Heat 712490.230.59(IS) 1.50E+18 US Forging to ONF Welds Heat 8T17620.190.57(IS) 1.50E+18 US to IS Circ. Weld SA-1484 Heat 72442 0.26 0.60 (*) 1.19E+19 IS to LS Circ. Weld SA-1101 Heat 71249 0.23 0.59 (*) 1.04E+20 LS to Dutchman Circ. Weld SA-1135 Heat 61782 0.23 0.52 (IS) 1.5E+18 Turkey Point Unit 4, 80-Year Fluence (E > 1.0 MeV)
US to INF Welds Heat 8T1762 0.19 0.57 (IS) 1.50E+18 Heat 8T1554B 0.16 0.57 (IS) 1.50E+18 Heat 299L44 0.34 0.68 (IS) 1.50E+18 US to ONF Welds Heat 8T1554B 0.16 0.57 (IS) 1.50E+18 Heat 299L44 0.34 0.68 (IS) 1.50E+18 US to IS Circ. Weld (ID 67%) WF-67 Heat 72442 0.26 0.60 (*) 1.21E+19 IS to LS Circ. Weld SA-1101 Heat 71249 0.23 0.59 (*) 1.03E+20 LS to Dutchman Circ. Weld SA-1135 Heat 61782 0.23 0.52 (IS) 1.5E+18 AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 3-3  Figure 3-1  Reactor Vessel-Turkey Point Unit 3
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 3-4  Figure 3-2  Reactor Vessel-Turkey Point Unit 4
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-1  4.0 MATERIAL PROPERTIES 4.1 J-Integral Resistance Model The J-integral resistance model for Mn-Mo-Ni/
Linde 80 welds in the reactor vessels of the B&WOG RVWG plants were developed using a large J-resistance data base. A detailed description of this model is provided in Appendix B of BAW-2192PA [1], Revision 00. This model was developed us ing specimens irradiated to 8.45E+18 n/cm 2 , and the range of applicability of the model was extended (qualitativel y) to approximately 1.90 E+19 n/cm 2 in Appendix B, Figure 3-1, to BAW-2251A [5]. See Appendix A of this report for a discussion of t he extension of the range of applicability of the B&WOG J-R model to fluence values expected at 80 years for Turkey Point Units 3 and 4.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-2  The coefficients a, d, and C4 are provided in Table 4-1. As required by ASME Section XI, ASME K-3300, when evaluati ng the vessel for Levels A, B, and C Service Loadings, the Jintegral resistance versus crackextension curve (JR curve) shall be a conservative representation of the toughness of the controlling beltline material at upper shelf temperatures in the oper ating range. As such, the Jd correlation minus 2 standard errors is used for evaluation of Levels A & B service loadings (i.e., equation (1) multiplied by
[  ] ). As discussed in Appendix B to BAW-2192PA, the JR curve was generated from a Jintegral database obtained from the same class of material with the same orientation as the applicable reactor vessel materials using co rrelations for the effe cts of temperature, chemical composition, and fluence level. Cra ck extension was by duc tile tearing with no cleavage. This complies with the ASME Code, Section XI, K-3300.
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-3  Table 4-1  Parameters in Jd Model 4B
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-4  4.2 Mechanical Properties of Weld Metals The following subsections provide representative properties for the Turkey Point reactor vessels. The temperature dependent mechanica l properties are developed from the 2007 Edition with 2008 Addenda of the ASME Code (Section II, Pa rt D) for the reactor base metal and cladding (the ASME Code does not provide separate mechanical properties for base and weld metal). Both ASME Code minimum and representative irradiated yield strengths are also provided.
The mechanical properties such as weld metal yield strengths typically used were the i rradiated properties but in some cases the ASME Code minimum properties were conser vatively considered. The irradiated
 
material properties used herein are consist ent with the PTN 60-year license renewal low upper shelf toughness analysis submitta l (See Section 1.1 above). 4.2.1 Mechanical Properties for the Turkey Point Reactor Vessels The Turkey Point reactor vessels are fabricated using A-508 Grade 2 Class 1 (3/4Ni-
 
1/2Mo-1/3Cr-V) Low Alloy Steel (LAS) and stainless steel (18C r-8Ni) cladding materials.
Table 4-2 provides the Young's modulus (E), the mean coefficient of thermal expansion
 
(), and the yield strength (Sy) for the RV shell regions.
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 4-5  Table 4-2 Mechanical Properties of Turkey Point RV Shell Materials  RV Base Metal  Weld Metals Temp. E (ksi)  (in/in/°F)
Sy (ksi) SA-1101SA-1135  (°F) (ksi) (ksi)      70 27800 6.40E-06 50.0
[  ] [  ]  200 27100 6.70E-06 47.0
[  ] [  ]  300 26700 6.90E-06 45.5
[  ] [  ]  400 26200 7.10E-06 44.2
[  ] [  ]  500 25700 7.30E-06 43.2
[  ] [  ]  600 25100 7.40E-06 42.1
[ ] [ ]  For the Turkey Point reactor vessel shell and RV nozzle regions, the normal operating steady state condition cold leg temperature value is
[  ] (Table 5-1).
Both the Turkey Point Units have SA-1101 a nd SA-1135 Linde 80 welds in the RV shell regions. The yield strength values for these weld metals at
[  ], respectively.
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-1 
 
===5.0 FRACTURE===
MECHANICS ANALYSIS
 
===5.1 Methodology===
 
In accordance with ASME Section XI, Appendi x K [16], Subarticle K-1200, the following analytical procedure was used for Levels A & B Service Loads. a. The postulated flaws in the reactor vessel shell welds, the transition welds as well as RV nozzle-to-shell welds were postulated in accordance with the acceptance criteria of Subarticle K-2200. b. Loading conditions at the locations of the postulate d flaws were determined for Levels A and B Service Loadings. For Levels A and B Service loadings the equations to calculate the stress intens ity factor (SIF) du e to pressure and thermal gradients for a given pressure and cooldown rate are given in Article K-4210. Consistent with Section 5 of BAW
-2192PA [1], the accumulation pressure is taken as ten percent above the desi gn pressure and the maximum cooldown rate is 100°F/hr. In the area of the nozzle-to-shell weld, applied loadings consist of pressure, thermal, and attached piping reactions. c. Material properties, including E, ,  y , and the Jintegral resistance curve (JR  curve), were determined at the locations of the postulated flaws. Young's
 
modulus, mean coefficient of therma l expansion and yield strength are addressed in Section 4.2. The JR curve is discussed in Section 4.1.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-2 
: d. The postulated flaws were evaluated in accordance with the acceptance criteria of Article K-2000. Requirement s for evaluating the applied Jintegral are provided in Subarticle K-3200, and for determining fl aw stability in Subarticle K-3400.
Subarticle K-3500(a) invokes the procedur e provided in Subarticle K-4200 (K-4220) for evaluating the applied J-integral for a specified amount of ductile flaw extension. Three permissible evaluation methods to address flaw stability are described in Subarticle K-3500(b). The eval uation method selected herein is the J-R curve crack driving force diagram procedure described in Subarticle K-4310.
 
===5.2 Procedure===
for Evaluating Levels A and B Service Loadings
 
For RV shell regions remote from structural discontinuities, the applied J-integral is calculated in accordance wit h Appendix K, Subsubarti cle K-4210, using an effective flaw depth to account for small scale yielding at the crack tip, and evaluated per K-4220 for upper-shelf toughness and per K-4310 for flaw stability, as outlined below. The generic equations provided below are provided for both axially and circumferentially oriented flaws. Since all the Linde 80 welds for Turkey Point Units 3 and 4 are circumferential welds, only the equations for circumferentia l oriented flaws are applicable and used in the analysis.
 
(1) For an axial flaw of depth a, the stress intensity factor due to internal pressure is calculated wit h a structural factor (SF) on pressure using the following:
15.0)(1)(Fa t RpSFK i Ip  where  50.020.0,006.1982.0 2 1t a t a F AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-3 (2) For a circumferential flaw of depth a, the stress intensity factor due to internal pressure is calculat ed with a structur al factor (SF) on pressure using the following:
25.0)(21)(Fa t RpSFK i Ip where  50.020.0,345.0233.0885.0 2 2t a t a t a F (3) For an axial or circumferential flaw of depth a , the stress intensity factor due to radial thermal gradients is calculated usi ng the following:
F/h r100)(0, o 35.2CRFtCRCKmIt  where for SA-508, Class 2 or SA-533, Grade B, Class1 steels the material coefficient C m is defined as:
,0051.0 1d E C m  CR = cooldown rate ( F/hr), and 50.020.0,6046.0273.15353.01181.0 3 2 3t a t a t a t a F  (4) The effective flaw depth for small scale yielding, a e, is calculated using the following:
2 6 1yItIp eKKaa AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-4 (5) For an axial flaw of depth a e , the stress intensity factor due to internal pressure is:
'15.0')(1)(Fa t RpSFK e i Ip  where  50.020.0,006.1982.0 2'1t a t a F e e  (6) For a circumferential flaw of depth a e , the stress intensity factor due to internal pressure is:
'25.0')(21)(Fa t RpSFK e i Ip  where  50.020.0,345.0233.0885.0 2'2t a t a t a F e e e (7) For an axial or circumferential flaw of depth a e , the stress intensity factor due to radial thermal gradients is:
  ,'F/hr100)(0o '35.2CRFCRCK tmIt  where  50.020.0,6046.0273.15353.01181.0 3 2'3t a t a t a t a F e e e e AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-5 (8) The J-integral due to applied loads for sm all scale yielding is calculated using the following:
'2''1 1000 EKK JItIp   
 
where  2'1E E  (9) Evaluation of upper-shelf toughness at a flaw extension of 0.10 in. is performed for a flaw depth,  in.,10.025.0ta  using SF = 1.15  p = P a  where P a is the accumulation pressure for Levels A and B Service Loadings, such that J 1  J 0.1  where J 1 = the applied J-integral for a safety fact or of 1.15 on pressure,  and a safety factor of 1.0 on thermal loading
 
J 0.1 = the lower bound J-integral resistance at a ductile flaw extension of 0.10 inches AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-6 (10) Evaluation of flaw stability is performed through use of a crack driving force diagram procedure, by com paring the slopes of the applied J-integral curve and the lower bound J-R curve. The applied J-integral is calculated for a series of flaw depths corres ponding to increasing am ounts of ductile flaw extension. The applied pressure is the accumulation pressure for Levels A and B Service Loadings, P a , and the safety factor (SF) on pressure is 1.25. Flaw stability at a given applied load is verified when the slope of the applied J-integral curve is less than the slope of the J-R curve at the point on the J-R curve where the two curves intersect.
For the Turkey Point reactor vessels, the applied J-integrals at the nozzle-to-shell welds and the upper transition welds were determined using stresses from a detailed three-dimensional finite element analysis. Path line stresses were used to determine applied J-integrals that included a plastic zone correct ion to account for small scale yielding.
Based on the results of analysis performed for similar B&W fabricated reactor vessels it was deemed that the effects of structural discontinuities at the lower transition welds need not be explicitly addressed. 
 
===5.3 Evaluation===
for Flaw Extension The applied J-integrals for the RV shell welds, the RV transition welds, and the RV nozzle welds are calculated as discussed in Section 5.2.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-7 
 
====5.3.1 Reactor====
Vessel Shell Welds The basic reactor vessel shell geometry and design pressure al ong with operating condition temperature information for the Turk ey Point reactor vessels is provided in Table 5-1. Initial flaw depths equal to 1/4 of the vessel wall thickness are analyzed for Levels A and B service loadings following the procedure outlined in Section 5.2 and evaluated for acceptance based on values for the J-integral resistance of the material
 
from the Linde 80 J-R model discu ssed in Section 4.1. Calculations are initially carried out to identify the controlling weld such that subsequent detailed low upper shelf toughness flaw evaluations can be performed using the controlling weld.
The results of the evaluations for each of the RV shell welds are presented in Table 5-2.
From the results of the eval uation in Table 5-2, the cont rolling RV shell welds can be observed. The controlling welds are determined by noting the minimum ratio of the material J-resistance (J 0.1) to the applied J-integral (J
: 1) (also referred to as "margin").
For the Turkey Point reactor vessels, the controlling RV shell weld is the circumferential weld SA-1101 which is located between t he intermediate and lower shell courses of both Units 3 and 4 reactor vessels. The mini mum ratio of materi al J-resistance to applied J-integral (J 0.1/J 1) is [  ] , which is significantly higher than the minimum acceptable value of 1.0. 5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds The reactor vessel nozzle welds are located in the substantially thicker cylindrical shell section (reinforced to account for the in let/outlet RV nozzle openings and typically referred to as the nozzle belt), located abov e the reactor vessel shell welds. The reactor vessel nozzle belt dimensi ons are reported in Table 5-3.  .
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-8 For the Turkey Point reactor vessels upper RV transition and RV nozzle weld regions, the J-applied results with the safety factor of 1.15 on applied pressure is compared against the lower bound J-integral resistance at a ductile flaw extension of 0.1 inches (J 0.1) in Table 5-4. The limiting item in this case is the inlet nozzle with a margin of
[  ], which is significantly higher than the minimum acceptable value of 1.0. For the lower transition weld SA-1135, it is noted that this circumferential weld is located at the thickness transition between the cylindrical shell and the thinner lower head. This location was additionally evaluated using the di mensions applicable to the lower head
[  ] since this will result in higher pressure stresses but lower thermal stresses when compared against the thi cker RV cylindrical shell. For evaluation of the SA-1135 weld, the margin reduces from
[  ] (considering the thicker RV shell as shown in Table 5-2) to
[  ] (as given in Table 5-5) when considering the thinner lower head. Since the reduced margin is still significantly larger than 1.0, this simplified analytical approach is deemed to be an acceptable means of addressing the structural discontinuity at the lower transition weld.
 
===5.4 Evaluation===
for Flaw Stability The flaw stability analysis is performed by ca lculating the applied J-integrals for various amounts of flaw extension with a safety fact or (on pressure) of 1.25. The resulting applied J-integral curve can then be compared against the lower bound J-R curve for the weld metal. It is noted that applied J-integrals are also calculated with a safety factor on pressure of 1.15 for illustration of the J 0.1/J 1 margin with respect to the lower bound J-R curve at a flaw extension of 0.1 inch.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-9 
 
====5.4.1 Reactor====
Vessel Shell Welds For the controlling SA-1101 weld material of Turkey Point Units 3 and 4, the applied J-integral values are calculated and shown in Table 5-6 with the corresponding mean and lower bound J-R curve values shown in Table 5-7.
The resulting J-applied curves are then compared against the lower bound J-R curve for this material in Figure 5-1. An evaluation line at a flaw extensio n of 0.1 inch is included to confirm the results of Table 5-2 by showing the margin between the applied J-integral with the safe ty factor of 1.15 and the lower bound J-integral resistance of t he material. The requirement for ductile and stable crack growth is dem onstrated by Figure 5-1 since the slope of the applied J-integral curve for a safety factor of 1.25 is less than slope of the lower bound J-R curve at the point where the two curves intersect. 5.4.2 Reactor Vessel Transition Welds and RV Nozzle Welds As discussed in Section 5.3.2 and shown in Table 5-4, the limiting location for the Turkey Point reactor vessels is the inlet nozzle with a margin of
[  ]. The applied J-integral for the inlet nozzle with a safety factor of 1.25 on pressure at various flaw extensions is plotted with t he lower bound J-resistance curve in Figure 5-2. The slope of the applied J-integral is less than the slope of the lower bound J-resistance curve at the point of intersection, which demonstrates that the flaw is stable as required by ASME Section XI, Appendix K. 
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-10 Table 5-1  Reactor Vessel Shell Dimens ions and Operating Conditions AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-11 Table 5-2 Plant Specific Flaw Evaluation Summary for RV Shell Regions AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-12 Table 5-3 Reactor Vessel Nozzle Belt Dimensions
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-13 Table 5-4  Flaw Evaluation Summary of Turkey Point Upper Transition and RV Nozzle-to-Shell Welds Table 5-5  Flaw Evaluation Summary of Turkey Point Lower Transition Welds AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-14 Table 5-6  Applied J-Integral versus Flaw Extensions of Turkey Point Controlling RV Shell Weld (SA-1101)
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 5-15 Table 5-7  Mean & Lower Bound J-R Curve Values for Turkey Point Controlling RV Shell Weld  (SA-1101)
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report  Page 5-16 Figure 5-1  J-Integral Flaw Extension for Turkey Point Controlling Reactor Vessel Shell weld (SA-1101)
 
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report  Page 5-17 Figure 5-2  J-Integral versus Flaw Extension for Turkey Point Inlet Nozzle
 
AREVA Inc. ANP-3646NP Revision0Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report  Page 6-1 6.0
 
==SUMMARY==
AND CONCLUSIONS
 
===6.1 Reactor===
Vessel Shell Welds The ASME Section XI, acceptance criteria for Levels A & B Service Loads for all reactor vessel shell welds are satisfied. The results of the limiting welds for Turkey Point Units 3 and 4 are reported below. The limiting RV shell weld is TP 3 and 4 ci rcumferential weld SA-1 101. With factors of safety of 1.15 on pressure and 1.0 on t hermal loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in.(J0.1)-(Figure 5-1). The ratio J0.1/J1 =
[ ] is significantly greater than the required value of 1.0. With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flawextensions are ductile and st able since the slope of the appl ied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect (Figure 5-1).
 
===6.2 Reactor===
Vessel Transition Welds and RV Nozzle Welds The acceptance criteria for Levels A & B Service Loads for RV transition welds and RV nozzle welds are satisfied. The results of the limiting weld considering transition welds and RV nozzle welds (inlet and outlet) fo r Turkey Point are reported below. The limiting weld for the Turkey Point Units 3 and 4 RV transition welds (upper andlower) and the RV nozzle welds is the RV inle t nozzle-to-shell weld. With factors ofsafety of 1.15 on pressure and 1.0 on thermal loading, the applied J-integral (J1) isless than the J-integral of the material at a ductile fl aw extension of 0.10 in. (J0.1)(Figure 5-2). The ratio J0.1/J1 =
[ ] is significantly great er than the required value of 1.0.
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report  Page 6-2    With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and st able since the slope of the appl ied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect (Figure 5-2).
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ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 7-1 
 
==7.0 REFERENCES==
: 1. BAW-2192PA, Revision 00,  "Low Upper Shelf Toughness Fracture Analysis of Reactor Vessels of B&W Owners Group Reactor Vessel Working Group for Levels A and B Conditions," April 1994, ADAMS
 
Accession (Legacy)  9406240261 (P), 9312220294 (NP) 2. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components
: 3. BAW-2192P, Supplement 1, Revision 00,  "Low Upper Shelf Toughness Fracture Analysis of Reactor Ve ssels of B&W Owners Group Reactor Vessel Working Group for Levels A and B Conditions, December 2017
: 4. Code of Federal Regulations, Title 10, Part 50 - Domestic Licensing of Production and Utilization Facilities, Appendix G - Fracture Toughness Requirements, Federal Register Vo
: l. 60. No. 243, December 19, 1995.
: 5. BAW-2251A, "Demonstrat ion of the Management of Aging Effects for the Reactor Vessel, The B&W Owners Group Generic License Renewal
 
Program," August 1999, ADAM S Accession Number 9909300150 6. Not used 7. Not used
: 8. Not used
: 9. Not used
: 10. Not used
: 11. TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING EXTENDED POWER UPRATE (TAC NOS. ME4907 AND ME4908)," Adams Accession number ML11293A365 AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 7-2 
: 12. LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE, ATTACHMENT 4, L-2010-113, Atta chment 4  ADAMS --ML103560177 13. Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA), L-2010-303, ADAMS-ML103610321 14. Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1, L-2011-029, ADAMS--ML110700068 15. NRC Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels With Charpy Upper Shelf Energy Less Than 50 ft-lb  16. 2007 Edition (with 2008 Addenda) ASME & Boiler Pressure Vessel Code, Section XI, Rules for Inservice In spection of Nuclear power Plant Components, Appendix K 17. 2013 Edition ASME & Boil er Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear power Plant Components, Appendix K 18. BAW-1975, Applicability of t he HSST Program Second and Third Irradiation Series Data to the Integr ity of Nuclear Reactor Vessels, A.L.
Lowe, June 1987 19. Not used 20. Not used
: 21. BAW-1543, Revision 4, Supplemen t 6-A, Supplement to the Master Integrated Reactor Vessel Surveillance Program, June 2007 22. ANP-3647P-000, Low Upper-She lf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels C & D Service Loads at 80-Years, January 2018 AREVA Inc. ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page 8-1 8.0 CERTIFICATION This report is an accurate description of the low upper-shelf toughness fracture analysis of Turkey Point Units 3 and 4 reactor vessels. d'cJ..J{tc~JIQ/8 Mark/CR' ckel 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 Turkey Point Units 3 and 4. Verification of independent review. This report is approved for release. A-z~ 1/;s/18 Ashok D. Nana .... t-Component Analysis and Fracture Mechanics Unit . ~-7 David R. Cofflin Component Anal Mechanics Unit ~~fwr Danielle Page I NSSS Project Management AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-1  APPENDIX A B&WOG J-R MODEL-DATA ANALYSI S AND EMPIRICAL MODEL DEVELOPMENT A.1 Background AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-2 AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-3  Table A-1 Model 4B, Range of Test Data AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-4  Figure A-1 BAW-2251A, Appendix B, Figure 3-1 
 
AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-5  A.2 New B&WOG J-a Data and Comparison to B&WOG J-R Model AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-6 AREVA Inc.
ANP-3646NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-7  Figure A-2 Jd (0.1) vs Fluence B&WOG J-R Model 4B and New Test Data (Normalized to Standard Conditions)
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ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-8 A.3  New B&WOG J-R Model
 
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ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-9 
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-10 Table A-3 Jd Model Coefficients (Models 4B, 5B, and 6B)
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-11 Figure A-3 Original and New Data and Model Fit Normalized at Standardized Conditions vs a AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-12 Figure A-4 Original and New Data and Model Fit Normalized at Standard Conditions vs Fluence
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-13 Figure A-5 Model 6B Residuals vs Fitted Values
 
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ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-14 Figure A-6 Model 6B Standardized Residuals vs Fitted Values
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-15 Figure A-7 Normal Q-Q Plot of Standardized Residuals
 
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ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-16 A.4 Model 6B Reconciliation to EMA Results Presented in Section 6.0 AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-17 Figure A-8 Comparison of Models 4B, 5B, and 6B at Standard Conditions 
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-18 Table A-4 EMA Reconciliation for Limiting RV Shell Welds-Models 4B and 6B
 
AREVA Inc.
ANP-3646NP  Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B Service Loads at 80-Years Topical Report Page A-19 A.5  Summary and Conclusions Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 3Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
Areva Topical Report, ANP-3647NP Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels C & D Service Loads at 80 Years January 5, 2018 (Non-proprietary)
(53 Total Pages, including cover sheets)
 
ANP-3647NP Revision 0
 
Low Upper-Shelf Toughness Fracture
 
Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report
 
January 2018 AREVA Inc.
(c) 2018 AREVA Inc.
ANP-3647NP Revision 0
 
Copyright © 2018 AREVA Inc.
All Rights Reserved AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page i    Nature of Changes Item Section(s) or Page(s) Description and Justification 1 All Initial Issue
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page ii    Contents Page
 
==1.0 INTRODUCTION==
............................................................................................... 1-1
 
===1.1 Equivalent===
Margins Analysis-A nalysis of Record .................................. 1-3
 
===2.0 REGULATORY===
REQUIREMENTS .................................................................... 2-1
 
===2.1 Regulatory===
Requ irements ....................................................................... 2-1
 
===2.2 Compliance===
with 10 CFR 50 Ap pendix G and Acceptance Criteria .................................................................................................... 2-2
 
====2.2.1 Acceptance====
Criteria Levels C and D ............................................ 2-3
 
==3.0 DESCRIPTION==
OF TURKEY P OINT REACTOR VESSELS ............................. 3-1
 
===4.0 MATERIAL===
PROPERTIES AND LEVELS C&D SERVICE
 
LOADINGS ........................................................................................................ 4-1 4.1 J-Integral Resist ance Model ................................................................... 4-1
 
===4.2 Mechanical===
Properties of Weld Metals .................................................... 4-3
 
====4.2.1 Mechanical====
Properties for the Turkey Point Reactor Vessels ........................................................................................ 4-4
 
===4.3 Levels===
C and D Serv ice Loadings ........................................................... 4-5
 
====4.3.1 Turkey====
Point ................................................................................. 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 S hell 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 Weld s and RV Nozzle Welds ........................ 6-2
 
==7.0 REFERENCES==
.................................................................................................. 7-1
 
===8.0 CERTIFICATION===
............................................................................................... 8-1 AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page iii List of Tables Table 3-1 Reactor Vessel Weld Locations
--Copper Content and  80-Year Fluence Projections ................................................................................... 3-2 Table 3-2 Reactor Vessel Shell Dimensions
.............................................................. 3-3 Table 3-3 Reactor Vessel Nozzle Be lt Dimensions ................................................... 3-4 Table 4-1 Parameters in Jd Model 4B ....................................................................... 4-3 Table 4-2 Mechanical Properties of Turkey Point RV Materials
................................. 4-4 Table 5-1  Turkey Point J-Integral ve rsus Flaw Extension for Levels C & D Service Loadings ..................................................................................... 5-12 Table 5-2  Turkey Point-J-R Curves for Evaluation of Levels C & D Service Loadings .................................................................................................. 5-13 Table 5-3  Turkey Point-Levels C & D Results for Nozzle to Shell and Upper Transition Welds ...................................................................................... 5-14
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page iv    List of Figures Figure 3-1  Reactor Vessel-Tu rkey Point Unit 3 ...................................................... 3-5 Figure 3-2  Reactor Vessel-Tu rkey Point Unit 4 ...................................................... 3-6 Figure 4-1  Turkey Point Steam Line Break Transients ............................................. 4-6 Figure 5-1  Turkey Point-KI versus Cr ack Tip Temperature for Levels C & D Service Loadings .................................................................................. 5-15 Figure 5-2 Turkey Point-J-Integral versus Flaw Extension for Levels C & D Service Loadings .................................................................................. 5-16 Figure 5-3  Turkey Point- Levels C
& D Applied J Integral vs Crack Tip Temperature for the Outlet Nozzle to She ll Weld .................................. 5-17 Figure 5-4  Turkey Point- Levels C & D Applied J Integral vs Crack Extension for the Outlet Nozzle to Shell We ld ....................................................... 5-18
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page v    Nomenclature 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 PTN Turkey Point Plant RV Reactor Vessel RVWG Reactor Vessel Working Group SLR Subsequent License Renewal Sy Yield Strength TSs Technical Specifications USE Upper Shelf Energy AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page vi    ABSTRACT This topical report presents the result s 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 fluence va lues expected at 80-years (subsequent license renewal--SLR). This topical report applies to the following Westinghouse-
 
designed reactor vessels fabricated by B&W:  Turkey Point Plant (PTN) Units 3 and 4.
Note that the Turkey Point EMA reported herei n is technically identical to the Turkey Point EMA reported in BAW-2178P, Supplement 1, Revision 0, which was submitted to the NRC by the PWROG on Decem ber 15, 2017. That is, Sect ions 1.0 through 7.0 of ANP-3647P were generated by extracting Turk ey Point 3 and 4-specific results from Sections 1.0 through 7.0 of BAW-2178, Supp lement 1, Revision 0. The B&WOG J-integral resistance model is discussed in  Appendix A to ANP-3646P, which is identical to Appendix A of BAW-2192, Supplement 1, Revision 0, with the exception that references to plants other than Turkey Poin t 3 and 4 were removed from Sections A.1, A.2, and A.4. The analytical procedure used in this supplement is in accordance with ASME Section XI, Appendix K, Subarticle K-1200, with se lection of design transients based on the guidance in Regulatory Guide 1.161, Section 4.0. EMA re sults are reported for all reactor vessel weld locations with 80-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 weld s are satisfied. The acceptance criteria for Levels C & D Service Loads for RV transition welds and RV nozzle welds are also satisfied. Consistent wit h BAW-2178PA, Revision 0, B&W OG J-R Model 4B is used for the Linde 80 welds.
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page vii The EMA utilizes the B&WOG J-integral re sistance (J-R) Model 4B reported in BAW-2192PA, Appendix B. Model 4B was develo ped based on fracture toughness test data obtained through approximately 1990, with specimen fluence that ranges from 0.0 to 8.45E+18 n/cm
: 2. Eighty-year fluence estimates for Turkey Point Units 3 and 4 exceeds 8.45E+18 n/cm 2 (e.g., maximum 80-year 1/4 T fluence at a crack extension of 0.1 inches is estimated at 6.38E+19 n/cm
: 2) and use of Model 4B 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 4B is compared to new fracture toughness test dat a (1990 to 2017) irradiated to fluence ranging from 8.0E+18 n/cm 2 to 5.8E+19 n/cm
: 2. The majority of te st data fell above the Model 4B mean and all of the test data fe ll above the Model 4B mean minus 2 standard error band. Therefore, use of Model 4B and associated uncertainty to extrapolate J-integral resistance for 80-year fluence appl ications was determined to be appropriate. This assessment is reported in ANP-3646P, Rev. 0, Appendix A.
To further substantiate the use of Model 4B, all of the original fracture toughness data used to develop Model 4B was combined with new fracture toughness data, using the
 
same model form, and a new Model 6B was generated. Model 6B was found to be essentially equivalent to Model 4B with res pect to model mean and 2 standard errors.
The EMA results reported herein using Model 4B were reconciled to Model 6B, with little or no change to the EMA results. Model 6B development and the EM A reconciliation to Model 4B are reported in ANP-3646P, Rev. 0, Appendix A.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 va lues expected at 80-years (subsequent license renewal--SLR). This topical report applies to the following Westinghouse-
 
designed reactor vessels fabricated by B&W:  Turkey Point Plant (PTN) Units 3 and 4 (also referred to as Turkey Point Units 3 and 4). Note that the Turkey Point EMA
 
reported herein is technically identical to the Turkey Point EMA reported in BAW-2178P, Supplement 1, Revision 0 [4], which was submitted to the NRC by the PWROG on December 15, 2017. That is, Sections 1.
0 through 7.0 of ANP-3647P were generated by extracting Turkey Point 3 and 4-specific results from Sections 1.0 through 7.0 of BAW-2178, Supplement 1, Revision 0. T he B&WOG J-integral resistance model is discussed in  Appendix A to ANP-3646P, which is identical to Appendix A of BAW-2192, Supplement 1, Revision 0 [6], with the exc eption that references to plants other than Turkey Point 3 and 4 were removed from Sections A.1, A.2, and A.4.
The EMA reported herein util izes the B&WOG J-integral resistance (J-R) Model 4B reported in BAW-2192PA, Appendix B. Justification for use of Model 4B for 80-year fluence is addressed in ANP-364 6P, Rev. 0, Appendix A [3].
Equivalent margins analyses ar e reported for all reactor vessel weld locations with 80-year fluence projections that exceed 1.0 E+17 n/cm 2 (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/cm 2, if needed, will be addressed separ ately in the Turkey Point Units 3 and 4 subsequent licens e renewal application.
 
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 welds for Turkey Point Units 3 and 4 (also referred to as Turkey Point 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. Transition Welds and RV Nozzle Welds-welds that are located above and below the reactor vessel shell welds that may experience 80-year fluence greater than 1.0 E+17 n/cm 2 and must consider the effects of neut ron irradiation em brittlement. 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.
The EMA evaluations in this report are for a ll weld locations expected to receive fluence
> 1.0E+17 n/cm 2 [19] at 80 years. The use of the terms "beltli ne" and/or "extended beltline" are not used in this report.
The 60-year EMA summary reports for Turkey Point Units 3 and 4 are reported in Section 1.1. Section 2.0 provides the current NRC regulatory requirements for the EMA. Section 3.0 provides a description of 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 refe rences. ANP-3646P, Rev. 0, Appendix A [3]
provides the technical justification for t he use of B&WOG J-R Model 4B for the EMA
 
reported herein.
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 1-3 
 
===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  Turkey Point Units 3 and 4 for 40 years. For 60 years, Turkey Point Units 3 and 4 reported plant-specific evaluations. T he summary reports for EMA analyses of record are as follows. Turkey Point Units 3 and 4 The Turkey Point Units 3 and 4 current licensing basis equivalent margins analysis at 48 EFPY is summarized in Section 2.1.2 of NRC document "TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING EXTENDED POWER UPRATE (TAC NOS. ME4907 AND ME4908)," Adam s Accession number ML 11293A365 [12]. NRC acceptance of the Turkey Point EMA at 48 EFPY for EPU is based on the following documentation. LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE, ATTACHMENT 4, L-2010-113, Attach ment 4  ADAMS --ML103560177 [13]  Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) Licens e Amendment Request (LAR) No. 205 and Equivalent Margin An alysis (EMA), L-2010-303, ADAMS-ML103610321 [14].
Reference [14] contains AREVA docum ent 77-2312-03 (P), LOW UPPER-SHELF TOUGHNESS FRACTURE MECHANICS ANALYSIS OF REACTOR VESSELS OF TURKEY POINT UNITS 3 AND 4 FOR EXTENDED LIFE THROUGH 48 EFFECTIVE FULL POWER YEARS AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 1-4 Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1, L-2011-029, ADAMS--ML110700068 [15]. Reference [15] provides a response to RAI CVIB-1.2 regarding the code year used to perform the equivalent margins analysis (i.e., 1998 Edition vs 2004 Edition). The NRC SER for the Turkey Point Extended Power Uprate, Section 2.1.2, contains the evaluation of upper shelf energy--ADAMS-ML 11293A365. 
 
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  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, Reacto r 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 R egulation or Director, Office of New Reactors, as appropriate, that lower va lues 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 anal ysis 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 t he beltline materials after exposure to neutron irradiation may be obtained from result s 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 y ears prior to the dat e 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, Of fice of New Reactors, as appropriate.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  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 f abricated, Charpy V-notch tests of the reactor vessel weld s 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. Turkey Point-ASME B&PV Code, Sect ion III, 1965 Edition through Summer 1966 Addenda In accordance with NRC Regulatory Guide 1.161 [16], the NRC has determined that the analytical methods described in ASME Sect ion 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. Selectio n of design transients and selection of material properties are addressed in Sections 3.0 and 4.0. The Linde 80 weld locations t hat are included within the scope of this report (i.e., weld locations with 80-year projected fluence > 1.0E+17 n/cm
: 2) 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 t he 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 ar e 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.
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 2-3 The material properties used in this anal ysis are based on ASME Section II, Part D, 2007 Edition with 2008 Addenda. The only change in the material properties listed in the 2013 Edition of ASME Section II for the applicable properties is the coefficient of thermal expansion for stainless steel at 600° F; this value was changed from 9.8E-6 in/in/°F to 9.9E-6 in/in/°F. 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 Crit eria Levels C and D ASME Section XI [17], Subarticles K-2300 and K-2400, provide acceptance criteria for Levels C and D Service Conditions. Consis tent with BAW-2178PA [1
], the evaluations reported herein will utilize acc eptance criteria applicable to Level C Service Loadings as summarized below. a) When evaluating adequacy of the upper s helf toughness for the weld material for Level C Service Loadings, interior semielliptical surface flaws with depths up to 1/10 th  of the base metal wall thickness, plus the cladding thickness, with total depths not exceeding 1 in. (25 mm), and a su rface 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 mate rial, both interior axial and circumferential flaws shall be post ulated, and toughness properties for the corresponding orientation shall be used. Flaws of various depths, ranging up to the maximum postulated depth, shall be an alyzed to determine the most limiting flaw depth. Smaller maximum flaw sizes may be used when justif ied. Two criteria shall be satisfied:
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 2-4 
: 1. The applied Jintegral shall be less than the Jintegral of the material at a ductile flaw extension of 0.10 in. (2.5 mm
), using a structural factor of 1 on loading. 2. Flaw extensions shall be ductile and st able, 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 li gament shall not be subject to tensile
 
instability.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-1 
 
==3.0 DESCRIPTION==
OF TURKEY POINT REACTOR VESSELS The Turkey Point reactor vessels with applic able weld locations are shown in Figure 3-1 and Figure 3-2. All weld locations evaluat ed for equivalent margins in this report are identified by an asterisk (*) in each figure. The Linde 80 welds copper and nickel content and 80-year fluence projections data needed for the equivalent margins analysis are provided in Table 3-1. The fl uence projections are reported for all reactor vessel weld locations that are expected to exceed 1.0E+17 n/cm 2 at 80 years. The 80-year fluence projections are conservati ve estimates based on detailed transport calculations completed by Westinghouse Elec tric Corporation usi ng a methodology that is in compliance with Regulatory Guide 1.190 (i.e., Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldo wn Limit Curves," May 2004.)
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 c ontent 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 Turkey Point reactor vessel fabrication reports.
The dimensions of the reactor vessel shel l geometry for the Turkey Point 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 are given in Table 3-3.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-2 Table 3-1 Reactor Vessel Weld Locations--Copper Content and  80-Year Fluence Projections Reactor Vessel Material Material ID and/or Heat Number  Cu, wt% Ni, wt% (IS) Inside Wetted Surface Fluence or (*) clad/base metal n/cm 2 E> 1.0 MeV)
Turkey Point Unit 3, 80-Year Fluence (E > 1.0 MeV)
US Forging to INF Welds Heat 8T17620.190.57(IS) 1.50E+18 Heat 8T1554B0.160.57(IS) 1.50E+18 Heat 712490.230.59(IS) 1.50E+18 US Forging to ONF Welds Heat 8T17620.190.57(IS) 1.50E+18 US to IS Circ. Weld SA-1484 Heat 72442 0.26 0.60 (*) 1.19E+19 IS to LS Circ. Weld SA-1101 Heat 71249 0.23 0.59 (*) 1.04E+20 LS to Dutchman Circ. Weld SA-1135 Heat 61782 0.23 0.52 (IS) 1.5E+18 Turkey Point Unit 4, 80-Year Fluence (E > 1.0 MeV)
US to INF Welds Heat 8T1762 0.19 0.57 (IS) 1.50E+18 Heat 8T1554B 0.16 0.57 (IS) 1.50E+18 Heat 299L44 0.34 0.68 (IS) 1.50E+18 US to ONF Welds Heat 8T1554B 0.16 0.57 (IS) 1.50E+18 Heat 299L44 0.34 0.68 (IS) 1.50E+18 US to IS Circ. Weld (ID 67%) WF-67 Heat 72442 0.26 0.60 (*) 1.21E+19 IS to LS Circ. Weld SA-1101 Heat 71249 0.23 0.59 (*) 1.03E+20 LS to Dutchman Circ. Weld SA-1135 Heat 61782 0.23 0.52 (IS) 1.5E+18
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-3 Table 3-2 Reactor Vessel Shell Dimensions
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-4 Table 3-3 Reactor Vessel Nozzle Belt Dimensions
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-5 Figure 3-1  Reactor Vessel-Turkey Point Unit 3
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 3-6 Figure 3-2  Reactor Vessel-Turkey Point Unit 4
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report  Page 4-1  4.0 MATERIAL PROPERTIES AND LEVELS C&D SERVICE LOADINGS 4.1 J-Integral 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 wa s 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 ANP-3646P, Rev. 0 for a disc ussion of the extension of the range of applicability of the B&WOG J-R model to fluence values expected at 80 years for Turkey Point reactor vessels.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report  Page 4-2  The coefficients, a, d, and C4 are provided in Table 4-1.
As required by ASME Section XI, ASME K-3300, when evaluati ng the vessel for Levels A, B, and C Service Loadings, the Jintegral resistance versus crackextension curve (JR curve) shall be a conservative representation of the toughness of the controlling beltline material at upper shelf temperatures in the operating range. When evaluat ing the vessel for Level D Service Loadings, the JR curve shall be a best estimate representation of the toughness of the controlling beltline materi al 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 Loadi ngs (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 JR curve was generated from a Jintegral 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 duct ile tearing with no cleavage. This complies with ASME Section XI, K-3300.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report  Page 4-3  Table 4-1 Parameters in Jd Model 4B
 
===4.2 Mechanical===
Properties of Weld Metals The following subsections provide representative properties for the Turkey Point reactor vessels. The temperature dependent mechanica l properties are developed from the 2007 Edition with 2008 Addenda of the ASME Code (Section III) 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 limit ing time points in the Levels C & D analysis where cladding effects are included the temper ature of the cladding is well below 600°F, and thus this change does not impact the EMA analyses summarized herein.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 properti es but in some cases the ASME Code minimum properties were conservatively considered. The irradiated material properti es 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 Turkey Point Reactor Vessels The Turkey Point reactor vessels are fabricated using A-508 Grade 2 Class 1 (3/4Ni-
 
1/2Mo-1/3Cr-V) Low Alloy Steel (LAS) and stainless steel (18C r-8Ni) cladding materials.
Table 4-2 provides the Young's modulus (E), the mean coefficient of thermal expansion
 
(), and the yield strength (Sy) for the RV base metal and weld material and the E and  properties for the RV cladding material.
Table 4-2  Mechanical Properties of Turkey Point RV Materials RV Base Metal Weld Metals Cladding Temp. E (ksi)  (in/in/°F)
Sy (ksi)SA-1101SA-1135 E (ksi)  (in/in/°F) (°F) (ksi)(ksi)        70 27800 6.40E-06 50.0 [  ]  [  ]
28300 8.50E-06 200 27100 6.70E-06 47.0 [  ]  [  ]
27500 8.90E-06 300 26700 6.90E-06 45.5 [  ]  [  ]
27000 9.20E-06 400 26200 7.10E-06 44.2 [  ]  [  ]
26400 9.50E-06 500 25700 7.30E-06 43.2 [  ]  [  ]
25900 9.70E-06 600 25100 7.40E-06 42.1 [ ]  [ ]
25300 9.80E-06 AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  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 Turkey Point Levels C and D Service Loadings were developed for the Turkey Point reactor vessels for the two steam line break transients listed below. The Turkey Po int Safety Injection (SI) pump shutoff head will limit RCS pressure increase to 1660 psi.
Level D: Steam Line Break (SLB Without Offsite Power)  Steam Line Break (SSDC 1.3 SLB)
The pressure and temperature steam line break transients for the Turkey Point reactor vessels are depicted in Figure 4-1. 
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report  Page 4-6  Figure 4-1  Turkey Point St eam Line Break Transients AREVA Inc.
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===5.0 FRACTURE===
MECHANICS ANALYSIS
 
===5.1 Methodology===
 
In accordance with ASME Section XI, Appendi x 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 accept ance criteria of Subarticles K-2300 and K-2400. b) Loading conditions at the locations of the postulat ed flaws were determined for Levels C and D Service Loadings. c) Material properties, including E, ,  y , and the Jintegral resistance curve (JR  curve), were determined at the locations of the postulated flaws. Young's modulus, mean coefficient of therma l expansion and yield strength are addressed in section 4.2. The JR curve is discussed in section 4.1. d) The postulated flaws were evaluated in accordance with the acceptance criteria of Article K-2000 by calculat ing the applied J-integral a ccording 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.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 se rvice loadings is performed as follows: 1 For each transient described in Section 4.3, calculate stress inte nsity factors for a semi-elliptical flaw of depth up to 1/10 th of the base metal wall thickness, as a function of time, due to internal pre ssure and radial thermal gradients with a factor of safety of 1.0 on loading. The applied stress intensity factor, K I , calculated for each of these transients is compared to the K Jc upper-shelf toughness curve of the weld material. The transient for which the applied K I most closely approaches the K Jc curve is chosen as the limiting transient, and the critical time in the limiting transient se lected for further eval uation occurs at the point where K I most closely approaches the K Jc curve. 2 At the critical transient time, devel op a crack driving force diagram with the applied J-integral and J-R curves plotted as a functi on 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 t he 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 li gament is not subject to tensil e instability. The internal pressure p shall be less than P I , where P I is the internal pressure at tensile instability of the remaining ligamen
: t. The pressure at instability, P I , is given in  K-5300, Appendix K of ASME Section XI for both axial and circumferential flaws.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-3 
 
====5.2.1 Processing====
of Transient Time-History Data 5.2.1.1 Turkey Point For the Turkey Point reactor vessels, the applie d 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 stre sses were subsequently used to calculate stress intensity factors and the applied J-integrals were determined based on consideration of small scale yielding. Fo r the controlling reactor vessel shell weld, stress intensity factors were calculated using the one-dim ensional, 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 re latively high constant energy level.
Similarly, fracture toughness can be addressed in three different regions on the temperature scale, i.e., a lower-shelf toughn ess region, a transition region, and an upper-shelf toughness region. Fracture toughness of reactor vessel steel and associated weld metals are conservative ly predicted by the ASME initiation toughness curve, K Ic, in the lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, K Jc , is derived from the upper-shelf J-integral resistance model described in Section 4.1. The upper-shel f 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 te mperature. Since the present analysis addresses the low upper-shelf fracture toughness issue, only the upper-shelf temperature range, which begins at the intersection of K Ic and the upper-shelf toughness curves, K Jc , is considered.
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====5.2.3 Cladding====
Effects AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-5 
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 Sections 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 tran sient events, is calculated and evaluated as discussed in Section 5.2.
The transition region toughness and upper s helf toughness are discussed in Section 5.2.2. Transition region toughness is obtained from the ASME Section XI equation for crack initiation, K Ic = 33.2 + 20.734 exp[0.02(T - RT ND T)]  using the applicable RTNDT value for a flaw depth of 1/10 th the wall thickness, where:
K Ic = transition region toughness, ksiin  T = crack tip temperature,  F Upper shelf toughness K Jc is derived from the J-integral resistance model of Section 4.1 for a flaw depth of 1/10 th the wall thickness, a crack extension of 0.10 inch, and the applicable fluence value at the crack tip: 
)(EJ K.Jc 21011000 where K Jc = upper-shelf region toughness, ksiin  J 0.1 = J-integral resistance at a = 0.1 in.
Using the above equations, the transiti on and upper shelf toughness values as a function of temperature are determined for the controlling weld and Levels C and D service loading conditions.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-8 
 
====5.3.1 Reactor====
Vessel Shell Welds 5.3.1.1 Turkey Point For the Turkey Point reactor vessels, the controlling weld was identified to be the  SA-1101 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 th  the base metal thickness for Levels C and D Service Loadings. The two main steam line break
 
transients (SSDC 1.3 SLB & SL B without Offsite Power) identified in Section 4.3.1 and illustrated in Figure 4-1 are evaluated.
The analysis is performed using the PCRIT Code. Figure 5-1shows the variation of t he applied stress intensity factor, K I , for these two transient cases as a function of the crack ti p temperature. This figure also shows the transition region toughness K Ic curve and the mean and lower bound upper-shelf toughness K Jc curves with crack tip temperature
. The K Ic curve is determined using the Adjusted Reference Temperature (ART) value at 1/10 th of the wall thickness for the limiting weld SA-1101, which at 80-years is
[  ] . The symbols on the K I 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 K Jc 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 K I curve is closest to the lower bound K Jc curve at 6.5 minutes into the transient. This time is selected as the critical time in the transi ent at which to perform the flaw evaluation for Levels C and D Service Loadings.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-9 The additional stress intensity fact or attributable to the cladding, K Iclad , at 6.5 minutes (limiting time point) into the SSD C SLB transient is determined to be
[  ] at a flaw depth corresponding to 1/10 th of the wall thickness. Applied J-Integrals are calculated for the controlling weld for various flaw depths in Table 5-1 using stress intensity factors from PCRIT for the SSDC SLB (at 6.5 min.) and adding
[  ] to account for cladding effects. Stre ss intensity factors are converted to J-integrals by the plane strain relationship, )1()(1000)(2 2 total appliedEaKaJ I Since the RV shell weld for the Turkey Point reactor vessels is
[  ] inch thick as given in Table 3-2, the initial flaw depth of 1/10 th  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 wi th 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 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 material J-resistance (J 0.1) to the applied J-integral (J
: 1) ratio can be determined using the lower bound and mean J-R curve values, from Table 5-2, for Levels C and D conditions, respectively. The margin for Level C Service Loading is
[  ] and the margin for Level D Service Loading is
[  ]  The requirements for ductile and stable cra ck 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.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-10 Referring to Figure 5-2, the Level D Service Loading requirement t hat 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 (less than 1%).
The last requirement for Level D Conditions is that the internal pressure p shall be less than P I , the internal pressure at tensile inst ability of the remaining ligament. The calculations for P I were determined for a circumferential flaw. An additional check is performed for circumferential flaws to ensure t hat internal pressure does not exceed the pressure at tensile instability caused by the applied hoop stress ac ting over the nominal wall thickness of the vessel. This valid ity limit on pressure is satisfied by i oyinstabilit R t07.1 P. To demonstrate that the remain ing ligament does not exceed the pressure at instability a conservative flaw depth equal to 1/10 th of the wall thickness plus 0.1 inch is used.
Although the internal pressure at tensile instability is ca lculated to be 12, 530 psi, the validity check on hoop stress requi res that the internal pre ssure not exceed 6240 psi, which is still much greater than any anticipated accident condition pressure. Therefore, the remaining ligament is not s ubject to tensile instability. 5.3.2 Reactor Vessel Transition Welds and RV Nozzle Welds The reactor vessel upper and lower transit ion welds are located above and below the reactor core, respectively (see Figures 3-1 and 3-2). The RV nozzle welds are located above the upper transition weld in the substant ially 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.
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-11 5.3.2.1 Turkey Point For the Turkey Point reactor vessels the appl ied 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 evaluat ed. For the SSDC transient the maximum pressure was capped at 1660 psi due to the s hutoff head of the safety injection pumps.
The bounding results with safety factor of 1.0 on the applie d 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 ti p temperature for each transi ent, the SSDC SLB transient and SLB without offsite power transient (also refe rred to as AIS transient), is plotted in Figure 5-3 for the outlet nozzle, along with the temperature dependent mean and lower
 
bound J 0.1 curves. As can be seen all points of the transient remain below the lower bound J 0.1. Additionally, Figure 5-3 shows the K Ic fracture toughness using an RT NDT of  [  ] , converted to an equivalent J using K Ic 2/(E/(1-2)); 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 vari ous flaw extensions is plotted with the lower bound J-resistance curve in Figure 5-4; the sl ope 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.
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-12 Table 5-1  Turkey Point J-Integral versus Flaw Extension for Levels C & D Service Loadings
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-13 Table 5-2  Turkey Point-J-R Cur ves for Evaluation of Levels C & D Service Loadings
 
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ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-14 Table 5-3  Turkey Point-Levels C & D Results for Nozzle to Shell and Upper Transition Welds
 
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-15 Figure 5-1  Turkey Point-KI vers us Crack Tip Temperature for Levels C & D Service Loadings AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-16 Figure 5-2  Turkey Point-J-Integral versus Flaw Extension for Levels C & D Service Loadings AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-17 Figure 5-3  Turkey Point- Levels C & D Applied J Integral vs Crack Tip Temperature for the Outl et Nozzle to Shell Weld
 
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 5-18 Figure 5-4  Turkey Point- Levels C
& D Applied J Integral vs Crack Extension for the Outlet Nozzle to Shell Weld
 
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 6-1 6.0
 
==SUMMARY==
AND CONCLUSIONS
 
===6.1 Reactor===
Vessel Shell Welds The ASME Section XI, acceptance criteria fo r 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 rati o 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 Turkey Point Units 3 and 4 are reported below.
The limiting weld among the Turkey Point reactor vessel shell welds is Turkey Point Units 3 and 4 circumferentia l weld SA-1101. 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 (SA-1101) is less than the lower bound J-integral of the material at a ductile flaw extension of 0.10 inch (J 0.1) with a ratio J0.1/J 1 =  [  ] =  [  ] , which is greater than the requ ired value of 1.0. Using the mean J-R curve permitted by Subarticle K-2400 for this Service Level D transient, the ratio J0.1/J 1 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-1101) 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-1101 it was demonstrated that flaw growth is stable at much less than 75% of the vessel wall thickness. It has al so been shown that the remaining ligament is sufficient to preclude tensile instability.
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 6-2 
 
===6.2 Reactor===
Vessel Transition Welds and RV Nozzle Welds The ASME Section XI, acceptance criteria fo r 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 acceptanc e 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 impo sed on the Level D transients evaluated in Section 5.0, although ASME Section XI, S ubarticle K-2400(b) permits use of a best estimate J-integral resistance curve for Lev el D Service Loadings.
The results of the limiting welds for Turkey Point Units 3 and 4 are reported below.
The upper transition weld and RV inlet and out let 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 (J
: 1) 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 J 0.1/J 1:  [  ] for the RV outlet nozzle-to-shell weld,  [  ] for the RV inlet nozzle-to-shell weld, and
[  ] for the upper transition we ld. All 3 ratios are greater than the required value of 1.0. 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 wa s 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 se ction of the reactor vessel is thicker compared to the RV shell welds, it is considered to be bounded by the RV shell location.
AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-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 B Conditions," April 1994, ADAMS
 
Accession (Legacy)  9406240261 (P), 9312220294 (NP) 3. ANP-3646P, Rev. 0, "Low Uppe r-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 Reactor Vessels for Levels A & B
 
Service Loads at 80-Years" (Proprietary),  January 2018
: 4. 43-2178P, Supplement 1, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reacto r Vessels of B&W Owners Reactor Vessel Working Group for Levels C & D Service Loads, December 2017 5. Code of Federal Regulations, Title 10, Part 50 - Domestic Licensing of Production and Utilization Facilities, Appendix G - Fracture Toughness Requirements, Federal Register Vo
: l. 60. No. 243, December 19, 1995
: 6. 43-2192P, Supplement 1, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reacto r Vessels of B&W Owners Reactor Vessel Working Group for Levels A & B Service Loads, December 2017 7. Not used 8. Not used
: 9. Not used
: 10. Not used
: 11. Not used AREVA Inc.
ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4  Reactor Vessels for Levels C & D Service Loads at 80-Years Topical Report Page 7-2 
: 12. TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING EXTENDED POWER UPRATE (TAC NOS. ME4907 AND ME4908)," Adams Accession number ML11293A365 13. LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE, ATTACHMENT 4, L-2010-113, Atta chment 4  ADAMS --ML103560177 14. Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA), L-2010-303, ADAMS-ML103610321 15. Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1, L-2011-029, ADAMS--ML110700068 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 Inservice In spection of Nuclear power Plant Components, Appendix K 18. 2013 Edition ASME & Boil er Pressure Vessel Code, Section XI, Rules for Inservice 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 AREVA Inc. ANP-3647NP Revision 0 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Turkey Point Units 3 and 4 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 Turkey Point Units 3 and 4 vessels. ~1f}XI'JJ 1k)8 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 Turkey Point Units 3 and 4. Verification of independent review. This report is approved for release. v4@.<Z~~ 1/.s/!8 Ashok D. Nana Component Analysis & Fracture Mechanics Unit David R. Cofflin Component Analy Mechanics Unit fMvn:i~,~
&;,rs Danielle Page Ir / NSSS Project Management Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 4Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
SIA Report No. 1700109.402P, Revision 3 - REDACTED Evaluation of Fatigue of ASME Section III, Class 1 Components for Turkey Point Units 3 and 4 for Subsequent License Renewal, December 2017 (44 Total Pages, including cover sheets)
 
Transient Number (3)Transient Count through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 1Station Heatup at 100&deg;F/hour20010955%15616482%
X200200200200 2Station Cooldown at 100&deg;F/hour20010955%15516482%
X200200200200 3Pressurizer Cooldown at 200&deg;F/hour (5)(17)200 9548%14214874%
X---200200200 5Station Loading at 5% power per minute145002932%27205334%
14500 (11)---145002200 (16)6Station Unloading at 5% power per minute145002422%21404403%
14500 (11)---145002200 (16)7Step Load Increase of 10% of Full Power 2000432%109794%2000---20002000 8Step Load Decrease of 10% of Full Power 2000905%2201648%
2000---20002000 9Step Load Decrease of 50% of Full Power2006834%1678241%
X 200---200200Steady State Fluctuations (12)0Exempted 0----Feedwater Cycling at Hot Standby (15)2000Exempted 2000------2000Boron Concentration Equalization (5)36000Not Counted---------36600 14Turbine Roll Test 10 1 (4)(20)10%1 (20)1 10%-------------15Hydrostatic Test at 3107 psig Pressure, 100&deg;F Temperature (6)(19)1 1 100%11 100%1---51 16Hydrostatic Test at 2485 psig Pressure and 400&deg;F Temperature (7)5 1 20%32 40%55---5 17Secondary Side Hydrostatic Test to 1356 psig---------------------------------------  Steam Generator Loop A (10)1017 / 9 (14)90%21 9 90%---1035 (23)10  Steam Generator Loop B (10)1013 / 7 (14)70%17 7 70%---1035 (23)10  Steam Generator Loop C (10)1013 / 7 (14)70%17 7 70%---1035 (23)10Primary to Secondary Side Leak Test to 2435 psig (7)150 1 1%----2 1%---150150150 18Primary to Secondary Side Leak Test to 2250 psig (6)15 1 7%02 13%---15---15 19(2)---------------------------------------  Steam Generator Loop A (13)50 9 18%42142%---505050  Steam Generator Loop B (13)50 7 14%71632%---505050  Steam Generator Loop C (13)50 7 14%41632%---505050Secondary to Primary Side Leak Test to 840 psig (6)(8)-----------------------------------  Steam Generator Loop A 15 8 53%88 53%---15---15  Steam Generator Loop B1515100%1515100%
---15---15  Steam Generator Loop C 15 9 60%99 60%---15---15TestNormalPTN-3Source of AllowablesThrough 201660 Year Projection (9)80-Year ProjectionsDesign Number Transient Number (3)Transient Count through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 21Loss of Load without Immediate Turbine Trip or Reactor Trip801519%432835%
---808080 22Loss of Off-Site AC Electrical Power 40 615%281025%
X---404040 23Loss of Flow in One Reactor Coolant Loop801418%432633%
---808080 25Reactor Trip40018346%29127268%
X400400400400 26Inadvertent Auxiliary Spray (18)(21)10 0 0%01 10%---10---10 27OBE (22)50 0 0%01020%---------20Loss of Secondary Pressure (Press Loss) (6)6 1 17%22 33%------6 6Footnotes(1)Weighted projection method used for counted normal and upset transients in which 60-year projections for either unit are over 70% of design numbers in SIR-00-089 [1].(2)Labelled as "Secondary Leak Test" in 0-ADM-553 [7]. Labelled as "Hydrostatic Pressure Test" in Table 4.1-10 [28].(3)Transient numbers from Table 3-1 of SIR-00-089 [1].(4)Not expected to have any additional cycles on RSGs.(5)Applies to Pressurizer only.(6)Applies to Steam Generator only. Labelled as "Secondary Leak Test" in 0-ADM-553 [7].(7)Limited by Reactor Coolant Pump Analysis [16, Attachment 1, pages 44 and 45].(8)Leak Test Procedure cancelled per [30].(9)60-year projections from [5, PTN-LR-00-0127 Table 10.3-1].(10)Not expected to have any additional cycles on RSGs.(11)Cycle limits for baffle-former bolts only is being lowered from 14,500 to 2,200 due EPU RCS conditions (Table 4.1-8 of UFSAR [8]).(12)Not counted, not significant contributor to fatigue usage factor.(13)80-year plant life projected cycles computed using 65 years of life for the RSGs.(14)Values are [ (pre-and post- 1987) / (post- 1987) ] cycles [5, PTN-LR-00-0127 Table 10.3-1]. (15)Not counted, intermittent slug feeding at hot standby not performed.(16)Limit of 2,200 cycles established for baffle former bolts only per UFSAR Table 4.1-8 [8].(17)(18)Spray water temperature differential to 560&deg;F.(19)Applies to Steam Generator only. Represents pre-operational test [16, Note 3 on Attachment 1, pages 44 and 45].(20)Adjustment in 60-year projection in [5, PTN-LR-00-0127 Table 10.3-1] - recorded as a value of 0 when 1 was assumed in pre-operational startup. (21)One cycle is projected for 80 years to remain within the analytical basis if that event occurs. (22)One cycle of 10 events is projected for 80 years to remain within the analytical basis if that event occurs. (23)Recommended revision 0-ADM-553 to align with UFSAR Table 4.1-10.UpsetPTN-3Source of AllowablesThrough 201660 Year Projection (9)80-Year ProjectionsDesign Number Transient Number (3)TransientCount through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 1Station Heatup at 100&deg;F/hour20012161%19118191%
X200200200200 2Station Cooldown at 100&deg;F/hour20012161%19018191%
X200200200200 3Pressurizer Cooldown at 200&deg;F/hour (5)(17)20010452%17915879%
X---200200200 5Station Loading at 5% power per minute145002602%23204843%
14500 (11)---145002200 (16)6Station Unloading at 5% power per minute145002422%21904513%
14500 (11)---145002200 (16)7Step Load Increase of 10% of Full Power 2000442%112824%
2000---20002000 8Step Load Decrease of 10% of Full Power 2000573%1231075%
2000---20002000 9Step Load Decrease of 50% of Full Power2004221%1105126%
X 200---200200Steady State Fluctuations (12)0Exempted 0----Feedwater Cycling at Hot Standby (15)2000Exempted 2000------2000Boron Concentration Equalization (5)36000Not Counted---------36600 14Turbine Roll Test 10 1 (4)(20)10%1 (20)1 10%-------------15Hydrostatic Test at 3107 psig Pressure, 100&deg;F Temperature (6)(19)11 100%11 100%1---51 16Hydrostatic Test at 2485 psig Pressure and 400&deg;F Temperature (7)51 20%32 40%55---5 17Secondary Side Hydrostatic Test to 1356 psig---------------------------------------  Steam Generator Loop A (10)1011 / 6 (14)60%15 6 90%---1035 (23)10  Steam Generator Loop B (10)1011 / 6 (14)60%15 6 70%---1035 (23)10  Steam Generator Loop C (10)109 / 5 (14)50%13 5 70%---1035 (23)10Primary to Secondary Side Leak Test to 2435 psig (7)150 1 1%----2 1%---150150150 18Primary to Secondary Side Leak Test to 2250 psig (6)15 1 7%0 2 13%---15---15 19(2)----------------------------------------  Steam Generator Loop A (13)50 6 12%4 14 28%---505050  Steam Generator Loop B (13)50 612%1114 28%---505050  Steam Generator Loop C (13)50 5 10%7 12 24%---505050Secondary to Primary Side Leak Test to 840 psig (6)(8)---------------------------  Steam Generator Loop A151493%141493%
---15---15  Steam Generator Loop B1515100%1515100%
---15---15  Steam Generator Loop C1515100%1515100%
---15---15NormalTestPTN-4Design NumberThrough 201660 Year Projection (9)80-Year ProjectionsSource of Allowables Transient Number (3)TransientCount through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 21Loss of Load without Immediate Turbine Trip or Reactor Trip801418%382734%
---808080 22Loss of Off-Site AC Electrical Power 401333%291948%
X---404040 23Loss of Flow in One Reactor Coolant Loop801114%432126%
---808080 25Reactor Trip40018747%33729273%
X400400400400 26Inadvertent Auxiliary Spray (18)(21)10 0 0%01 10%---10---10 27OBE (22)50 0 0%01020%---------20Loss of Secondary Pressure (Press Loss) (6)(21)60 0%0 1 17%------6 6Footnotes(1)Weighted projection method used for counted normal and upset transients in which 60-year projections for either unit are over 70% of design numbers in SIR-00-089 [1].(2)Labelled as "Secondary Leak Test" in 0-ADM-553 [7]. Labelled as "Hydrostatic Pressure Test" in Table 4.1-10 [28].(3)Transient numbers from Table 3-1 of SIR-00-089 [1].(4)Not expected to have any additional cycles on RSGs.(5)Applies to Pressurizer only.(6)Applies to Steam Generator only. Labelled as "Secondary Leak Test" in 0-ADM-553 [7].(7)Limited by Reactor Coolant Pump Analysis [16, Attachment 1, pages 44 and 45].(8)Leak Test Procedure cancelled per [30].(9)60-year projections from [5, PTN-LR-00-0127 Table 10.3-1].(10)Not expected to have any additional cycles on RSGs.(11)Cycle limits for baffle-former bolts only is being lowered from 14,500 to 2,200 due EPU RCS conditions (Table 4.1-8 of UFSAR [8]).(12)Not counted, not significant contributor to fatigue usage factor.(13)80-year plant life projected cycles computed using 66 years of life for the RSGs.(14)Values are [ (pre-and post- 1987) / (post- 1987) ] cycles [5, PTN-LR-00-0127 Table 10.3-1]. (15)Not counted, intermittent slug feeding at hot standby not performed.(16)Limit of 2,200 cycles established for baffle former bolts only per UFSAR Table 4.1-8 [8].(17)(18)Spray water temperature differential to 560&deg;F.(19)Applies to Steam Generator only. Represents pre-operational test [16, Note 3 on Attachment 1, pages 44 and 45].(20)Adjustment in 60-year projection in [5, PTN-LR-00-0127 Table 10.3-2] - recorded as a value of 0 when 1 was assumed in pre-operational startup. (21)One cycle is projected for 80 years to remain within the analytical basis if that event occurs. (22)One cycle of 10 events is projected for 80 years to remain within the analytical basis if that event occurs. (23)Recommended revision 0-ADM-553 to align with UFSAR Table 4.1-10.UpsetPTN-4Design NumberThrough 201660 Year Projection (9)80-Year ProjectionsSource of Allowables 0 20 40 60 80 100 120 140 160 1808/13/19728/13/19748/13/19768/13/19788/13/19808/13/19828/13/19848/13/19868/13/19888/13/19908/13/19928/13/19948/13/19968/13/19988/13/20008/13/20028/13/20048/13/20068/13/20088/13/20108/13/20128/13/20148/13/20168/13/20188/13/20208/13/20228/13/20248/13/20268/13/20288/13/20308/13/20328/13/20348/13/20368/13/20388/13/20408/13/20428/13/20448/13/20468/13/20488/13/2050Unit 3 Heatup Actual and Projected
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 5Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
SIA Report No. 1700109.401P, Revision 3 - REDACTED Evaluation of Environmentally-Assisted Fatigue for Turkey Point Units 3 and 4 for Subsequent License Renewal, January 2018 (62 Total Pages, including cover sheets)
 
Transient Number (3)Transient Count through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 21Loss of Load without Immediate Turbine Trip or Reactor Trip801519%432835%
---808080 22Loss of Off-Site AC Electrical Power 40 615%281025%
X---404040 23Loss of Flow in One Reactor Coolant Loop801418%432633%
---808080 25Reactor Trip40018346%29127268%
X400400400400 26Inadvertent Auxiliary Spray (18)(21)10 0 0%01 10%---10---10 27OBE (22)50 0 0%01020%---------20Loss of Secondary Pressure (Press Loss) (6)6 1 17%22 33%------6 6Footnotes(1)Weighted projection method used for counted normal and upset transients in which 60-year projections for either unit are over 70% of design numbers in SIR-00-089 [1].(2)Labelled as "Secondary Leak Test" in 0-ADM-553 [7]. Labelled as "Hydrostatic Pressure Test" in Table 4.1-10 [28].(3)Transient numbers from Table 3-1 of SIR-00-089 [1].(4)Not expected to have any additional cycles on RSGs.(5)Applies to Pressurizer only.(6)Applies to Steam Generator only. Labelled as "Secondary Leak Test" in 0-ADM-553 [7].(7)Limited by Reactor Coolant Pump Analysis [16, Attachment 1, pages 44 and 45].(8)Leak Test Procedure cancelled per [30].(9)60-year projections from [5, PTN-LR-00-0127 Table 10.3-1].(10)Not expected to have any additional cycles on RSGs.(11)Cycle limits for baffle-former bolts only is being lowered from 14,500 to 2,200 due EPU RCS conditions (Table 4.1-8 of UFSAR [8]).(12)Not counted, not significant contributor to fatigue usage factor.(13)80-year plant life projected cycles computed using 65 years of life for the RSGs.(14)Values are [ (pre-and post- 1987) / (post- 1987) ] cycles [5, PTN-LR-00-0127 Table 10.3-1]. (15)Not counted, intermittent slug feeding at hot standby not performed.(16)Limit of 2,200 cycles established for baffle former bolts only per UFSAR Table 4.1-8 [8].(17)(18)(19)Applies to Steam Generator only. Represents pre-operational test [16, Note 3 on Attachment 1, pages 44 and 45].(20)Adjustment in 60-year projection in [5, PTN-LR-00-0127 Table 10.3-1] - recorded as a value of 0 when 1 was assumed in pre-operational startup. (21)One cycle is projected for 80 years to remain within the analytical basis if that event occurs. (22)One cycle of 10 events is projected for 80 years to remain within the analytical basis if that event occurs. (23)Recommended revision 0-ADM-553 to align with UFSAR Table 4.1-10.UpsetPTN-3Source of AllowablesThrough 201660 Year Projection (9)80-Year ProjectionsDesign Number Transient Number (3)TransientCount through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 120012161%19118191%
X200200200200 220012161%19018191%
X200200200200 3(5)(17)20010452%17915879%
X---200200200 5Station Loading at 5% power per minute145002602%23204843%
14500 (11)---145002200 (16)6Station Unloading at 5% power per minute145002422%21904513%
14500 (11)---145002200 (16)7Step Load Increase of 10% of Full Power 2000442%112824%
2000---20002000 8Step Load Decrease of 10% of Full Power 2000573%1231075%
2000---20002000 9Step Load Decrease of 50% of Full Power2004221%1105126%
X 200---200200Steady State Fluctuations (12)0Exempted 0----Feedwater Cycling at Hot Standby (15)2000Exempted 2000------2000Boron Concentration Equalization (5)36000Not Counted---------36600 14Turbine Roll Test 10 1 (4)(20)10%1 (20)1 10%-------------15(6)(19)11 100%11 100%1---51 16(7)51 20%32 40%55---5 17Secondary Side Hydrostatic Test to 1356 psig---------------------------------------  Steam Generator Loop A (10)1011 / 6 (14)60%15 6 90%---1035 (23)10  Steam Generator Loop B (10)1011 / 6 (14)60%15 6 70%---1035 (23)10  Steam Generator Loop C (10)109 / 5 (14)50%13 5 70%---1035 (23)10Primary to Secondary Side Leak Test to 2435 psig (7)150 1 1%----2 1%---150150150 18Primary to Secondary Side Leak Test to 2250 psig (6)15 1 7%0 2 13%---15---15 19(2)----------------------------------------  Steam Generator Loop A (13)50 6 12%4 14 28%---505050  Steam Generator Loop B (13)50 612%1114 28%---505050  Steam Generator Loop C (13)50 5 10%7 12 24%---505050Secondary to Primary Side Leak Test to 840 psig (6)(8)---------------------------  Steam Generator Loop A151493%141493%
---15---15  Steam Generator Loop B1515100%1515100%
---15---15  Steam Generator Loop C1515100%1515100%
---15---15NormalTestPTN-4Design NumberThrough 201660 Year Projection (9)80-Year ProjectionsSource of Allowables Transient Number (3)TransientCount through 2016Percent of Design Number80 Year ProjectionPercent of Design NumberWeighted Projection Method (1)UFSAR Table 4.1-8UFSAR Table 4.1-100-ADM-553Minimum 21Loss of Load without Immediate Turbine Trip or Reactor Trip801418%382734%
---808080 22Loss of Off-Site AC Electrical Power 401333%291948%
X---404040 23Loss of Flow in One Reactor Coolant Loop801114%432126%
---808080 25Reactor Trip40018747%33729273%
X400400400400 26Inadvertent Auxiliary Spray (18)(21)10 0 0%01 10%---10---10 27OBE (22)50 0 0%01020%---------20Loss of Secondary Pressure (Press Loss) (6)(21)60 0%0 1 17%------6 6Footnotes(1)Weighted projection method used for counted normal and upset transients in which 60-year projections for either unit are over 70% of design numbers in SIR-00-089 [1].(2)Labelled as "Secondary Leak Test" in 0-ADM-553 [7]. Labelled as "Hydrostatic Pressure Test" in Table 4.1-10 [28].(3)Transient numbers from Table 3-1 of SIR-00-089 [1].(4)Not expected to have any additional cycles on RSGs.(5)Applies to Pressurizer only.(6)Applies to Steam Generator only. Labelled as "Secondary Leak Test" in 0-ADM-553 [7].(7)Limited by Reactor Coolant Pump Analysis [16, Attachment 1, pages 44 and 45].(8)Leak Test Procedure cancelled per [30].(9)60-year projections from [5, PTN-LR-00-0127 Table 10.3-1].(10)Not expected to have any additional cycles on RSGs.(11)Cycle limits for baffle-former bolts only is being lowered from 14,500 to 2,200 due EPU RCS conditions (Table 4.1-8 of UFSAR [8]).(12)Not counted, not significant contributor to fatigue usage factor.(13)80-year plant life projected cycles computed using 66 years of life for the RSGs.(14)Values are [ (pre-and post- 1987) / (post- 1987) ] cycles [5, PTN-LR-00-0127 Table 10.3-1]. (15)Not counted, intermittent slug feeding at hot standby not performed.(16)Limit of 2,200 cycles established for baffle former bolts only per UFSAR Table 4.1-8 [8].(17)(18)(19)Applies to Steam Generator only. Represents pre-operational test [16, Note 3 on Attachment 1, pages 44 and 45].(20)Adjustment in 60-year projection in [5, PTN-LR-00-0127 Table 10.3-2] - recorded as a value of 0 when 1 was assumed in pre-operational startup. (21)One cycle is projected for 80 years to remain within the analytical basis if that event occurs. (22)One cycle of 10 events is projected for 80 years to remain within the analytical basis if that event occurs. (23)Recommended revision 0-ADM-553 to align with UFSAR Table 4.1-10.UpsetPTN-4Design NumberThrough 201660 Year Projection (9)80-Year ProjectionsSource of Allowables
 
Component Design Number Pressurizer Spray Nozzle Pressurizer Lower HeadRSG Divider PlateRSG TubesPzr Upper Head and ShellRV Core Support BlocksCRDM Bi-Metallic (Nozzle-to-Adapter) WeldCRDM Latch HousingRVCH CRDM Nozzle and J-Groove WeldVessel Flange Minimum Cycles CRDM Lower Joint (from RVCH installation ot 80 years for bounding Unit 4)200 18118118118118118118118118192200 181181181181 181181181181Station Loading at 5% power per minute14500484533533533533533533484Station Unloading at 5% power per minute14500450533533 451451440440Step Load Increase of 10% of Full Power 2000 821641648282Step Load Decrease of 10% of Full Power 2000 106164164 106Step Load Decrease of 50% of Full Power 200 5182825182Loss of Load without Immediate Turbine Trip or 80 2728282713Loss of Flow in One Reactor Coolant Loop 80 21262621Inadve rtent Auxiliary Spray 10 2 (1)05 3 2 2Primary to Secondary Side Leak Test to 2435 psig 150 4 4Loss of AC Power 40191919 8Reactor Trip 400 311292292292 O BE 501010 10Rod Trips (2)2600 20002000144
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 6Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
SIA Environmentally Assisted Fatigue Calculations Pressurizer Lower Head 1700804.316P - REDACTED, Revision 0, 3-D Finite Element Model of Pressurizer Bottom Head, Skirt Assembly and H eater Wells, September 28, 2017 1700804.317, Revision 0, Pressurizer Low er Head Green's Functions and Unit Pressure, October 5, 2017 1700804.318, Revision 0, Pressurize r Lower Head Loads, Fatigue and EAF Analysis, November 7, 2017 Pressurizer Spray Nozzle 1700804.313P - REDACTED, Revision 1, Pressurizer Spray Nozzle Loads, December 7, 2017 1700804.314P - REDACTED, Revision 1, Pressurizer Spray Nozzle Finite Element Model and Stress Analyses, December 7, 2017 1700804.315P - REDACTED, Revision 1, Pressurizer Spray Nozzle Fatigue Analysis, December 7, 2017 (120 Total Pages, including cover sheets)
 
BottomHead.inp ANSYS geometry input file for stress analysi s
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 7Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
Westinghouse Environmentally Assisted Fatigue Calculations LTR-SDA-II-17-13-NP, Rev. 2 Environmentally Assisted Fatigue Evaluation of the Turkey Point Unit 3 and Unit 4 Pressurizer Upper Head and Shell and Reactor Vessel Core Support Blocks, November 30, 2017 LTR-CECO-17-025-NP, Rev. 1 Environmentally Assisted Fatigue Evaluation of the Turkey Point Unit 3 and Unit 4 R eplacement Steam Generators, November 30, 2017 (8 Total Pages, including cover sheets)
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 8Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
Areva Environmentally Assisted Fatigue Calculations:
Areva Letter No. AREVA-17-02742 , dated December 6, 2017 Final CUF EN Results - Turkey Point 3 & 4 - SLR EAF Analyses 1. 32-9280707, Rev. 0, Turkey Point 3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Eval uation, December 15, 2017 2. 32-9280708, Rev. 0, Turkey Po int 3 & 4 Replacement RVCH J Groove, December 12, 2017 3. 32-9280709, Rev. 0, 12/15/17, TP CRDM Latch Housing Environmentally Assisted Fatigue, December 15, 2017 4. 32-9280710, Rev. 0, TP Vent Nozzle Environmentally Assisted Fatigue, December 14, 2017 5. 32-9280711, Rev. 0, Turkey Point SLR EAF Analysis for Reactor Vessel Flange, December 14, 2017 6. 32-9280712, Rev. 0, TP CR DM Lower Joint Environmentally Assisted Fatigue, December 15, 2017 (146 Total Pages, including cover sheets)
 
LOCATION CUF EN (MAXIMUM) =
AREVA DOCUMENT SOURCE CRDM Nozzle / J
-Groove 0.2 74 32-9279212-000 CRDM Latch Housing 0.269 32-9279367-000 CRDM Nozzle to Adapter Weld 0.695 32-9279174-0 00 CRDM Lower Connection 0.420 [Unit 3]
0.749 [Unit 4]
32-9280202-000 RV Flange 0.373 32-9279161-000 RV Vent Nozzle 0.2 30 32-9279362-000 
 
Page 1 of 12 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280707 - 000 Safety Related:  Yes    No Title Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS:
Purpose:  The purpose of this document is to calculate the Cumula tive Usage Factor for the CRDM Nozzle to Adapter Weld Connection and adjust for Environmentally Assisted Fatigue per NUREG/CR-6909 Rev 1(Reference [1]).
 
The proprietary version of this document is 32-9279174-001.
Results: For 80 years plant operation, the CUF criterion is met with 0.695 < 1
[  ]       
 
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No A AREVA 0402-01-F01 (Rev. 020, 11/17/2016)
Document No. 32-9280707-000 Turkey Point -3 & 4 CROM Nozzle to Adapter Weld Connection EAF Evaluation
-Non-Proprietary Review Method: Design Review (Detailed Check) D Alternate Calcu l ation Does this document estab l ish design or technical requirements?
DYES NO Does this docwnent contain Customer Required Fomiat? DYE S NO Signature Block Name and Title (printed or typed) Caleb Tomlin Engineer II CAFM Kaihong Wang Principal Engineer CAFM David Cofflin Manager CAFM Signature P/R/A/M and LP/LR R A Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Mentor (M) Date Pages/S ect ion s Prepared/Reviewed/Approved All All In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver sha ll use 'All' or 'All except _' in the pages/sections reviewed/approved.
'All' or 'AU except_' means that the changes and the effect of the changes on the entire document have been prepared/reviewed/approved.
It does not mean that t he lead preparer/reviewer/approver has prepared/reviewed/approved a ll the pages of the document Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Date NIA NIA NIA NIA Page2 Document No. 32-9280707-0000402-01-F01 (Rev. 020, 11/17/2016)
Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 000 All Original Issue. The proprietary version of this document is 32-9279174-001.
Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 4 Table of Contents Page  SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5
 
===1.0 PURPOSE===
..................................................................................................................................... 6
 
===1.1 Background===
....................................................................................................................
................... 6
 
===2.0 METHODOLOGY===
...................................................................................................................
....... 6 2.1 General Approach to Calculating the F en and Usage Factor
............................................................. 6
 
===3.0 ASSUMPTIONS===
...................................................................................................................
......... 7
 
===3.1 Unverified===
Assumptions.........................................................................................................
............ 7
 
===3.2 Justified===
Assumptions
........................................................................................................................ 7
 
===4.0 COMPUTER===
USAGE ................................................................................................................
.... 7 5.0 INPUTS ........................................................................................................................
................. 7
 
===6.0 ANALYSIS===
......................................................................................................................
............... 7
 
===6.1 Evaluation===
using Existing CUF .................................................................................................
......... 7
 
===6.2 Refinement===
of CUF
............................................................................................................................ 8 6.2.1 F en Calculation ..................................................................................................................
.. 8 6.2.2 CUF Considering EAF ...................................................................................................... 11
 
==7.0 CONCLUSION==
....................................................................................................................
........ 12
 
==8.0 REFERENCES==
....................................................................................................................
........ 12
 
Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 5 List of Tables Page Table 6-1:  Input Data and Stress Difference Calculation ......................................................................
... 9 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen ........................ 10 Table 6-3:  [  ] .................................................................... 1 1 Table 6-4:  [  ] ................................................................ 12
 
Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 6 1.0 PURPOSE The purpose of this document is to address the concerns of Environmentally Assisted Fatigue for the CRDM Nozzle and Adapter Weld Connection utilizing NUREG/CR-6909 Rev 1 (Reference [1]). The analysis is subsequent to the analysis that was originally performed for the replacement components in Reference [2]. 1.1 Background
 
The ASME Code provides rules for the design of Class 1 components. The Code also provides fatigue design curves for applicable structural materials. However, the curves were developed using in-air testing and did not account for light water reactor coolant environments. Subsequent testing has shown that the LWR environment can substantially reduce fatigue lives for components,  [  ]  [    ]  2.0 METHODOLOGY 1.) Evaluate the existing in air cumulative usage factors (CUF) for the reported wetted surface to determine if applying the maximum F en can produce a CUF en less than 1.0.
2.) [  ] . A detailed F en will be calculated and applied to the recalculated in-air CUF. The in-air CUF will be re-calculated using the updated fatigue curves from NUREG/CR-6909 Rev 1 (Reference [1]). 2.1 General Approach to Calculating the F en and Usage Factor (1) Calculate the stress range between time point n-1 and n on the component stresses according to NB-3216.2. The stress ranges are 'ij = (n)ij - (n-1)ij. The principal stresses ( 1 ,  2 ,  3) are then calculated based on the stress ranges 'ij , and arranged such that  1 >  2 >  3. The stress intensity (SI) range is then int =  1 -  3. (2) Calculate the strain range by  n =  int  / E, where E is the Young's modulus of the material at the average metal temperature between the two time points.
(3) Calculate the strain rate by 'n =  n/t n , where t n in seconds is the time increment between point n-1 and n. (4) Determine whether or not the F en for step n shall be calculated. Since only the strain increment of increasingly tensile is considered (i.e., no negative ' is permitted), the methodology presented in Reference [3]  is adopted to determine if the strain range shall be kept:
If l 3l < l 1l, the F en for step n shall be calculated based on the strain rate (Step 3) and other applicable parameters (S*, T* and O*) according to corresponding equations listed in Sub-Section 6.1; If l 3l>l 1l, the strain range is excluded in the next step, and no F en is calculated.
(5) Calculate the F en factor between the two extreme time points of pair number k using the Multi-Linear Strain Based F en approach presented in Reference [1] (Section 6 Equation (68)):
Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 7    (6) Calculate the overall CUF with EAF for the location:  U en = F en , 1U1 + F en , 2 U 2 + F en , 3 U 3 + ***. 3.0 ASSUMPTIONS 3.1 Unverified Assumptions There are no unverified assumptions within this calculation. 3.2 Justified Assumptions 1.) [  ]  2.) Transient and External loads as analyzed in Reference [2] are unchanged.
3.) [    ]  4.0 COMPUTER USAGE No engineering software is used in this calculation.
 
===5.0 INPUTS===
The inputs for this calculation are taken from Reference [2]
[  ]  6.0 ANALYSIS 6.1 Evaluation using Existing CUF Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 8 Applying the resulting F en to the existing CUF results in a CUF en greater than of 1:  This is an unacceptable result and therefore additional analysis is required. 6.2 Refinement of CUF
[    ]  The transients and their combinations which were analyzed in Reference [2] are utilized within this analysis. 
[    ]  As stated in NB-3213.17 of the ASME Code "Fatigue strength reduction factor is a stress intensification factor which accounts for the effects of a local structural discontinuity (stress concentration) on the fatigue strength."
[  ]  The existing K e factor, from Reference [2] Section 5.4.2, of
[  ]  is then applied to the maximum Total SI Range of  [  ]    [  ]  6.2.1 F en Calculation Table 6-1 provides the output from
[  ]  in Reference [2] which is used as input for the detailed F en and Table 6-2 calculates the Detailed F en.  [  ]  [    ]
DocumentNo. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 9 Table 6-1:  Input Data and Stress Difference Calculation
 
DocumentNo. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 10 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 11  Based on the above calculations for  n and F en
* n, the usage factor can be recalculated to account for the EAF.  =  [  ]  6.2.2 CUF Considering EAF Table 6-3: 
[  ]
Document No. 32-9280707-000 Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection EAF Evaluation - Non-Proprietary Page 12 Table 6-4: 
[  ] 
 
==7.0 CONCLUSION==
For 80 years plant operation, the CUF criterion is me t with 0.695 < 1
[  ] 
 
==8.0 REFERENCES==
: 1. NUREG/CR-6909, Rev 1, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials," March 2014 (Draft Report for Comments). 2. AREVA Document, 32-5029862-009, "Turkey Point -3 & 4 CRDM Nozzle to Adapter Weld Connection." 3. Mark A. Gray, et al., "Strain Rate Calculation Approach in Environmental Fatigue Evaluations," Transaction of the ASME Journal of Pressure Vessel Technology, Vol. 136, August 2014. 4. AREVA Document 38-9279661-000, Formal Transmittal of inputs for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4.
Page 1 of 30 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280708 - 000 Safety Related:  Yes    No Title Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS:
Purpose: The purpose of this document is to evaluate the Environmentally Adjusted Cumulative Usage Factor (CUF en) at the wetted surface of CRDM nozzle and J-Groove weld for the replacement RVCH at Turkey Point Units 3 and 4.
The proprietary version of this document is 32-9279212-001.
Summary of Results: The CUF en = 0.274  [  ] , which meets the criterion of CUF en < 1.0.     
 
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No  ANSYS 15.0.7    StressRange20.exe 
 
A AREVA 0402-01-F01 (Rev. 020, 11/17/2016)
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CROM Nozzle EAF Analysis-Non-Proprietary Review Method: Design Review (Detailed Check) D Alternate Calculation Does this document establish design or technical requirements?
DYES ~NO DYES ...... ~NO Does this document contain Customer Required Format? -----------------
-------------
-----------------
-----------------
-----------------------------
-----------
-----------------------
Signature Block Name and Title (printed or typed) Signature P/R/A/M and LP/LR Date Jasmine Cao Principal Engineer 12/12/2017 Don Kim Superviso1y Engineer David Cofflin Manager Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Mentor (M) Pages/Sections Prepared/Reviewed/Approved All All All In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use 'All' or 'All except ****-*--***--
--*--='-inJ:he.pages/sections..reviewed/approved.-'.AlLor..:.AILexcept
' neans..thaUhe..changes..and.the..effecto&#xa3;the._.
__ . __ _ changes on the entire document have been prepared/reviewed/approved.
It does not mean that the lead preparer/reviewer/approver has prepared/reviewed/approved all the pages of the document.
Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Date NIA N/A NIA N/A Page2 Document No. 32-9280708-0000402-01-F01 (Rev. 020, 11/17/2016)
Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 000 All  Initial Release. The proprietary version of this document is 32-9279212-001.
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 4 Table of Contents Page SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5 LIST OF FIGURES ...............................................................................................................
.................... 6
 
==1.0 INTRODUCTION==
..................................................................................................................
......... 7
 
===2.0 METHODOLOGY===
...................................................................................................................
....... 7 3.0 ASSUMPTIONS ...................................................................................................................
......... 7
 
===3.1 Unverified===
Assumptions.........................................................................................................
............ 7
 
===3.2 Justified===
Assumptions
........................................................................................................................ 7
 
===4.0 DESIGN===
INPUT ..................................................................................................................
........... 8
 
===4.1 Geometry===
........................................................................................................................................... 8
 
===4.2 Finite===
Element Model
......................................................................................................................... 9
 
===4.3 Materials===
.....................................................................................................................
....................... 9
 
===4.4 Boundary===
Co nditions ...........................................................................................................
............ 10
 
===4.5 Loads===
.........................................................................................................................
...................... 10
 
===5.0 COMPUTER===
USAGE ................................................................................................................
.. 12 5.1 Software and Hardware..........................................................................................................
......... 12
 
===5.2 Computer===
files ................................................................................................................
................. 12
 
===6.0 RESULTS===
.......................................................................................................................
............. 14
 
===6.1 Design===
Conditi on ..............................................................................................................
............... 14
 
===6.2 Thermal===
Analysis Results ......................................................................................................
.......... 15
 
===6.3 Structural===
Analysis Results ...................................................................................................
.......... 17
 
===6.4 Stress===
from Exter nal Load due to OBE ........................................................................................... 19 6.5 In-Air Fatigue Usage Calculation ..............................................................................................
...... 19 6.6 CUF en Calculation
............................................................................................................................ 21
 
==7.0 CONCLUSION==
S ...................................................................................................................
....... 22
 
==8.0 REFERENCES==
....................................................................................................................
........ 22 APPENDIX A :
COMPARISON BETWEEN RESULTS FROM THE ORIGINAL AND NEW MODELS ..................................................................................................................... A-1 APPENDIX B :
MESH-SENSITIVITY STUDY AND SCF DETERMINATION ..................................... B-1 Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 5 List of Tables Page Table 4-1:  [
] .......................................... 11 Table 4-2:  [  ] ............... 11 Table 5-1:  Computer Files ....................................................................................................
................. 12 Table 6-1:  Nodes of interest for evaluation of temperature gradients .................................................... 16 Table 6-2:  Temperature Gradients of Interest
........................................................................................ 16 Table 6-3:  Time Points of Interest for the Transients ........................................................................
..... 16 Table 6-4:  Path Lines - Node Number and Coordinates ....................................................................... 17 Table 6-5:  [  ] ............................................................. 19 Table 6-6:  [
] ................................................... 20 Table 6-7:  [  ] ................................................. 20
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 6 List of Figures Page Figure 4-1:  Finite Element Model .............................................................................................
................ 9 Figure 6-1:  Deformation and Stress Intensity Contours - Design Condition ......................................... 15 Figure 6-2:  Path Lines for Stress Linearization ..............................................................................
........ 18 Figure A-1:  Displacement Contours of the Original (Left) and Refined (Right) FEMs ......................... A-2 Figure A-2:  Stress Intensity Contours of the Original (Left) and Refined (Right) FEMs
....................... A-2 Figure A-3:  Original (Left) vs. Refined (Right) - dT  Time History -  [  ] ......................... A-3 Figure A-4:  Original (Left) vs. Refined (Right) - Temperature Contour -  [  ] ................. A-3 Figure A-5:  Original (Left) vs. Refined (Right) - dT  Time History -  [  ] ......................... A-4 Figure A-6:  Original (Left) vs. Refined (Right) - Temperature Contour -  [  ] ................. A-4 Figure B-1:  FEM1 vs. FEM2 ....................................................................................................
............ B-1 Figure B-2:  Displacement Contours - FEM1 (Left) vs. FEM2 (Right) .................................................. B-2 Figure B-3:  Stress Intensity Contours (Overall) - FEM1 (Left) vs. FEM2 (Right) ................................ B-3 Figure B-4:  Stress Intensity Contours (Local) - FEM1 (Left) vs. FEM2 (Right) ................................... B-3 Figure B-5:  Stress Comparison of Key Locations - FEM1 (Left) vs. FEM2 (Right) ............................. B-4
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 7
 
==1.0 INTRODUCTION==
Florida Power & Light Co. (FPL) is planning to submit their License Amendment Request (LAR) for Subsequent License Renewal (SLR = life to 80 years) to the NRC, for the replacement Reactor Vessel Closure Head (RVCH) and Control Rod Drive Mechanisms (CRDM) assembly at Turkey Point 3 and 4. In order to be approved for 80 year life, the NRC requires utilities to address Environmentally-Assisted Fatigue (EAF) at susceptible locations in the RCS. Six susceptible locations are identified within the replacement RVCH and CRDM assembly as described in the scope document [4].
The purpose of this document is to evaluate the EAF within the replacement CRDM assembly. The scope of the document herein is limited to the wetted surface of the locations evaluated in Reference [3], i.e., the path lines defined on the wetted surface of the nozzle and the J-Groove weld.
 
===2.0 METHODOLOGY===
The methodology is outlined below:
(1) Resurrect the geometry from the existing ANSYS model used in the existing analyses. Redefine the node and element by re-meshing the geometry to meet discretization requirements of the analysis documented herein. Modify the ANSYS input files accordingly and create new ANSYS input files as needed.
(2) Removal of obvious conservatisms, such as in transient groupings, FSRF, etc., as needed.
(3) Execute the modified input files with the new database (model) and compare key results to those documented in the previous analyses in order to validate that the modified input file is correct.
(4) Perform the thermal and structural analyses considered in the previous analysis performed by AREVA.
(5) Define stress classification path lines and determine the maximum fatigue cumulative usage factor at the wetted surface using the in-air fatigue curve given in NUREG/CR-6909 (Reference [1]). The attempt will be made to calculate the in-air CUF according to the design number of cycles of transients. If the in-air CUF is lower enough that the expected CUF en would meet the criteria of less than 1.0, no further analysis will be performed.
[  ]  (6) Calculate F en, per RG 1.207 Rev. 1 (Reference [2]) and NUREG 6909 Rev. 1 (Reference [1]) for the wetted location having the highest in-air cumulative usage factor (CUF) to arrive at the CUF en for the locations of interest being examined.
(7) Complete documentation and perform final review. 3.0 ASSUMPTIONS 3.1 Unverified Assumptions There are no unverified assumptions used for the calculation herein. 3.2 Justified Assumptions Justified assumptions are listed in the body of the calculation where applicable.
(1) The CRDM nozzle joint consists of a piece of RV head, a CRDM nozzle, J-Groove weld and a piece of cladding. The cladding is exposed to the fluid. However, since cladding is not a structural component and is assumed to be failed in the analysis herein, it is not evaluated for the EAF requirement.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 8 (2) The RV head is not exposed to fluid since it is covered by the cladding on its inside surface. However, since the cladding is assumed to be failed, the head is evaluated against the EAF requirement. This is conservative and acceptable.
(3) [      ]  (4) [    ]    4.0 DESIGN INPUT
 
===4.1 Geometry===
As indicated in Step (1) of Section 2.0, the geometry is from the existing analysis as documented in Reference [3]. 
[  ]  The key dimensions are taken from Reference [3] and listed below for convenience: RV Head inside radius to base metal =
[  ]  RV Head thickness =  [  ]  RV Head cladding thickness =  [  ]  Penetration diameter =  [  ]  CRDM nozzle OD =  [  ]  CRDM nozzle ID =
[  ]  J-groove outer radius =  [  ]  Buttering Weld layer thickness =  [  ]  Angle between weld and nozzle =  [  ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 9 4.2 Finite Element Model The finite element model is shown in Figure 4-1. 
[  ]  The finite element model is documented in database file "TP_CRDM_Model_new.cdb", in the form of ANSYS commands. 
[    ]    A mesh sensitivity study is performed and documented in Appendix B. The results show that the discretization of the model, as shown in Figure 4-1, for the regions of interest is able to properly capture the stresses needed for the purpose of the analysis.
Figure 4-1:  Finite Element Model
 
===4.3 Materials===
Material properties of analyzed components are taken from Reference [3]:
RV Head =
[  ]  CRDM nozzle =
[  ]  Cladding =
[  ]  J-Groove buttering
=  [  ]  J-Groove filler
=  [  ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 10 4.4 Boundary Conditions The same boundary conditions, as described in Reference [3], are applied to the new finite element model, with the exception of the simulation of the interface between the outside diameter of the nozzle and inside diameter of the head penetration in the thermal analysis:
[    ]  This is confirmed from the comparison documented in Appendix A;
[  ]  [  ]  This is conservative and acceptable;  [    ]  4.5 Loads The same loads, as documented in Section 3.5 of Reference [3], are used in the analyses performed herein.
External load
: The external loads due to OBE will be used in the fatigue analysis. The OBE loads are taken from Reference [3] and listed below for convenience:
Operating Transient Load
: The heat transfer coefficients used in the thermal analysis are taken from subsection 3.5.3 of Reference [3] and listed below for convenience:  [ ]  [    ]  [  ]  The same practice is adopted in the analysis documented herein. The transient loads are taken from Tables 10 through 15 of Reference [3].
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 11 [    ]    Table 4-1: 
[  ]  Table 4-2: 
[  ]  Note:
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 12 5.0 COMPUTER USAGE 5.1 Software and Hardware The finite element analysis documented in this report is performed using ANSYS 15.0.7, Reference [6]. The error notices of this version of ANSYS have been reviewed and concluded that ANSYS 15.0.7 is acceptable to use. All computer runs are performed on the following computer: 
(1) AUSLYNCHPCI06; Intel Xeon CPU E5-2640 v3 @ 2.60GHz; 396 GB RAM; Operating System: Red Hat Enterprise Server v6.4 (Linux); Kernel: 2.6.32-358.el6.x86_64 (2) Person running the tests: Jasmine Cao (3) ANSYS Verification Files Test Date: 11/09/2017 and 11/17/2017. Table 5-1 lists the ANSYS verification run results. Evaluation of the verification results shows all tests were successful. 5.2 Computer files The computer files have been uploaded to ColdStor and listed in Table 5-1 below.
Table 5-1:  Computer Files
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 13 Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 14    6.0 RESULTS 6.1 Design Condition Stress analysis under the design pressure is performed to verify the behavior of the new model is as expected. The design condition stress analysis is documented in ANSYS output file "TP_CRDM_DesignCond.out".
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 15 Figure 6-1 shows the model deformation and stress intensity contours of the model under the design pressure. It is seen that the deformation and the stress intensity distribution of the model are as expected.
Figure 6-1:  Deformation and Stress Intensity Contours - Design Condition 6.2 Thermal Analysis Results Thermal analysis is performed using the enveloped or grouped transients as documented in Reference [3]. The transient loads are documented in the ANSYS input files "TP_*_tr.mac", where
* represents the name of the transients, i.e.,  [  ]    The thermal analysis is documented in ANSYS output files "TP_CRDM_*_th.out", where
* represents the name of the transients, i.e.
[  ]  The results of the thermal analysis are evaluated to identify the peaks and valleys of the temperature gradients time history between key locations of interest. The temperature distributions at the time points of these peaks and valleys are used as temperature loads in the structural analysis to calculate thermal stresses. The same key locations as those documented in Reference [3] are selected for temperature gradient evaluation. Due to re-discretization of the finite element model, node numbers are different from the original document. Table 6-1 summarizes the nodes of interest for evaluation of temperature gradients. The temperature gradients of the same pairs of the locations, as documented in Table 17 of Reference [3], are used in the evaluation. The temperature gradient of interest is su mmarized in Table 6-2 for convenience. The temperature gradient evaluation is documented in the ANSYS output files "TP_CRDM_*_th_post.out", where
* represents the name of the transients, i.e.
[  ]  Besides the time points when the extrema occur in the temperature gradient time history, the time points when the pressure time history start, change, or end are also selected in the structural analysis. The time points to be used for structural analysis are summarized in Table 6-3. 
[    ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 16 Table 6-1:  Nodes of interest fo r evaluation of temperature gradients Table 6-2:  Temperature Gradients of Interest Table 6-3:  Time Points of Interest for the Transients In addition, the results from the thermal analysis are compared to the results from the original analysis. The comparison shows that the results from the new model are in agreement with the original results. The comparison is documented in Appendix A.
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 17 6.3 Structural Analysis Results  Structural analysis is performed to calculate the stresses due to thermal and pressure transient loads. The transient loads and the time points are documented in ANSYS files "TP_*_Pr_tr.mac", where
* represents the name of the transients, i.e.
[  ]  Note that the transient data stored in these files are exactly the same as those for thermal transient analysis, except the time points at which analysis is to be performed are different. The analysis is documented in ANSYS output files "tp_st_*.out The path lines for stress linearization are shown in Figure 6-2 and listed in Table 6-4.
Table 6-4:  Path Lines - N ode Number and Coordinates Note:
* These nodes are not located on the wetted surface (refer to Figure 6-2) and the cumulative usage factors of these nodes are not calculated and reported in the document herein.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 18 Figure 6-2:  Path Lines for Stress Linearization
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 19 6.4 Stress from External Load due to OBE Stresses due to OBE external loads are calculated and documented in ANSYS output file "TP_CRDM_OBE_afterEUP_rough.out". The stresses for each path line is extracted and documented in "TP_CRDM_OBE_afterEUP_rough_post.out". The stress intensity range is calculated using EASI list engineering tool StressRange20.exe (Reference [7]) and documented in following files:  "TP_CRDM_OBE_afterEUP_rough_post.Class_Line_Summary(M+B).dat"  "TP_CRDM_OBE_afterEUP_rough_post.Class_Line_Summary(Total).dat" The stress range calculated based on total stress is used in the fatigue calculation. The stress due to OBE load is combined with the stress due to transient loads in the fatigue calculation. 
[  ]  This is conservative and acceptable. 6.5 In-Air Fatigue Usage Calculation ANSYS Fatigue module is utilized to calculate the in-air cumulative usage factor. Total stresses are used in the fatigue calculation for all locations investigated. Per Appendix B, the discretization of the model is refined enough to properly capture the peak stress of the locations of interest. 
[      ]  The calculation of the maximum in-air CUF is detailed in Table 6-5.
Table 6-5: 
[  ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 20  CUF based on
[  ] : [    ]  The maximum in-air CUF occurs at the uphill side of the weld fillet. The calculation is detailed in Table 6-7.
Table 6-6: 
[  ]  Note: Stress due to OBE load is included in the CUF calculation.
Table 6-7: 
[  ]  Note:  (1) This value reflects the number of used cycles for event pair
[  ]  (2) The value is based on the alternating stress including OBE loads;
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 21 6.6 CUF en Calculation As mentioned in Assumption (1) in Subsection 3.2, cladding (stainless steel) is not evaluated for the EAF requirement. The EAF evaluations for J-groove weld and CRDM
[  ]  and RV head
[  ]  are as follows: Locations in J-Groove Weld and CRDM Nozzle  [  ] : The wetted location with the highest in-air CUF is WPATH2,  [  ]  The CUF reported in Section 6.5 for this location is  [  ]  and the maximum stress intensity range is  [  ]    This is an acceptable result and therefore no additional analysis is required.
Locations in Head  [  ] : The base metal of head does not expose to the fluid. However, it is conservatively assumed that the cladding has crack therefore the inside surface of the head need to be qualified to the EAF criterion.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page 22 This is an acc eptable result and theref ore no addit ional analy s is is required.
 
==7.0 CONCLUSI==
ONS The environmental aided fatigue cumulative usage factor
[  ]  is 0.274, which is well below the requirement of 1.0, therefore meet the fatigue criterion.
 
==8.0 REFERENCES==
[1] NUREG/CR-6909, Rev. 1,  Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Draft Report for Comment, U.S. NRC [2] REGULATORY GUIDE 1.207, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components due to the Effects of the Light-Water Reactor Environment for New Reactors [3] AREVA Document, 32-5037379-008, Turkey Point -3 & 4 Replacement RVCH CRDM Nozzle Connection [4] AREVA Document 51-9277194-000, Turkey Point Units 3 and 4 Environmentally Assisted Fatigue Evaluation [5] AREVA Document 38-9279661-000, Formal Transmittal of inputs for AREVA Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4  [6] ANSYS Release 15.0.7, UP20140420, LINUX x64, ANSYS Inc., Canonsburg, P.A. [7] AREVA EASI Software, StressRange2.0.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page A-1  APPENDIX A: COMPARISON BETWEEN RESULTS FROM THE ORIGINAL AND NEW MODELS A.1 Background The original FEM documented Reference [3] is created in ANSYS 7.0. The FEM is able to be loaded in ANSYS 15.0.7. The model has been converted into a database file in the form of ANSYS commands file and uploaded into ColdStor. The original FEM was meshed in
[  ]  were conservatively used in the fatigue calculation to cover any uncaptured peak stresses due to discretization of the original mesh. In order to remove this conservatism, a refined model is built based on the same geometry as in the original document. A refined model is created and benchmarked to the original results. The following comparisons are made between the results from the refined and original FEMs. For Design Condition:
o Total displacement contour; o Stress intensity contour;  For thermal analysis:
o Differential temperature (dT) time history; o Temperature contour comparison for a randomly selected time point. A.2 Design Condition Comparison The total displacement and stress intensity contours from the original and the refined FEMs are shown Figure A-1 and Figure A-2. It is seen from Figure A-1 that the patterns of the displacement contours of the two FEMs are very similar. 
[  ]  It is seen from Figure A-2 that the patterns of the stress intensity contours of the two FEMs are similar. The maximum stress intensities from both models occur at the weld fillet regions. The magnitude of the maximum stress intensity from the refined model is higher than that of the original model, as expected.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page A-2 Figure A-1:  Displacement Contours of the Original (Left) and Refined (Right) FEMs Figure A-2:  Stress Intensity Contours of the Original (Left) and Refined (Right) FEMs A.3 Thermal Analysis Comparison The dT time histories and temperature distribution contours at a randomly selected time point for  [  ]  transients are shown in Figure A-3, Figure A-4, Figure A-5 and Figure A-6. It is seen the dT time histories and the temperature distribution from both models are very close, though the transition of temperature contour is smoother for the refined model than that of the original coarse model.
It is concluded that the refined FEM produces comparable (very close) results as the original FEM for the thermal analysis, as expected.
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page A-3  Figure A-3:  Original (L eft) vs. Refined (Right) - dT  Time History -
[  ]  Figure A-4:  Original (Left) vs. Refined (Right) - Temperature Contour -
[  ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page A-4 Figure A-5:  Original (L eft) vs. Refined (Right) - dT  Time History -
[  ]  Figure A-6:  Original (Left) vs. Refined (Right) - Temperature Contour -
[  ]
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page B-1 APPENDIX B: MESH-SENSITIVITY STUDY AND SCF DETERMINATION B.1 Background  The FEM from the original document has been refined to form the FEM used in the analysis documented herein, in order to properly capture peak stress. This FEM is labeled as FEM1. Mesh sensitivity study is performed for the regions of interest, i.e., the weld fillet and the CRDM nozzle wall. Two cases are compared:  Case 1:  [  ]  This model is used in the analysis;  Case 2:  [  ]  This model is considered to be able to produce more accurate stresses in the region of interest. The comparison of the results from the two cases is taken as the base for FSRF determination.
Design pressure produces tensile and bending stresses in the region of interest, therefore is good for use of mesh sensitivity study and FSRF determination. B.2 The Two Finite Element Models A comparison of the two FEMs is shown in Figure B-1.
Figure B-1:  FEM1 vs. FEM2
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page B-2 B.3 Results and Conclusion The overall deformations of the two models are shown in Figure B-2. It is seen that the overall deformation of the two models are comparable, as expected. The over stress intensity of the two models are shown in Figure B-3. The overall maximum SINT from both FEMs occur at the crevice between the CRDM nozzle and the bore of the head. 
[  ]  Figure B-4 shows the detailed stress intensity distribution at the weld fillet region from the two models. 
[    ]  Figure B-5 shows a comparison of the stress intensities at the key locations of the two models. It is seen that the differences of the results from the two models are negligible indicating that model FEM1 is good enough for the purposed analysis. 
[  ]  Figure B-2:  Displacement Contours - FEM1 (Left) vs. FEM2 (Right)
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page B-3 Figure B-3:  Stress Intensity Contours (Overall) - FEM1 (Left) vs. FEM2 (Right)
Figure B-4:  Stress Intensity Contours (Local) - FEM1 (Left) vs. FEM2 (Right)
 
Document No. 32-9280708-000 Turkey Point 3 & 4 Replacement RVCH CRDM Nozzle EAF Analysis- Non-Proprietary Page B-4 Figure B-5:  Stress Comparison of Key Locat ions - FEM1 (Left) vs. FEM2 (Right)
 
Page 1 of 24 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280709 - 000 Safety Related:  Yes    No Title TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS:
Purpose The purpose of this calculation is to determine the effects of environmental assisted fatigue on the Latch Housing. The basis used for the determination is DRAFT Regulatory Guide 1.207 and DRAFT NUREG/CR-6909  Revision 1.
 
The proprietary version of this document is 32-9279367-001. Results The results indicate that the Latch Housing fatigue remains acceptable when the Environmental Assisted Fatigue (EAF) is considered. The EAF usage factor is calculated
[  ]  and is acceptable. The maximum calculated EAF usage factor including the environmental assisted fatigue is, U en = 0.269  [  ]
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No A AREVA 0402-01-F01 (Rev. 020, 11/17/2016)
Document No. 32-9280709-000 TP CROM Latch Hou sing Environmentally Assisted Fatigue -Non-Prop rietary Review Method: Design Review (Deta il ed Check) D Alte rn ate Ca l culat i on D oes this docume n t estab li s h de sig n or tec hn ical r e quir e m en t s? D YES N O Does this document contain Customer Required Format? D YES NO Signature Block P/R/A/M Name and Title and Pages/Sections (printed or typed) Signature LP/LR Date Prepared/Reviewed/Approved Th omas M. WasWco ~VW\Wv--p I?..~ I s--wq HT Harrison :17~* R IZ.!;5 -~ /17 David Coffl i n \D.. {Miee_ .X A j,% E n gineering
%r Manager No te s: P/R/A desi g nate s Preparer (P), Reviewer (R), Approver (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Mentor (M) A ll Al l A ll In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use 'All' or 'A ll except _' in the pages/sections reviewed/approved.
'All' or 'All except_' means that th e changes and the effect of the changes on the entire document have been prepared/reviewed/approved. It doe s not mean that the lead preparer/reviewer/approver has prepared/reviewed/approved all t h e pages of th e document.
Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Date NIA Page 2 Document No. 32-9280709-0000402-01-F01 (Rev. 020, 11/17/2016)
TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 0 All Original Issue. The proprietary version of this document is 32-9279367-001.
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 4 Table of Contents Page  SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5 LIST OF FIGURES ...............................................................................................................
.................... 6
 
===1.0 OBJECTIVE===
.....................................................................................................................
............. 7
 
===2.0 METHODOLOGY===
...................................................................................................................
....... 7 2.1 General Approach to Calculating the Fen and EAF Usage Factor ................................................... 7
 
===2.2 Detailed===
Method of Calculation of the Fen for Latch Housing........................................................... 7
 
===3.0 ASSUMPTIONS===
...................................................................................................................
......... 8
 
===4.0 DESIGN===
INPUTS .................................................................................................................
......... 8
 
===5.0 COMPUTER===
USAGE ................................................................................................................
.... 8 6.0 CALCULATIONS ..................................................................................................................
......... 8
 
===6.1 Transients===
To Be Evaluated ....................................................................................................
.......... 8
 
===6.2 EXCEL===
Spreadsheet Calculations ................................................................................................
.. 12 7.0 RESULTS .......................................................................................................................
............. 20
 
==8.0 REFERENCES==
............................................................................................................................ 21 APPENDIX A :
BASIS FOR PRINCIPAL STRESS CALCUATIONS
................................................... A-1
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 5 List of Tables Page Table 6-1:  Input Data and Stress Difference Calculation for Transients 3 and 16 ................................. 14 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen ........................ 15 Table 6-3  Input Data and Stress Difference Calculation for Transient 1
............................................... 16 Table 6-4  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transient 1 ...................................................................................................................
......... 17 Table 6-5:  [  ]
................................................................. 18 Table 6-6:  [  ] ......................................... 19
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 6 List of Figures Page Figure 6-1:  Time History Stresses for Transients 16 and 3 ..................................................................... 9 Figure 6-2:  Transients 16 and 3 Stress Versus Time ............................................................................ 10 Figure 6-3:  Strain Versus Time for Two Transients ............................................................................
... 10 Figure 6-4:  Fen Versus Strain for Two Transients .............................................................................
.... 11 Figure 6-5  Time History Stresses for Transients 16 and 1 .................................................................... 11 Figure 6-6:  S-N Curve Showing Significance of NUREG/CR-6909 to ASME III .................................... 20
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 7 1.0 OBJECTIVE The component to be qualified for EAF in this calculation is the CRDM latch housing. Only the inside wetted location is to be considered. A fatigue usage factor is to be calculated considering the EAF and the objective is to obtain a value that is less than 1.0.
 
===2.0 METHODOLOGY===
Calculations are performed based on the criteria below, which is based on DRAFT NUREG/CR-6909, Reference [1] . The applicable code is ASME III, 1989 edition with no Addenda Reference [2]. 2.1 General Approach to Calculating the Fen and EAF Usage Factor Time-history stresses are provided for various transients in Reference [3]. From this information, the following procedure is used to calculate the Fen and resulting EAF usage factor.
(1)  Calculate the stress range between time point n-1 and n on the component stresses according to NB-3216.2. The stress ranges are 'ij = (n)ij - (n-1)ij. The principal stresses ( 1 ,  2 ,  3) are then calculated based on the stress ranges 'ij , and arranged such that  1 >  2 >  3. The stress intensity (SI) range is then int =  1 -  3. (2) Calculate the strain range by  n =  int  / E, where E is the Young's modulus of the material at the average metal temperature between the two time points.
(3) Calculate the strain rate by 'n =  n/t n , where t n in second is the time increment between point n-1 and n. (4) Determine whether or not an F en for step n shall be calculated. Since only the strain increment of increasingly tensile strain is considered (i.e., no negative ' is permitted), the following methodology is adopted to determine if the strain range shall be kept:
If l  3l < l  1l, the F en for step n shall be calculated based on the strain rate (Step 3) and other applicable parameters (S*, T* and O* ) according to Eqs. (2), (3) or (4);
If l_3 l>l_1 l, the strain range is excluded in the next step, and no Fen is calculated.
(5) Calculate the F en factor between the two extreme time points of pair number k using the "Multi-Linear Strain Based Method" F en approach presented in Reference [1]. Equation (68) of  Reference [1].
(6) Calculate the overall EAF en usage factor for the location:  U en = F en , 1 U1 + F en , 2 U 2 + F en , 3 U 3 + ***. 2.2 Detailed Method of Calculation of the Fen for Latch Housing Detailed description of the F en calculation applicable to the latch housing:
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 8  3.0 ASSUMPTIONS No assumptions or Modeling Simplifications were used. 4.0 DESIGN INPUTS The plant design, past and projected operating information was provided in Reference [3], Reference [5], and Reference [6]. 5.0 COMPUTER USAGE EXCEL spreadsheet only. 6.0 CALCULATIONS Calculations are performed based on the methodology of Reference [1]. Based on the information provided in Reference [3], the total "in-air" usage factor was calculated to be [  ]      This total usage is mostly from a range between transients 16 and 1, 16 and 3. The stress time-histories are also provided in a table in this same calculation and an example is displayed in the following plot, Figure 6-1. 
[    ]  6.1 Transients To Be Evaluated The following transients will be shown to be the contributors to the EAF usage factor:
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 9  Figure 6-1:  Time History Stresses for Transients 16 and 3 The plots in Figure 6-2 are shown to establish the basis for the strain rate calculations. 
[  ]
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 10 Figure 6-2:  Transients 16 and 3 Stress Versus Time An example of two transients to be integrated are shown below in Figure 6-3. The hand-drawn figures below are for demonstration of methodology and do not represent the exact data.
 
Figure 6-3:  Strain Versus Time for Two Transients
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 11 Figure 6-4:  Fen Versus Strain for Two Transients
 
Figure 6-5  Time History Stresses for Transients 16 and 1 Transients 1 and 16 are plotted below. Due to the differences in the time durations of transients 1 and 16, refer to the preceding figures for a better view of the time history of transient 16. 
[  ]
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 12 6.2 EXCEL Spreadsheet Calculations The calculations necessary to perform this evaluation were done with EXCEL. The following is an explanation of the equations used in the spreadsheet.
For Sheet: "Latch Housing"
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 13 Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 14 [  ]  Table 6-1:  Input Data and Stress Difference Calculation for Transients 3 and 16
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 15 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 16 Table 6-3  Input Data and Stress Difference Calculation for Transient 1
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 17 Table 6-4  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transient 1
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 18 For Transients 3 and 16:
Based on the above calculations for n and Fen
* n, the usage factor can be recalculated to account for the EAF.  =  [  ]  For Transients 1 and 16:
Based on the above calculations for n and Fen
* n, the usage factor can be recalculated to account for the EAF.  =  [  ]  [    ]  Table 6-5: 
[  ]
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 19 [  ]  Table 6-6: 
[  ]  Total EAF Usage =0.269 Reported previous "in-air" ASME usage factor without EAF was
[  ]  per Reference [3], page 13. Additional notes:
(4) The EAF Usage Factors above were calculated with the Fatigue Design Curve from Reference [1], Table A.2.   
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 20 The following S-N curve is based on NUREG/CR-6909, Reference [1] and ASME III 1989, Reference [2] and shows the significance of including the NUREG/CR-6909 curve which reduces the number of allowable cycles significantly for an alternating stress of 50.246 ksi.
Figure 6-6:  S-N Curve Showing Significance of NUREG/CR-6909 to ASME III
 
===7.0 RESULTS===
Per Table 6-4, the EAF usage factor has been recalculated and shown to be less than 1.0
[  ]
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page 21
 
==8.0 REFERENCES==
References identified with an (*) are maintained within Turkey Point Nuclear Power Plant Records System and are not retrievable from AREVA Records Management. These are acceptable references per AREVA Administrative Procedure 0402-01, Attachment 8. See page 2 for Project Manager Approval of customer references.
: 1. NUREG/CR-6909, Revision 1, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials", DRAFT, March 2014. 2. ASME B&PV Code Sec. III, SubSection 1989 edition with no Addenda. 3. AREVA Document 38-9279681-000, JSPM Document 17NI0548, Rev. B, "Turkey Point Power Plant - Data for the fatigue analysis of the critical points of latch housing and canopy joint". 4. Textbook, "Advanced Strength and Applied Elasticity The SI Version", A.C. Ugural, S.K. Fenster, 1982, Elsevier Science Publishing Co. 5. AREVA Document 33-9127371-001, "Turkey Point Plant Extended Power Uprate Control Rod Drive Mechanism Pressure Housing Assembly Appurtenances ASME Class 1". 6. AREVA Document 08-5036747-003, JSPM Document 6GA4571 Rev. H, "Turkey Point Nuclear Power Plant Units 3 & 4 Control Rod Drive Mechanism Pressure Housing Assembly Appurtenances ASME III
 
Class 1 Design Specification". 7. Areva Document 38-9279661-000, [NNPARV-17-0241, NNPARV-17-0243], "Formal Transmittal of inputs for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4".
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page A-1  APPENDIX A: BASIS FOR PRINCIPAL STRESS CALCUATIONS A.1 Equations for Principal Stress Calculations The source equations from the text, "Advanced Strength and Applied Elasticity The SI Version",  [4] are shown here for documentation purposes.
 
Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page A-2 Document No. 32-9280709-000 TP CRDM Latch Housing Environmentally Assisted Fatigue - Non-Proprietary Page A-3 Page 1 of 29 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280710 - 000 Safety Related:  Yes    No Title TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS: PURPOSE: The purpose of the analysis is to evaluate the Environmentally Assisted Fatigue (EAF) of the Reactor Vessel Closure Head Vent Nozzle for Turkey Point Units 3 and 4, according to the F en methodology provided in NUREG/CR-6909 Rev. 1 (Reference [1]).
 
The proprietary version of this document is 32-9279362-001.
 
RESULTS: The maximum cumulative fatigue usage factor considering EAF is 0.23
[  ]         
 
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No  ANSYS 16.0 A .AR EVA 0402-01-F01 (Rev. 020, 11/17/2016)
Document No. 32-9280710-000 TP Vent Nozzle Envi r onme n tally Assis t ed Fatigue -Non-Proprietary Review Method: IZJ Design Review (D etai l ed Check) D Alternate Calculation Does this document establish design or technical requirements?
DYES IZJ NO Does this document contain Customer Required Format? DYES IZJ NO Signature Block P/R/A/M Name and Titl e and Pag es/Sections (printed or typed) Signature LP/LR Dat e Prepared/R ev iewed/Appro v ed Kaihong Wang, Principal Eng i neer p 12113117 T h omas Washko, *~~vJ,~ Engineer R l'Z.-\,\-~q David Cofffo1, ~J.J:._t~/}
-A~ Manager A !z~ 14/1,-:r N o tes: P/R/A designates Preparer (P), Reviewer (R), Approver (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Mentor (M) All. All, detailed review. All. In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use 'All' or 'AU except _' in the page s/sections reviewed/approved.
'All' or 'All except_' means that the changes and the effect of the changes on the entire document have been prepared/reviewed/approved.
It does not mean that the lead preparer/rev i ewer/approver has prepared/reviewed
/approved all the pages of the document.
Project Manager Approval of Custom e r References and/or Customer Formatting (N/A if not a p plicable)
Name Title (printed or typed) (p r inted or typed) Signatur e Date NIA NIA NIA NIA Page2 Document No. 32-9280710-0000402-01-F01 (Rev. 020, 11/17/2016)
TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 000 Initial release The proprietary version of this document is 32-9279362-001.
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 4 Table of Contents Page SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5 LIST OF FIGURES ...............................................................................................................
.................... 6
 
==1.0 INTRODUCTION==
..................................................................................................................
......... 7
 
===2.0 PURPOSE===
AND SCOPE .............................................................................................................
.. 7 3.0 ANALYTICAL METHODOLOGY ................................................................................................... 7 3.1 ASME Section III In-Air Fatigue Usage .........................................................................................
.... 7 3.2 F en Methodol ogy ..................................................................................................................
.............. 8
 
===4.0 ASSUMPTIONS===
...................................................................................................................
......... 9
 
===4.1 Unverified===
Assumptions.........................................................................................................
............ 9
 
===4.2 Justified===
Assumptions
........................................................................................................................ 9
 
===4.3 Modeling===
Simplifications ......................................................................................................
............ 10
 
===5.0 DESIGN===
INPUTS .................................................................................................................
....... 10
 
===5.1 Geomet===
ry ......................................................................................................................
................... 10
 
===5.2 Material===
......................................................................................................................
...................... 11
 
===5.3 Finite===
Element Model
....................................................................................................................... 11
 
===5.4 Loads===
.........................................................................................................................
...................... 14
 
===6.0 COMPUTER===
USAGE ................................................................................................................
.. 16 6.1 Software ......................................................................................................................
.................... 16
 
===6.2 Computer===
Files ................................................................................................................
................ 16
 
===7.0 CALCULATIONS===
..................................................................................................................
....... 18
 
===7.1 Design===
Condi tions .............................................................................................................
.............. 18
 
===7.2 Thermal===
Analysis ..............................................................................................................
............... 19
 
===7.3 Stress===
Analysis ...............................................................................................................
................. 24
 
===7.4 Cumulative===
Fatigue Usage Factor
................................................................................................... 25
 
===8.0 RESULTS===
 
==SUMMARY==
AND CONCLUSION ............................................................................... 28
 
==9.0 REFERENCES==
....................................................................................................................
........ 29
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 5 List of Tables Page Table 5-1: Enveloped Affecting Transients for Fatigue Evaluation ......................................................... 15 Table 5-2: Enveloped Transients ...............................................................................................
............. 15 Table 7-1: Locations for Temperature Gradients ................................................................................
.... 20 Table 7-2: Time Points in Structural Analysis .................................................................................
........ 24 Table 7-3: Nozzle EAF Usage Factor ............................................................................................
......... 26 Table 7-4: J-Groove Weld EAF Usage Factor .....................................................................................
... 27 Table 7-5: RV Head EAF Usage Factor
.................................................................................................. 27
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 6 List of Figures Page Figure 5-1: 2-D Solid Model ...................................................................................................
................. 12 Figure 5-2: Meshed Finite Element Model .......................................................................................
....... 13 Figure 5-3: Thermal Boundary Conditions .......................................................................................
....... 13 Figure 5-4: Structural Boundary Conditions ....................................................................................
........ 14 Figure 7-1: Displacement in Design Conditions (inch) ..........................................................................
.. 18 Figure 7-2: Stress Intensity in Design Conditions (psi) .......................................................................
.... 19 Figure 7-3: Locations for Evaluation of Temperature Gradients ............................................................. 20 Figure 7-4:  [  ] ........................................................................
.. 21 Figure 7-5:  [  ] ........................................................................
.. 21 Figure 7-6:  [  ] ........................................................................
... 22 Figure 7-7:  [  ] ........................................................................
.. 22 Figure 7-8:  [
].........................................................................
... 23 Figure 7-9:  [  ] ........................................................................
.. 23 Figure 7-10: Locations for EAF Evaluation .....................................................................................
........ 25
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 7
 
==1.0 INTRODUCTION==
Florida Power & Light Co. (FPL) is planning to submit a License Amendment Request (LAR) for Subsequent License Renewal (SLR = life to 80 years) to the NRC for the replacement Reactor Vessel Closure Head (RVCH) and CRDM assembly at Turkey Point Units 3 and 4. As required by the NRC, the Environmentally-Assisted Fatigue (EAF) at susceptible locations in the RCS needs to be evaluated. 
 
===2.0 PURPOSE===
AND SCOPE The purpose of this document is to calculate the EAF usage factors for the RVCH Vent Nozzle. The EAF evaluation is based on the F en methodology provided in NUREG/CR-6909 Rev. 1 (Reference [1]). The original ASME Section III analysis of the Vent Nozzle is documented in Reference [2], in whic h the plant EPU transients are considered. Since the original analysis does not include the fatigue usage calculation for the insi de surfaces of the J-Groove weld and the nozzle that are exposed to the RCS coolant, a finite element analys is is re-performed to calculate the stresses and fatigue usage factors at the inside surfaces, followed by the evaluation of EAF usage based on the F en methodology.
The scope is to calculate the cumulative usage factor (CUF) considering EAF on the inside surfaces of:  Vent Nozzle near the J-Groove weld
[  ]  J-Groove weld including the buttering
[  ]  RV closure head near the J-Groove weld
[  ]  The list above covers all material types in the component.
ASME Section III stress and fatigue analyses are contained in the original analysis (Reference [2]), which remains valid. Note that the original stress and fatigue analysis qualifies the Vent Nozzle to the ASME Code 1989 Edition with no addenda. The current analysis calculate the in-air fatigue usage factor based on the fatigue design curves developed in NUREG/CR-6909 Rev. 1 (Reference [1]), as part of the LAR commitment.
 
===3.0 ANALYTICAL===
METHODOLOGY 3.1 ASME Section III In-Air Fatigue Usage A finite element analysis is performed, with the same steps as described in Reference [2] to calculate stresses due to design transients and nozzle external loads:
: 1. [  ]  There are two finite element models consisting of thermal and structural elements respectively so as to enable the thermal and structural analysis.
: 2. Applying the design conditions of pressure and temperature to the structural finite element model and obtaining the deformation and stresses in the model. The deformation field is used to verify the correct behavior of the model and correct modeling of boundary and load conditions.
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 8 3. Applying the thermal loads pertaining to the power transients (in the form of transient temperatures and corresponding heat transfer coefficients versus time). Each of the major power transients requires a separate run on the thermal finite element model.
: 4. Evaluating the results of the thermal analysis by examination of the magnitude of temperature differences between key locations of the model at corresponding times (for example between nozzle and wall). The time points of maximum temperature gradient are the time points at which the maximum thermal stresses develop or the biggest pressure exists.
: 5. Applying the corresponding mechanical (pressure) and thermal loads (temperature gradients) at each time point identified in step 4 on the structural finite element model.
: 6. Applying the stress intensity due to the OBE loads calculated in Reference [2] to applicable locations on
 
the nozzle.
: 7. Calculating the in-air CUF on the inside surfaces of the nozzle, J-Groove weld and RV head based on the fatigue design curves as summarized in Appendix A of Reference [1]. This analysis does not use the simplified methodology described in NRC RIS 2008-30 where the influence function and only one value of stress is used in the fatigue evaluation. This is a detailed finite element analysis, meeting the requirements of ASME III, NB-3216.2, where all six stress components are used in calculating fatigue usage factors.
3.2 F en Methodology The F en method is developed in NUREG/CR-6909 Rev. 1 (Reference [1]). This sub-section summarizes the formulas provided in Appendix A of Reference [1].
The F en method is an acceptable approach in the EAF evaluation, in which the F en factor is a nominal correction value defined as the ratio of fatigue life in air at room temperature (Nair,RT) to that in LWR coolant environments at service temperature (Nwater): 
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 9 The final fatigue usage considering EAF is then: . where U i is the in-air fatigue usage factor.
 
===4.0 ASSUMPTIONS===
4.1 Unverified Assumptions There are no unverified assumptions within this calculation. 4.2 Justified Assumptions
: 1. The loading conditions (design transients, nozzle external loads) are the same as those used in the original analysis.
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 10 2. [  ]  3. The RV head inside surfaces are covered with cladding. For EAF evaluation, the head inside surface near the J-Groove weld is considered being exposed to the RCS coolant although it is covered with cladding.
: 4. [  ]  5. [ ]  6. In NUREG/CR-6909 Rev. 1 (Reference [1]), when calculating T* or  in Appendix A, the upper bound limit of equation for T is 325&deg;C (617&deg;F).
[  ]  4.3 Modeling Simplifications Simplifications in modeling and simulation in the analysis are the same as in the original analysis.
 
===5.0 DESIGN===
INPUTS
 
===5.1 Geometry===
The geometry of the model is based on the drawings provided in References [3], [4]
and [5], same as in the original analysis. RV Head inside radius to base metal  =  [  ]  RV Head thickness    =
[  ]  RV Head cladding thickness, min.  =
[  ]
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 11 Vent nozzle OD, min.    =  [  ]  Vent nozzle ID, max.    =  [  ]  J-Groove weld height  =
[  ]  Buttering weld layer height  =
[  ]  Angle between weld and nozzle  =  [  ]  [  ]  5.2 Material The materials designation is provided in Reference [3] and [7]. Same as in the original analysis, the material designations of various sub-components as listed as follows:
RV Head =
[  ]  Reference [3]
Vent nozzle =
[  ]  Reference [7]
Cladding(1) =  [  ]  Reference [7] J-Groove buttering (2) =  [  ]  Reference [4]
J-Groove filler (2) =  [  ]  Reference [4] Notes: (1) Thermal properties in ASME code are grouped by their chemical composition.
[  ]  (2) Section 4.4 of Reference [7] identifies the weld material in accordance with ASME Code Cases 2142-1 and 2143-1. These code cases provided specification of weld filler metal and welding electrode, respectively.
[  ]  The analysis herein uses the thermal properties - mean coefficient of thermal expansion (), specific heat (C), thermal conductivity (k) and the mechanical properties - modulus of elasticity (E), Poisson's ratio (), density (). The detailed values (thermal & structural) for these materials are listed in Tables 3-1 to 3-3 of Reference [2].
 
===5.3 Finite===
Element Model The finite element model is developed with ANSYS (Reference [8]). The solid model is shown in Figure 5-1. The model is meshed with
[  ]  The meshed model is shown in Figure 5-2. The ANSYS finite element model is documented in the file "TPVent_Geo.dat" and "TPVent_Geo.out."
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 12 The same boundary conditions as used in the original analysis are applied in both thermal and structural analyses.
[  ]  In structural analysis, symmetric boundary conditions are applied to the head cut-off cross-section. End-cap pressure load is applied to the nozzle cut-off cross-section.
[  ]    Figure 5-1: 2-D Solid Model
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 13 Figure 5-2: Meshed Finite Element Model Figure 5-3: Thermal Boundary Conditions 
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 14 Figure 5-4: Structural Boundary Conditions 
 
===5.4 Loads===
The external nozzle loads are provided in Reference [9]. The forces and moments along with the corresponding locations are summarized in the tables in Sub-Section 3.5.1 of the original analysis (Reference [2]). As with the fatigue usage, a conservative value of  [  ]  due to nozzle external loads is determined from the original analysis (Sub-Section 5.4.2.2 in Reference [9]), which will be added to the stress intensity ranges for nodes on the nozzle.  [  ]  The reactor vessel head is designed to satisfy the ASME code criteria when operating at a pressure of
[  ]  and a temperature of  [  ]  (Reference [7]). The design conditions are simulated on the model by applying a uniform and reference temperature of  [  ]  throughout the model and a uniform pressure of
[  ]  on all those surfaces in contact with the primary reactor coolant. These surfaces include the inside surfaces of the reactor vessel head, inside surfaces of the vent nozzle, and the surface of the weld joining the nozzle to the reactor head. The Normal and Upset transients are listed in Table 3-8 of Reference [2]. The enveloped affecting transients for fatigue evaluation are listed in Table 3-9 of Reference [2], as repeated in Table 5-1. Per Reference [6], the number of cycles for the analyzed transients is applicable for 80 years plant operation.
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 15 Table 5-1: Enveloped Affecting Transients for Fatigue Evaluation In Table 5-1the enveloped transients as well as the total number of cycles are the same as liste d in Table 3-9 of Reference [2]. The corresponding transient data as listed in Tables 3-10 to 3-15 of Reference [2] are listed as follows.
[  ]  Table 5-2: Enveloped Transients
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 16 6.0 COMPUTER USAGE 6.1 Software ANSYS Release 16.0 is used in this calculation. Verification tests are listed as follows: Computer programs tested: ANSYS Release 16.0.
Verification Tests:
[  ]  Verification output files are identified in the computer listing, using the naming convention of pre/post_
* where "*" represents the Test Name. Computer hardware used: 
: 1) Computer Name: SC-MJEHGTA. Hardware: Intel (R) Xeon&#x17d; CPU E5645@2.40GHz, 24GB RAM. Operating System: Windows 7 Enterprise, Service Pack 1, 64-Bit.
: 2) Name of person running the tests: Kaihong Wang.
: 3) Date of tests: 11/7/2017 and 11/12/2017. Acceptability: Output files of all verification tests have been reviewed and found to be successful. Verification output files are also uploaded to AREVA ColdStor. 6.2 Computer Files Computer files generated for this analysis and the verification tests have been uploaded to AREVA ColdStor in the directory"\cold\General-Access\32\
32-9000000\32-9279362-000\official\." A complete directory listing of all filenames, sizes, dates and sub-directories has been provided in the following under the abovementioned directory. All times in the listing file and below are the U.S. Central Standard Time (CST).  [last modified date/time]--------------------------[file size]--[filename] -------------------------------------------------------------------------------------------------------------------------------------- 11/12/2017  06:36 PM        1,274,273 InAirCUF_Head.out 11/12/2017  06:36 PM            4,325 InAirCUF_HEAD.txt
 
11/12/2017  06:38 PM        3,024,099 InAirCUF_JWeld.out
 
11/12/2017  06:38 PM            10,523 InAirCUF_JWeld.txt
 
11/12/2017  06:39 PM        2,447,746 InAirCUF_Noz.out 11/12/2017  06:39 PM            8,457 InAirCUF_NOZ.txt 11/07/2017  09:49 AM            1,821 TH_dt.mac
 
11/12/2017  06:27 PM            31,373 TPVent_Des.out
 
11/04/2017  12:08 PM        1,621,522 TPVent_Geo.dat
 
11/07/2017  09:50 AM            85,943 TPVent_Geo.out 11/05/2017  12:16 PM              946 TR01.mac 11/07/2017  09:58 AM            93,854 TR01_st.out
 
11/07/2017  09:56 AM          673,036 TR01_th.out
 
11/05/2017  12:16 PM              978 TR02.mac
 
11/07/2017  11:43 AM            93,101 TR02_st.out 11/07/2017  11:41 AM          673,036 TR02_th.out 11/05/2017  11:25 PM            1,512 TR03.mac
 
11/07/2017  10:17 AM            85,817 TR03_st.out
 
11/07/2017  10:15 AM          593,619 TR03_th.out
 
11/05/2017  11:26 PM            1,682 TR04.mac 11/07/2017  10:24 AM            85,825 TR04_st.out 11/07/2017  10:22 AM          723,844 TR04_th.out
 
11/12/2017  06:23 PM            1,420 TR05.mac Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 17 11/12/2017  06:35 PM            81,253 TR05_st.out 11/12/2017  06:33 PM          871,187 TR05_th.out 11/05/2017  11:30 PM            1,848 TR06.mac 11/07/2017  10:41 AM            87,506 TR06_st.out
 
11/07/2017  10:39 AM          902,396 TR06_th.out
 
11/12/2017  06:40 PM            31,422 VM112_post.out
 
11/07/2017  09:39 AM            31,608 VM112_pre.out 11/12/2017  06:45 PM            79,248 VM211_post.out 11/07/2017  09:46 AM            79,434 VM211_pre.out
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 18 7.0 CALCULATIONS 7.1 Design Conditions Design conditions are simulated in the model by applying a uniform and reference temperature of
[  ]  throughout the model and uniform pressure of
[  ]  (Reference [7]). The purpose is to provide a basis for verification of the correct behavior of the model, the structural boundary conditions, and to verify stress attenuation at regions away from the nozzle. The ASME Section III primary stress analysis of Design Conditions is included in Reference [2]. The general deformation and the stress intensity (SI) plots are shown in Figure 7-1and Figure 7-2, respectively. Design Conditions analysis is documented in the ANSYS file "TPVent_Des.out."
Figure 7-1: Displacement in Design Conditions (inch)
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 19  Figure 7-2: Stress Intensity in Design Conditions (psi)
 
===7.2 Thermal===
Analysis The results of the thermal analyses are evaluated to identify the maximum and minimum temperature gradients between critical locations in the model and the corresponding time points. These temperature gradients generate maximum and minimum thermal stresses, which in turn contribute to the maximum range of stress intensities in the model. Similar locations as in the original stress analysis (Reference [2]) are selected. The node numbers corresponding to the two locations for evaluation of temperature gradient are listed in Table 7-1 for the model, as shown in Figure 7-3. The corresponding ANSYS output files are listed in Section 6.2 with the file name convention as "TR*_th.out" for thermal output file. Here the asterisk "*" is to be substituted by the transient number as 01, 02, -, 06 as listed in Table 5-2. Transient input files are named as "TR*.mac" that include the temperature and internal pressure all as a function of time and are called by each of the "TR*_th.inp" file. The input file for location definition is "TH_dt.mac."
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 20 Table 7-1:
Locations for Temperature Gradients The temperature gradients between the two nodes as a function of time are shown in Fig ure 7-4 to Figure 7-9. These figures are provided to show the trend and for visual aid only. Specific values are taken from the ANSYS output files
 
Figure 7-3: Locations for Evaluation of Temperature Gradients
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 21 Figure 7-4: 
[  ]  Figure 7-5:
[  ]
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 22 Figure 7-6:
[ ]  Figure 7-7:
[ ]
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 23 Figure 7-8:
[  ]  Figure 7-9:
[  ]
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 24 7.3 Stress Analysis Nodal solution of the thermal analysis is loaded into the structural analysis with ANSYS. Time points selected from the thermal analysis include those with max/min temperature gradients as well as those where the internal pressure changes in linear interpolations. These time points are tabulated in tables below. Internal pressure at each time point is added as the mechanical load. The ANSYS output files are listed in Section 6.2 with the file name convention as "*_st.out" where "*" is to be substituted by the transient index as listed in Table 5-2. Pressure input tables are defined in "TR*.mac."
Table 7-2:
Time Points in Structural Analysis Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 25 7.4 Cumulative Fatigue Usage Factor For consideration of fatigue usage, the total stress intensity ranges are calculated.
[  ]  Based on the stress contours, fatigue locations are selected as shown in Figure 7-10 where the FEM node numbers are indicated. The bounding transients (and the corresponding cycles) that have a potential impact on fatigue usage factor are listed in Section 5.4.
[  ]  The in-air cumulative fatigue usage factor is calculated using the ranges obtained from the stress results and the design cycles listed in Table 5-1. The in-air CUF is calculated based on the design fatigue curves in NUREG/CR-6909 Rev. 1 (Reference [1]). The corresponding maximum F en value for the subject material (see Sub-Section 3.2) is then multiplied with the in-air CUF to obtain the final EAF CUF en. Figure 7-10: Locations for EAF Evaluation 
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 26 In the following tables (Table 7-10 to Table 7-12), only the locations having the highest cumulative usage factors and only SI ranges having nonzero partial usage factors for each component are documented. The detailed calculation includes: (1) Req'd Cycles - the number of cycles.
(2) Peak SI range - the extreme Total stress intensity range between two load steps, including SI range due to external loads if applicable.
 
(3) S a = (Peak SI range*E_ratio)/2, where E_ratio = E_curve/ E_material Table 7-3: Nozzle EAF Usage Factor Note:  (1) E_ratio=E_curve/E_material, where E_material is the nozzle material Young's module at  [  ]  and E_curve is the Young's module of the testing material in the design fatigue curve.
(2) [ ]
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 27 Table 7-4: J-Groove Weld EAF Usage Factor Note:  (1) E_ratio= E_curve/E_material, where E_material is the nozzle material Young's m odule at  [  ]  and E_curve is the Young's module of the testing material in the design fatigue curve.
Table 7-5: RV Head EAF Usage Factor Note:  (1) E_ratio= E_curve/E_material, where E_material is the nozzle material Young's m odule at  [  ]  and E_curve is the Young's module of the testing material in the design fatigue curve.
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 28 8.0 RESULTS
 
==SUMMARY==
AND CONCLUSION Finite element analysis of the Turkey Point Units 3 and 4 RVCH Vent Nozzle is performed, similar to the original stress analysis (Reference [2]), in order to obtain the in-air CUF at the wetted surfaces near the J-Groove weld. An EAF analysis using the F en method documented in Reference [1] is then performed.
[  ]  the highest CUF en considering EAF is found to be 0.23
[  ]   
 
Document No. 32-9280710-000 TP Vent Nozzle Environmentally Assisted Fatigue - Non-Proprietary Page 29
 
==9.0 REFERENCES==
: 1. NUREG/CR-6909, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials," Rev. 1, March 2014 (draft for comments). 2. AREVA Document 32-5037640-006, "Turkey Point Un its 3 and 4 Reactor Vessel Head Vent Nozzle Qualification." 3. AREVA Drawing 02-5023321E-10, "Specification Drawing For Replacement Reactor Vessel Closure Head Turkey Point, Units 3 and 4." 4. AREVA Drawing 02-5027458E-04, "Turkey Point Units 3 and 4, J-Groove Weld Details." 5. AREVA Drawing 02-5028736D-01, "Vent Pipe Assembly for FPL, Turkey Point, Units 3 and 4." 6. AREVA Document 38-9279661-000, "Formal Transmittal of inputs for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4." 7. AREVA Document 08-5023846-04, "Certified Design Specification - Reactor Vessel Closure Head Replacement, Florida Power and Light Company, Turkey Point, Units 3 and 4," as amended by SA-PTN 3/4-02. 8. "ANSYS" Finite Element Computer Code, Version 16.0, ANSYS, Inc., Canonsburg, Pa. 9. AREVA Document 18-5027466-05, "Loading Specification and Design Transients for Reactor Vessel Closure Head, Control Rod Drive and Integrated Head Assembly Replacements Turkey Point - Units 3 and 4."
Page 1 of 22 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280711 - 000 Safety Related:  Yes    No Title Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS: PURPOSE: The purpose of the analysis is to evaluate the Environmentally Assisted Fatigue (EAF) of the Reactor Vessel Flange for Turkey Point Units 3 and 4, according to the F en methodology provided in NUREG/CR-6909 Rev 1 [1].
The proprietary version of this document is 32-9279161-001.
 
RESULTS: [  ]  the cumulative fatigue usage factor considering EAF is 0.373.
[  ]
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No  None A .AREVA 0402-01-F0 1 (Rev. 020, 11/17/2016)
Document No. 32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange -Non-Proprietary Review Method: !XI Design Review (Detailed Check) D Alternate Calculation Does this document estab lish design or technical requirements?
DYES !XI NO Does this document contain Customer Required Format? DYES NO Signature Block P/R/A/M and Pages/Sections Name and Title (printed or typed) Signature LP/LR Date Prepared/Reviewed/
Approved Milan KOPINEC Engineer Kaihong Wang Principal Engineer David Cofflin Manager \p R A iy$4v4~ 12./tll /17 l--2(~~ 1,1-Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Mentor (M) All. All, detailed review. All. In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use 'All' or 'All except _' in the pages/sections reviewed/approved.
'AIJ' or 'All except_' means that the changes and the effect of the changes on the entire document have been prepared/reviewed/approved. It does not mean that the lead preparer/reviewer/approver has prepared/reviewed/approved al l the pages of the document.
Project Manage r Approval of Customer References and/or Customer Formatt i ng (N/A if not applicable)
Name Title (printed or typed) (printed or typed) S ign ature Date NIA NIA NIA NIA Page2 Document No. 32-9280711-0000402-01-F01 (Rev. 020, 11/17/2016)
Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 000 All Initial Revision. The proprietary version of this document is 32-9279161-001.
Document No. 32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 4 Table of Contents Page  SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5 LIST OF FIGURES ...............................................................................................................
.................... 6
 
==1.0 INTRODUCTION==
..................................................................................................................
......... 7
 
===2.0 PURPOSE===
AND SCOPE .............................................................................................................
.. 7 3.0 ANALYTICAL METHODOLOGY ................................................................................................... 7
 
===4.0 ASSUMPTIONS===
...................................................................................................................
......... 8
 
===4.1 Unverified===
Assumptions.........................................................................................................
............ 8
 
===4.2 Justified===
Assumptions
........................................................................................................................ 8
 
===5.0 DESIGN===
INPUTS .................................................................................................................
......... 8
 
===6.0 COMPUTER===
USAGE ................................................................................................................
.... 9 7.0 CALCULATIONS ..................................................................................................................
....... 10 7.1 In-Air Fatigue Usage Factor ...................................................................................................
......... 10
 
===7.2 Stresses===
(prior to EPU) .......................................................................................................
............ 15 7.3 F en Calculation (prior to EPU) ...................................................................................................
....... 18
 
===7.4 Environmentally===
Assi sted Fatigue Usage........................................................................................
20
 
==8.0 CONCLUSION==
....................................................................................................................
........ 21
 
==9.0 REFERENCES==
....................................................................................................................
........ 22
 
Document No. 32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 5 List of Tables Page Table 7-1:  Fatigue Usage Factor from [2] prior to EPU ......................................................................... 11 Table 7-2:  Fatigue Usage Factor from [2] with EPU ............................................................................
.. 11 Table 7-3:  In-Air Fatigue Usage Factor using ANL Fatigue Design Curve, prior to EPU
....................... 12 Table 7-4:  In-Air Fatigue Usage Factor using ANL Fatigue Design Curve, with EPU ........................... 13 Table 7-5:  Total Stresses prior to EPU .......................................................................................
........... 15 Table 7-6:  [  ] ............................. 17 Table 7-7:  Calculation Summary of Strain Increments ..........................................................................
19 Table 7-8:  Calculation Summary of F en .................................................................................................. 19 Table 7-9:  [  ] .............................................................................
.... 21 Document No. 32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 6 List of Figures Page Figure 7-1:  Critical Fatigue Location ........................................................................................
.............. 10 Figure 7-2:  Temperature Gradients in Thermal Analysis .......................................................................
10 Figure 7-3:  Transients Contributing to the Fatigue Usage .....................................................................
14 Figure 7-4:  Temperature Difference Plots and Time Points Chosen for Structural Analysis ................. 16
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 7
 
==1.0 INTRODUCTION==
Florida Power & Light Co. is planning to submit a License Amendment Request for Subsequent License Renewal (SLR = life to 80 years) to the NRC for the replacement Reactor Vessel Closure Head (RVCH) and CRDM assembly at Turkey Point Units 3 and 4. In order to be approved for 80 year life, the NRC requires utilities to address Environmentally-Assisted Fatigue (EAF) at susceptible locations in the RCS. 
 
===2.0 PURPOSE===
AND SCOPE Six susceptible locations are identified within the replacement RVCH and CRDM assembly as described in the scope document [1]. This report is focused on the EAF analysis of the Reactor Vessel Flange which is one of the identified locations. The analysis takes advantage of the calculation results documented in [2] , which is a detailed stress and fatigue calculation of Turkey Point 3 and 4 replacement RVCH following rules specified in ASME Code Section III 1998 ed ition no addenda. The EAF analysis herein is based on the guidelines given in NUREG/CR-6909 Rev 1 [3]  so that the environmental effects are incorporated into ASME Code Section III fatigue evaluation performed in [2] with environmental correction factor 'F en'. 3.0 ANALYTICAL METHODOLOGY The EAF analysis of the Reactor Vessel Flange (RV Flange) is performed with the stress and fatigue results documented in [2] along with environmental correction factor 'F en' determined in accordance with [3].
The following steps outline the analytical methodology used:
: 1) The critical location for the fatigue damage determined in [2] is identified.
: 2) Recalculation of in-air fatigue usage factor with the ANL fatigue design curve from [3] and number of cycles specified in [4]. As reported in the original stress analysis [2] for the usage factor with transients prior to EPU and with EPU, the in-air usage factor is re-calculated also with both sets of transients. The results are reviewed to determine which set of transients is bounding in the subsequent EAF usage calculation.
: 3) The component stresses from [2] are retrieved at the critical location at each time point of the transients contributing to the total fatigue usage factor. 
: 4) Review of the calculated stress states in the course of the transients and their applicability for EAF analysis.
: 5) Calculation of F en in accordance with NUREG/CR-6909 Rev 1 [3].
: 6) Multiplication by the F en factor for each contributing transient pair identified in step no. 2) to obtain final EAF usage factor.
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 8 4.0 ASSUMPTIONS 4.1 Unverified Assumptions There are no unverified assumptions within this calculation. 4.2 Justified Assumptions
: 1. The loading conditions (design transients, nozzle external loads) are the same as those used in the original analysis.
: 2. The RV head and vessel inside surfaces are covered with cladding.
[    ]  3. [    ] . 
 
===5.0 DESIGN===
INPUTS The analysis of EAF of RV Flange is based on:
- calculation results from [2] (see also Section 6.0)
- calculation guidelines [3]
- [  ]  - [  ]  - [  ]
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 9 6.0 COMPUTER USAGE All calculations are made within MS Excel 2010. No further software was used. Computer files retrieved from AREVA ColdStor which were generated in [2] are listed as follows:  Total Stress Intensity Ranges:
Ranges(total).out 11/25/2008 Ranges(Total)_EPU.out 11/16/2009 Temperature differences through the vessel wall:  HUCD_DeltaTs.out 11/25/2008  PLUL_DeltaTs.out 11/25/2008 RIRD_DeltaTs.out 11/25/2008 Stress components:
TP_ST_HUCD_Path.out 11/25/2008 TP_ST_PLUL_Path.out 11/25/2008  TP_ST_PLPU_EPU_Path.out 11/15/2009  TP_ST_RIRD_Path.out 11/25/2008 TP_ST_RIRD_EPU_Path.out 11/12/2009
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 10 7.0 CALCULATIONS 7.1 In-Air Fatigue Usage Factor The in-air ASME Code Section III maximum usage factor of the RV Flange calculated in [2] is
[  ]  The location is shown in Figure 7-1 (inside node of Path 6, al so the location labeled with G in thermal analysis shown in Figure 7-2). The usage is based on the transients and design number of cycles defined in [5]. The fatigue calculation summary from [2] at the inside node of Path 6 is provided in Table 7-1.
Figure 7-1:  Critical Fatigue Location Figure 7-2:  Temperature Gradients in Thermal Analysis
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 11 In the following fatigue usage calculation, [  ]  that can be combined with any other transient stress conditions to calculate the stress intensity range.
Table 7-1:  Fatigue Usage Factor from [2] prior to EPU The Revision 004 of [2] has also analyzed t he impact of the Extended Power Uprate (EPU).
[  ]  However, this assumption is not necessarily conservative in terms of EAF evaluation, since 'F en' factor might be actually larger for the contributing transient pair having smaller stress amplitude with a smaller strain rate in comparison to the one with larger stress amplitude and larger strain rate. The fatigue calculation summary from [2] at the inside node of Path 6 with EPU transients is provided in Table 7-2.
Table 7-2:  Fatigue Usage Factor from [2] with EPU In order to calculate the EAF usage factor using the F en method,  [  ]  needs to be applied to calculate the in-air fatigue usage factor. Since the ANL fatigue design curve is located above the ASME fatigue curve used in the original analysis, the usage factor decreases along with Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 12 reduction of conservatisms described in [3].
[    ] This simplification increases the E ratio factor and so the adjusted alternating stress.
With the ANL fatigue design curve
[ ]  the in-air fatigue usage calculation is re-performed with aforementioned changes. The stresses and metal temperatures at the node no. 202260 (inside node of Path 6) for the load combinations contributing to usage factor are retrieved from these computer outputs: TP_ST_HUCD_Path.out, TP_ST_PLUL_Path.out, TP_ST_PLPU_EPU_Path.out TP_ST_RIRD_Path.out,  TP_ST_RIRD_EPU_Path.out Ranges(total).out,  Ranges(Total)_EPU.out The fatigue usage recalculation is documented in Table 7-3 for the inside node of Path 6 prior to EPU, and in Table 7-4 for the same node with EPU. In Table 7-3 and Table 7-4, [  ]  In the table:  LC 1 and 2 - ANSYS Load Cases comprising the load combination  Time 1 and 2 - time points associated with the load cases and transient pair under consideration  Range - total stress intensity range Emat,min - Young's modulus at metal temperature.
[  ]    Ecurve - Young's modulus
[  ]  S alt - calculated alternating stress intensity range from ANSYS Table 7-3:  In-Air Fatigue Usage Factor using ANL Fatigue Design Curve, prior to EPU
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 13 Table 7-4:  In-Air Fatigue Usage Factor using ANL Fatigue Design Curve, with EPU The fatigue recalculation shows a significant decrease of in-air usage fact or from [  ]  for transients prior to EPU and from
[  ]  for transients with EPU. 
[  ]  For the total usage factor with EPU transients, the second combination is between  [  ]  The  [  ]  transients are plotted in Figure 7-3.
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 14  Figure 7-3:  Transients Cont ributing to the Fatigue Usage
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 15 7.2 Stresses (prior to EPU) The component stresses for
[  ]  transients for all calculated time points prior to EPU are retrieved from the "Ranges(total).out" computer file and are summarized in Table 7-5.
Table 7-5:  Total Stresses prior to EPU
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 16 Temperature difference plots between inside and outside vessel walls at the location of Path 6 for both transients are shown in Figure 7-4. The plots also depict the time points at which the stresses were calculated (black dots).
[ ]  Figure 7-4:  Temperature Difference Plots and Time Points Chosen for Structural Analysis Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 17 Table 7-6: 
[  ]  [  ]  Transient The logic used for
[  ]  transient is also used for the
[  ]  transient.
[  ]
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 18 7.3 F en Calculation (prior to EPU)
[    ]  This approach is described in detail in Section 6 of [3]. The calculation steps are repeated herein:
: 1) Calculate the stress range between time points  and  on the component stresses according to NB-3216.2. The stress ranges are . The principal stresses () are then calculated based on the stress ranges , and arranged such that . The stress intensity (SI) range is then . 2) Calculate the strain range by , where  is the Young's module of the material at the average temperature of the metal between the two time points.
: 3) Calculate the stain rate by  , where  in second is the time increment between time points  and  4) Determine whether or not a  for step  shall be calculated. Since only the strain increment of increasingly tensile is considered (i.e., no negative  is permitted), the following criterion from [6] is adopted to determine if the strain range shall be kept:
If , the  for step n shall be calculated based on the strain rate (Step 3) and other applicable parameters (S*, T* and O* ) according to equations shown below in text; If , the strain range is excluded in the next step, and no  is calculated.
: 5) Calculate the  factor between the two extreme time points of pair number  using the Multi-Linear Strain Based approach provided in [3] (see Section 6 Equation (68) in [3]): . 6) Calculate the overall EAF usage factor for the location: . The value for dissolved oxygen (DO) is obtained from [4]: 
[  ]  Finally, the F en can be calculated using
[  ]  Table 7-7 and Table 7-8 document the calculation of strain increments and calculation of 'F en', respectively.
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 19 Table 7-7:  Calculation Summary of Strain Increments Note 1): Negative sign signifies a compressive strain increment.
Table 7-8:  Calculation Summary of F en Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 20 7.4 Environmentally Assisted Fatigue Usage The combination of in-air cumulative usage factors () from Table 7-3 with corresponding  factors from Table 7-8 gives the final environmentally assisted fatigue usage:
  [    ]  [    ]  [    ]  The environmentally assisted fatigue usage for
[  ]  is:  In Section 7.1 it was stated that in-air fatigue usage assuming EPU could potentially result in a larger fatigue usage provided that F en factors determined for the combinations including EPU transient are sufficiently larger than the ones calculated in Table 7-8 because of their smaller strain rates.
[fulfilled
]    Moreover, the values O* and S* are unchanged and so these two variables also cannot increase the F en factors.
[    ]  For these reasons, the calculated CUF en which is based on the transients prior to EPU is bounding. 
  [ ] [ ]  [  ]
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 21 Table 7-9: 
[  ] 
 
==8.0 CONCLUSION==
 
The former in-air ASME Code Section III fatigue usage analysis made in [2] calculates the usage of
[  ]  [    ]  The above calculated usage factors are based on
[  ]  the usage is significantly reduced to 0.373. 
 
Document No.32-9280711-000 Turkey Point SLR EAF Analysis for Reactor Vessel Flange - Non-Proprietary Page 22 
 
==9.0 REFERENCES==
: 1. AREVA Document 51-9277194-000, "Turkey Point Units 3 and 4 Environmentally Assisted Fatigue Evaluation." 2. AREVA Document 32-5040304-006, "Turkey Point Units - 3 & 4 Closure Analysis w/ Replacement Head." 3. NUREG/CR-6909, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials," Rev. 1, March 2014 (Draft Report for comment). 4. AREVA Document 38-9279661-000, "Formal Transmittal of inputs for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4." 5. AREVA Document 18-5027466-05, "Loading Specification and Design Transients for Reactor Vessel Closure Head, Control Rod Drive and Integrated Head Assembly Replacements Turkey Point - Units 3 and 4." 6. Mark A. Gray, et al., "Strain Rate Calculation Approach in Environmental Fatigue Evaluations," Transaction of the ASME Journal of Pressure Vessel Technology, Vol. 136, August 2014.
 
Page 1 of 26 0402-01-F01 (Rev. 020, 11/17/2016)
CALCULATION
 
==SUMMARY==
SHEET (CSS) Document No. 32 - 9280712 - 000 Safety Related:  Yes    No Title TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary PURPOSE AND
 
==SUMMARY==
OF RESULTS:
Purpose The purpose of this calculation is to determine the effects of EAF (environmental assisted fatigue) on the Lower Joint. The basis used for the determination is DRAFT NUREG/CR-6909 Revision 1.
The proprietary version of this document is 32-9280202-001. Results The results indicate that the Lower Joint fatigue remains acceptable when the Environmental Assisted Fatigue is considered. The calculated EAF usage factor including the environmental assisted fatigue is 0.420 for Unit 3 and 0.749 for Unit 4,  [  ]
If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODE/VERSION/REV CODE/VERSION/REV Yes  No A AREVA 0402-01-F01 (Rev. 020 , 11/17/2016)
Document No. 32-9280712-000 TP CROM Lowe r Joint Environmentally Assisted Fatigue -Non-P r opr ietary Review M ethod: D esig n Review (Detailed Check) D Alternate Calcu l ation Do es th i s document establish design or technical requirements?
DYES NO Does this document contain Customer Required Format? D YES NO S i gnat u re B l ock P/R/A/M Na m e and T itl e a n d P ages/Sect i ons (pr i nted or typed) S i gna t u r e L P/L R D a t e P repared/Rev i ewe d/App r oved Thomas M. Washko p ~Jni~ ('L-( ) -U, tr HT Harrison /7/k. R IZ/ /5/17 D avid Coffiin ~P(~ ()---)4 *L,,~ Engineering ~1 Manager rv71' ' No te s: PIRIA designates Preparer (P), Reviewer (R), App r over (A); LP/LR designates Lead Preparer (LP), Lead Reviewer (LR); M designates Men tor (M) All All A ll In prepari n g, r ev i ewing and approv in g r ev i sions, the lead preparer/reviewer/approver shall use 'All' or 'All except _' i n th e pages/sections reviewed/approved. 'A ll' or 'A ll except_' means that the changes and the effect of the changes on the entire document have b ee n prepared/reviewed/approved.
It does n ot mean that the lead preparer/reviewer/approver has prepared/reviewed/approved all t h e pages of the document.
Pr o ject M anage r App r o v a l of Cus t omer Re f e r e n ces a n d/or C u s to me r Formatti ng (N/A i f n ot a p p li c a b l e) N ame Ti t l e (p rin ted o r typed) (pri n ted o r typed) Signature D ate NIA NIA NIA NIA Page2 Document No. 32-9280712-0000402-01-F01 (Rev. 020, 11/17/2016)
TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 3 Record of Revision Revision No. Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 0 All Original Issue. The proprietary version of this document is 32-9280202-001.
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 4 Table of Contents Page  SIGNATURE BLOCK ...............................................................................................................
................. 2 RECORD OF REVISION ............................................................................................................
.............. 3 LIST OF TABLES ................................................................................................................
..................... 5 LIST OF FIGURES ...............................................................................................................
.................... 6
 
===1.0 OBJECTIVE===
.....................................................................................................................
............. 7
 
===2.0 METHODOLOGY===
...................................................................................................................
....... 7 2.1 General Approach to Calculating the Fen and Usage Factor ........................................................... 7
 
===2.2 Detailed===
Method of Calculation of the Fen for Lower Joint ............................................................... 7
 
===3.0 ASSUMPTIONS===
...................................................................................................................
......... 8
 
===4.0 DESIGN===
INPUTS .................................................................................................................
......... 8
 
===5.0 COMPUTER===
USAGE ................................................................................................................
.... 8 6.0 CALCULATIONS ..................................................................................................................
......... 8
 
===6.1 Transients===
To Be Evaluated ....................................................................................................
.......... 8
 
===6.2 EXCEL===
Spreadsheet Calculations ................................................................................................
.. 11 7.0 RESULTS .......................................................................................................................
............. 22
 
==8.0 REFERENCES==
............................................................................................................................ 23 APPENDIX A :
BASIS FOR PRINCIPAL STRESS CALCULATIONS
................................................. A-1
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 5 List of Tables Page Table 6-1:  Input Data and Stress Difference Calculation for Transients 1 and 7 ................................... 14 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transients 1 and 7 ............................................................................................................
.... 15 Table 6-3:  Input Data and Stress Difference Calculation for Transients 10 and 11 ............................... 16 Table 6-4:  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transients 10 and 11 ..........................................................................................................
.. 17 Table 6-5  [  ]
................................................................................... 19 Table 6-6  [  ]
................................................................................... 19 Table 6-7  [
] ............................................................................................................ 20 Table 6-8  [
] ............................................................................................................ 21
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 6 List of Figures Page Figure 6-1 Time History Stresses for Transient 1 ..............................................................................
....... 9 Figure 6-2 Time History Stresses for Transient 7 ..............................................................................
..... 10 Figure 6-3 Time History Stresses for Transient 10 .............................................................................
.... 10 Figure 6-4 Time History Stresses for Transient 11 .............................................................................
.... 11 Figure 6-5  [  ] .................................................................................................. 19
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 7 1.0 OBJECTIVE The component to be qualified for EAF (environmental assisted fatigue) in this calculation is the CRDM Lower Joint. Only the inside wetted location is to be considered. A fatigue usage factor is to be calculated considering the EAF and the objective is to obtain a value that is less than 1.0.
 
===2.0 METHODOLOGY===
Calculations are performed based on the criteria below, which is based on DRAFT NUREG/CR-6909, Reference [1]. The applicable code is ASME III, 1989 edition with no Addenda Reference [2]. 2.1 General Approach to Calculating the Fen and Usage Factor Time-history stresses are provided for various transients in Reference [4]. From this information, the following procedure is used to calculate the Fen and resulting EAF usage factor.
(1)  Calculate the stress range between time point n-1 and n on the component stresses according to NB-3216.2. The stress ranges are 'ij = (n)ij - (n-1)ij. The principal stresses ( 1 ,  2 ,  3) are then calculated based on the stress ranges 'ij , and arranged such that  1 >  2 >  3. The stress intensity (SI) range is then int =  1 -  3. (2) Calculate the strain range by  n =  int  / E, where E is the Young's modulus of the material at the average metal temperature between the two time points.
(3) Calculate the strain rate by 'n =  n/t n , where t n in second is the time increment between point n-1 and n. (4) Determine whether or not an F en for step n shall be calculated. Since only the strain increment of increasingly tensile strain is considered (i.e., no negative ' is permitted), the following methodology is adopted to determine if the strain range shall be kept:
If l  3l < l  1l, the F en for step n shall be calculated based on the strain rate (Step 3) and other applicable parameters (S*, T* and O* ) according to Eqs. (2), (3) or (4);
If l_3 l>l_1 l, the strain range is excluded in the next step, and no Fen is calculated.
(5) Calculate the F en factor between the two extreme time points of pair number k using the "Multi-Linear Strain Based Method" Fen approach presented in Reference [1]. Equation (68) of  Reference [1].
(6) Calculate the overall EAF en usage factor for the location:  U en = F en , 1 U1 + F en , 2 U 2 + F en , 3 U 3 + ***. 2.2 Detailed Method of Calculation of the Fen for Lower Joint Detailed description of the F en calculation applicable to the Lower Joint:
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 8 3.0 ASSUMPTI ONS [    ]  4.0  DESIGN INPUTS The plant design, past and projected operating information was provided in Reference [3], Reference [4] and Reference [7]. 5.0 COMPUTER USAGE EXCEL spreadsheet only. 6.0 CALCULATIONS Calculations are performed based on the methodology of Reference [1]. Based on the information provided in Reference [4], the total usage factor associated with various transients including consideration of dynamic seismic events was calculated to be  [  ]  The stress time-histories are also provided in a table in this same calculation and are displayed in the following plot. 
[  ]  6.1 Transients To Be Evaluated The following transients were considered to be included in the calculation of the EAF usage factor for this component.
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 9 [  ]  The following plots show the time history loading of the transients involved. The decreases in stresses in these plots will generally correspond to the times of negative strain shown later in Tables 6-2 and 6-4.
Figure 6-1 Time History Stresses for Transient 1
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 10 Figure 6-2 Time History Stresses for Transient 7 Figure 6-3 Time History Stresses for Transient 10 Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 11 Figure 6-4 Time History Stresses for Transient 11
 
===6.2 EXCEL===
Spreadsheet Calculations The calculations necessary to perform this evaluation were done with EXCEL. The following is an explanation of the equations used in the spreadsheet.
For Sheet: "Lower Joint" Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 12 Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 13 Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 14 Table 6-1:  Input Data and Stress Difference Calculation for Transients 1 and 7
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 15 Table 6-2:  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transients 1 and 7
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 16 Table 6-3:  Input Data and Stress Difference Calculation for Transients 10 and 11
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 17 Table 6-4:  Calculation of Various Intermediate Parameters Needed to Determine Fen for Transients 10 and 11
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 18 Based on the above calculations for n and Fen
* n, the usage factor can be recalculated to account for the EAF. For Transients 1 and 7,  =  [  ]  For Transients 1 and 10,
=  [  ]  For Transients 1 and 11,  =  [  ]  [  ]  This will be used for the Transient 1 and 5 combination.
Transient 16  [  ]  The temperature loading is shown to be a maximum of
[  ]  on page 39 of Reference [6]. Conservatively using this temperature for the lower joint:
[  ]  [  ]  [  ]  [  ]  This will be used for the Transient 1 and 16 combination.
The following Tables 6-5 and 6-6 show the calculation of cycles of transients available
[  ]
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 19 Figure 6-5
[  ]  Table 6-5 
[  ]  (1) Reference [3],  (2)  Reference [7] .
Table 6-6 
[  ]  (1) Reference [3],  (2)  Reference[7].
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 20 Table 6-7 
[  ]    EAF Usage Factor Total Uen (2) = 0.420  Notes: (1) The EAF Usage Factors above were calculated with the Fatigue Design Curve from Reference [1], Table A.2.
(2) Per Section 2.1, the overall EAF en usage factor for the location is:  U en = F en , 1 U1 + F en , 2 U 2 + F en , 3 U 3 + ***. (3) [  ]  (4) [  ]
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 21 Table 6-8 
[  ]    EAF Usage Factor Total Uen (2) = 0.749  Notes: (1) The EAF Usage Factors above were calculated with the Fatigue Design Curve from Reference [1], Table A.2. 
(2) Per Section 2.1, the overall EAF en usage factor for the location is:  U en = F en , 1 U1 + F en , 2 U 2 + F en , 3 U 3 + ***. (3)  [  ]  (4)  [  ]
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 22  7.0 RESULTS Per Tables 6-7 and 6-8, the EAF usage factors have been recalculated and shown to be less than 1.0
[      ] However removing this conservatism provides a more accurate and realistic usage factor as presented in these calculations.
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page 23
 
==8.0 REFERENCES==
References identified with an (*) are maintained within Turkey Point Nuclear Power Plant Records System and are not retrievable from AREVA Records Management. These are acceptable references per AREVA Administrative Procedure 0402-01, Attachment 8. See page 2 for Project Manager Approval of customer references.
: 1. NUREG/CR-6909, Revision 1, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials", DRAFT, March 2014. 2. ASME B&PV Code Sec. III, SubSection 1989 edition with no Addenda. 3. Areva Document 38-9279661-000, [NNPARV-17-0241, NNPARV-17-0243], "Formal Transmittal of inputs for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4". 4. AREVA Document 38-9279681-000, JSPM Document 17NI0548, Rev. B, "Turkey Point Power Plant - Data for the fatigue analysis of the criticals points of latch housing and canopy joint". 5. Textbook, "Advanced Strength and Applied Elasticity The SI Version", A.C. Ugural, S.K. Fenster, 1982, Elsevier Science Publishing Co. 6. AREVA Document 33-9127371-001, "Turkey Point Plant Extended Power Uprate Control Rod Drive Mechanism Pressure Housing Assembly Appurtenances ASME Class 1". 7. AREVA Document 38-9280092-000, [NNPARV-17-0268], "Formal Transmittal of input for Areva Evaluation of Environmentally Assisted Fatigue at Turkey Point Units 3 and 4".
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page A-1  APPENDIX A: BASIS FOR PRINCIPAL STRESS CALCULATIONS A.1 Equations for Principal Stress Calculations The source equations from the text, "Advanced Strength and Applied Elasticity The SI Version",  [5] are shown here for documentation purposes.
 
Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page A-2 Document No. 32-9280712-000 TP CRDM Lower Joint Environmentally Assisted Fatigue - Non-Proprietary Page A-3 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 9Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)
PWROG-17031-NP, Rev. 0 Update for Subsequent Licen se Renewal: WCAP-15338-A A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants, August 2017 (34 Total Pages, including cover sheets)
 
PWROG-17031-NP Revision 0 Update for Subsequent License Renewal: WCAP-15338-A, "A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants" Materials Committee PA-MSC-1497 August 2017 PWROG-17031-NP Revision 0 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* August 2017 ACKNOWLEDGEMENTS WESTINGHOUSE ELECTRIC COMPANY LLC PROPRIETARY LEGAL NOTICE:
COPYRIGHT NOTICE
:
 
DISTRIBUTION NOTICE
 
PWR Owners Group United States Member Participation* for PA-MSC-1497 Utility Member Plant Site(s)
Participant Yes No PWR Owners Group United States Member Participation* for PA-MSC-1497 Utility Member Plant Site(s)
Participant Yes No
* 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.
PWR Owners Group International Member Participation* for PA-MSC-1497 Utility Member Plant Site(s)
Participant Yes No
* 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.
TABLE OF CONTENTS 1 Background and Introduction
 
2 Mechanisms of Cracking Associated with Weld Deposited Cladding
 
3 Plant Experience with Defects in and under the Weld-deposited Cladding 3.1 PWR Plants 3.1.1 Reactor Vessel Experience:  Underclad Cracks
 
====3.1.2 Reactor====
Vessel Experience:  Exposed Base Metal
: 1. Yankee Rowe Reactor Vessel
: 2. Connecticut Yankee Reactor Vessel
: 3. Watts Bar Unit 1 Reactor Vessel Inlet Nozzles
: 4. McGuire Unit 2 Bottom Head Dutchman
 
====3.1.3 Other====
Primary System Component Experience
 
3.2 BWR Plants
 
3.3 PWR Service Experience Since 1999
 
4 Effects of Cladding on Fracture Analysis
 
5 Vessel Integrity Assessment 5.1 Potential for Inservice Exposure of the Vessel Base Metal To Reactor Coolant Water
 
===5.2 Fatigue===
Usage 
 
===5.3 Acceptance===
Criteria 5.3.1 ASME Secti on XI - IWB-3500 5.3.2 ASME Secti on XI - IWB-3600
 
5.3.2.1 Criteria Based on Flaw Size 5.3.2.2 Criteria Based on Applied Stress Intensity Factors 
 
===5.4 Fatigue===
crack growth
 
Table 5-1: Fatigue Crack Growth Result for Beltline Region, Axial Flaw (Water Environment) Flaw Shape AR = l/a = 2Flaw Shape AR = l/a = 6Continuous Flaw (l/a = 100)
Table 5-2: FCG Results for Beltline Region, Circumferential Flaw in Water Flaw Shape AR = l/a = 2 Flaw Shape AR = l/a = 6 Continuous Flaw (l/a = 100)
Table 5-3: FCG Results for Inlet Nozzle to Shell Weld, Axial Flaw in Water Flaw Shape AR = l/a = 2Flaw Shape AR = l/a = 6Continuous Flaw (l/a = 100)
Figure 5-1: Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels Exposed to Water Environment [27, Fig. A-4300-2] 
 
===5.5 Allowable===
Flaw Size - Normal, Upset & Test Conditions
 
Table 5-4: Allowable Flaw Size Summary for Beltline Region - Normal, Upset & Test Conditions Figure 5-2: Allowable Flaw Size Determination - Beltline Region, Axial Flaw, AR 2:1 Figure 5-3: Allowable Flaw Size Determination - Beltline Region, Axial Flaw, AR 6:1 Figure 5-4: Allowable Flaw Size Determination - Beltline Region, Axial Flaw, Continuous
 
===5.6 Sensitivity===
of Inservice Inspections
 
===5.7 Allowable===
Flaw Size - Emergency & Faulted Conditions
 
Table 5-5: Critical Flaw Size Summary for Beltline Region - Emergency & Faulted Conditions Table 5-6: Allowable Axial Flaw Sizes for Beltline Region - Emergency and Faulted Conditions
 
6 Summary and Conclusions
 
7 References
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 10Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)  0 PWR Owners Group, PWROG-17011-NP, Rev. 0 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, November 2017 (64 Total Pages, including cover sheets)
 
PWROG-17011-NP Revision 0 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" Materials Committee PA-MSC-1500 November 2017 PWROG-17011-NP Revision 0 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* Raymond E. Schneider* November 2017 ACKNOWLEDGEMENTS WESTINGHOUSE ELECTRIC COMPANY LLC PROPRIETARY LEGAL NOTICE
 
COPYRIGHT NOTICE
 
DISTRIBUTION NOTICE
 
PWR Owners Group United States Member Participation* for PA-MSC-1500 Utility Member Plant Site(s)
Participant Yes No
* 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.
PWR Owners Group International Member Participation* for PA-MSC-1500 Utility Member Plant Site(s)
Participant Yes No
* 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.
TABLE OF CONTENTS List of Tables
 
List of Figures List of Acronyms
 
1 INTRODUCTION
 
2 BACKGROUND
 
Table 2-1: Westinghouse Domestic Plant Alpha Designation Listing Plant Alpha Designation(s)Plant(s) 
 
===2.1 DESIGN===
AND FABRICATION
 
Table 2-2: Summary of Westinghouse Domestic Plant RCP Motor Flywheel Information [
3] Group Outer Diam. (inch) Bore (inch) Keyway Radial Length (inch) Pump & Motor Inertia (lbm-ft 2) Material Type Applicable Plants (Plant Alpha Designation)
 
Figure 2-1: Example of a Typical Westinghouse RCP Motor Flywheel 
 
===2.2 INSPECTION===
 
Inspection History
 
Table 2-3: Flywheel Inspection Result from MUHP-5042 Study [2, 3] Plant Alpha Designation Plant Number of FlywheelsTotal Number ofFlywheel InspectionsTotal Number of Inspection with No Indications or Non-recordableIndications Total Number of Inspection With Recordable Indications Number of IndicationsAffecting Flywheel Integrity Table 2-3: Flywheel Inspection Result from MUHP-5042 Study [2, 3], continued Plant Alpha Designation Plant Number of FlywheelsTotal Number ofFlywheel InspectionsTotal Number of Inspection with No Indications or Non-recordableIndications Total Number of Inspection With Recordable Indications Number of IndicationsAffecting Flywheel Integrity TOTALS 57 217 729 686 43 0 Table 2-4: Summary of Recordable Indications from MUHP-5042 Study [2, 3]
Plant Alpha Designation Year Description of Recordable Indications
 
Table 2-4: Summary of Recordable Indications from MUHP-5042 Study [2, 3], continued Plant Alpha Designation Year Description of Recordable Indications
 
Inspection History Update
 
Table 2-5: Flywheel Inspection Data per [8]
Plant Number of Flywheels Total Number of Flywheel Inspections Total Number of Inspection with No Indications or Non-recordable Indications Total Number of Inspection With Recordable Indications Number of IndicationsAffecting Flywheel Integrity Total  75 81 77 4 0 Table 2-6: Flywheel Inspection Data per [8] Plant Year Description of Recordable Indications
 
===2.3 STRESS===
AND FRACTURE EVALUATION
 
Figure 2-2:  Result of Typical Westinghouse RCP Motor Flywheel Overspeed Evaluation 2.3.1 Selection of Flywheel Groups for Evaluation Table 2-7: Flywheel Groups Evaluated for Program MUHP-5043 [3] 
 
====2.3.2 Ductile====
Failure Analysis
 
Table 2-8: Ductile Failure Limiting Speed
 
====2.3.3 Nonductile====
Failure Analysis
 
Table 2-9: Critical Crack Lengths for Flywheel Overspeed of 1500 rpm (Considering LBB) 2.3.4 Fatigue Crack Growth Table 2-10: Fatigue Crack Growth Assuming 6000 RCP Starts and Stops 2.3.5 Excessive Deformation Analysis
 
Table 2-11: Flywheel Deformation at 1500 rpm
 
Table 2-12: Shrink Fit Lost at 1500 rpm [2]
2.4
 
==SUMMARY==
OF STRESS AND FRACTURE RESULTS
 
3 RISK ASSESSMENT
 
3.1 RISK-INFORMED REGULATORY GUIDE 1.174 METHODOLOGY
 
Figure 3-1:  NRC Regulatory Guide 1.174 Basic Steps Step 1:  Define the proposed change Step 2:  Perform engineering analysis
 
Step 3:  Define implementation and monitoring program Step 4:  Submit proposed change Figure 3-2:  Principles of Risk-Informed Regulation [9] Principle 1: Change meets current regulations unless it is explicitly related to a requested exemption or rule change Principle 2:  Change is consistent with defense-in-depth philosophy
 
Principle 3:  Maintain sufficient safety margins Principle 4: Proposed increases in CDF or risk are small and are consistent with the Commissions Safety Goal Policy Statement
 
Principle 5: The impact of the proposed change should be monitored using performance-measurement strategies to monitor the change 
 
===3.2 FAILURE===
MODES AND EFFECTS ANALYSIS Failure Modes
 
.
 
Failure Effects
 
===3.3 FLYWHEEL===
FAILURE PROBABILITY
 
====3.3.1 Method====
of Calculation Failure Probabilities
 
Table 3-1:  Variables for RCP Motor Flywheel Failure Probability Model No. Name Description of Input Variable Usage Type 
 
Table 3-2:  Input Values for RCP Motor Flywheel Failure Probability Model No. Name Median Distribution Uncertainty*
Flywheel Group ORadius (inch) IRadius (inch) DLength (inch)
 
Table 3-3:  Cumulative Probability of Failure over 40, 60 and 80 Years with and without Inservice Inspection (1) Flywheel Group Design Limiting Speed (rpm)  Cumulative Probability of Flywheel Failure with ISI at 4-Year Intervals Cumulative Probability of Flywheel Failure with ISI at 4-Year Intervals Prior to 10 Years and without ISI after 10 Years % Increase in Cumulative Failure Probability for Eliminating Inspections Over 40, 60 & 80 Years Over 40 Years Over 60 Years Over 80 Years Over 40 Years Over 60 Years Over 80 Years
 
====3.3.2 Sensitivity====
Study
 
Table 3-4:  Effect of Flywheel Risk Parameter on Failure Probability (Flywheel Group 10) Description of Flywheel Risk Parameter Varied Probability of Flywheel Failure after 40 years with ISI prior to and after 10 years Probability of Flywheel Failure after 40 years with ISI prior to 10 years and without ISI after 10 years 3.3.2.1 Sensitivity to Change in Flaw Detection Probability
 
3.3.2.2 Sensitivity to Initial Flaw Length
 
====3.3.3 Failure====
Probability Assessment Conclusions
 
Figure 3-3: Westinghouse PROF Program Flow Chart for Calculating Failure Probability
 
Figure 3-4: Probability of Failure for Flywheel Evaluation Group 1
 
Figure 3-5: Probability of Failure for Flywheel Evaluation Group 2
 
Figure 3-6:  Probability of Failure for Calvert Cliffs Units 1 and 2
 
3.4 CORE DAMAGE EVALUATION
 
Table 3-5:  Summary of Flywheel Analysis Parameters Westinghouse RCP/Flywheel (rpm) Calvert Cliffs RCP/Flywheel (rpm)
 
3.4.1 What is the Likelihood of the Event
 
Table 3-6:  Estimated RCP Motor Flywheel Failure Probabilities Flywheel Group and Conditions* Cumulative Probabilities of Flywheel Failure over 60 Years* Cumulative Probabilities of Flywheel Failure over 80 Years*
With ISI at 4-Year Intervals With ISI at 4-Year Intervals Prior to 10 Years and without ISI after 10 Years With ISI at 4-Year Intervals With ISI at 4-Year Intervals Prior to 10 Years, and without ISI after 10 Years 3.4.2 What are the Consequences?
 
3.4.3 Risk Calculation
 
Table 3-7:  Westinghouse RCP Motor Flywheel Evaluation Group 1 Condition Initiating Event Frequency Likelihood of RCP Motor Flywheel Failure (@80 years) Event with RCP Motor Flywheel Failure (and Core Damage Frequency given CCDP = 1.0)  (per year) (per year) With ISI after 10 Years Without ISI after 10 Years With IS After 10 Years Without ISI after 10 Years
 
Totals Table 3-8:  RCP Motor Flywheel Evaluation Group 2 Condition Initiating Event Frequency Likelihood of RCP Motor Flywheel Failure (@80 years) Event with RCP Motor Flywheel Failure (and Core Damage Frequency given CCDP = 1.0)  (per year) (per year) With ISI after 10 Years Without ISI after 10 Years With IS After 10 Years Without ISI after 10 Years
 
Totals Table 3-9:  Calvert Cliffs Units 1 and 2 RCP Motor Flywheel Evaluation Condition Initiating Event Frequency Likelihood of RCP Motor Flywheel Failure (@80 years) Event with RCP Motor Flywheel Failure (and Core Damage Frequency given CCDP = 1.0)  (per year) (per year) With ISI after 10 Years Without ISI after 10 Years With IS After 10 Years Without ISI after 10 Years 1.88E-101.90E-101.50E-081.52E-08 2.11E-102.12E-10
 
7.76E-147.87E-14Totals 1.54E-08 1.557E-08  1.30E-10  5.20E-10 
 
===3.5 CONSIDERATION===
OF UNCERTAINTY
 
====3.5.1 Initiating====
Event Frequency
 
====3.5.2 Conditional====
Flywheel Failure Probability
 
====3.5.3 Conditional====
Core Damage/Large Early Release Probability Given Flywheel Failure Event Table 3-10:  CDF Sensitivity to Variations in PRA evaluation assumptions  for  RCP Flywheel Failure Risk Assessment for Extending 10 year inspection intervals to 20 years for Flywheel operation to 80 years -(Flywheel Group 1)
Incremental Change in CDF (per Year) Risk Impact of Single Flywheel Failure Risk impact of Flywheel Failure (4 RCP Plant)
 
====3.5.4 Conclusion====
Regarding Treatment of Uncertainty 3.6 RISK RESULTS AND CONCLUSIONS
 
Table 3-11:  Evaluation with Respect to Regulatory Guide 1.174 (Key Principles) Key Principles Evaluation Response
 
4 CONCLUSIONS (Note however that the LBB exclusion for LBLOCA does not pertain to the risk assessment contained in the WCAP-15666-A [3] report, which does consider the overspeed due to LBLOCA.)
 
5  REFERENCES
 
APPENDIX A:  CALVERT CLIFFS UNIT 1 & 2 RCP MOTOR FLYWHEEL EVALUATIONS FOR EXTENSION OF ISI INTERVAL  Background and Purpose Ductile Failure Analysis
 
Nonductile Failure Analysis
 
Table A-1:  Critical Crack Length in Inches and % Through Flywheel RT NDT 0&deg;F 30&deg;F 60&deg;F Fatigue Crack Growth Excessive Deformation Analysis
/
Conclusion
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 11Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)  1 WCAP-15354-NP, Revision 1 Technical Justification for Eliminating Primary Loop Pipe Rupture as a Structural Design Basis for Turkey Point Units 3 and 4 Nuclear Power Plants for the Subsequent License Renewal Time-Limited Aging Analysis Program (80 Years) Leak-Before-Break Evaluation, August 2017 (59 Total Pages, including cover sheets)
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 12Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version) 2 SIA Leak-Before-Break Evaluation for Auxiliary Lines 0901350.401, Revision 3 Leak-Before-Break Evaluation Accumulator, Pressurizer Surge and Residual Heat Removal Lines Turkey Point Units 3 and 4, September 2017 0901350.304, Revision 3 Fatigue Crack Growth Evaluation, September 18, 2017  (149 Total Pages, including cover sheets)
 
Figure 1-1. Representation of Postulated Cracks in Pipes for Fracture Mechanics Leak
-Before-Break Analysis
 
Figure 1-2. Conceptual Illustration of ISI (UT)/Leak Detection Approach to Protection Against Pipe Rupture
 
Figure 1-3. Leak-Before-Break Approach Based on Fracture Mechanics Analysis with In
-service Inspection and Leak Detection
 
Table 4-1. Normal Operating Conditions for Leakage Evaluation Table 4-2. Operating Conditions for Critical Flaw Size Evaluation Table 4-3. Pipe Geometry Inputs for Leakage Evaluation
 
Table 4-4. Pipe Geometry Inputs for Critical Flaw Size Evaluation Table 4-5. ASME Code Strength at Normal Operating Temperatures for Leakage Calculation
 
Table 4-6. ASME Code Strength at Normal Operating Temperatures for Critical Flaw Size Calculation Table 4-7. Ramberg-Osgood Parameters for Leakage Calculation
 
Table 4-8. 
 
Table 4-10.
Figure 4-1. Schematic of Piping Model and Selected Node Points for Accumulator Line s (Loops A, B and C), PTN Unit 3 [34 , 35 , 36]
Figure 4-2. Schematic of Piping Model and Selected Node Points for Accumulator Line s (Loops A, B and C), PTN Unit 4
[39 , 40 , 41]
Figure 4-3. Schematic of Piping Model and Selected Node Points for Pressurizer Surge Line , PTN Unit 3
[38]
Figure 4-4. Schematic of Piping Model and Selected Node Points for Pressurizer Surge Line , PTN Unit 4
[43]
Figure 4-5. Schematic of Piping Model and Selected Node Points for RHR Line, PTN Unit 3
[37] Figure 4-6. Schematic of Piping Model and Selected Node Points for RHR Line, PTN Unit 4
[42]
 
Figure 5-1. Leakage Flaw Size versus Normal Operating Stress of Accumulator Line s Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM
 
Figure 5-2. Leakage Flaw Size versus Normal Operating Stress of Pressurizer Surge Line s (Pipe Side at Pressurizer End) Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM Figure 5-3. Leakage Flaw Size versus Normal Operating Stress of Pressurizer Surge Line s (Nozzle Side at Pressurizer End) Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM Figure 5-4. Leakage Flaw Size versus Normal Operating Stress of Pressurizer Surge Line (Nozzle Side at Hot Leg End) Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM Figure 5-5. Leakage Flaw Size versus Normal Operating Stress of Pressurizer Surge Line at Hot Leg End Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM Figure 5-6. Leakage Flaw Size versus Normal Operating Stress of RHR Line at Hot Leg End
 
Leakage Flaw Length (2a), inchesNormal Operating Stress, ksi10 GPM5 GPM2 GPM
 
Figure 5-7. BACs and Load Points for Accumulator Lines
 
Figure 5-8. BACs and Load Points for RHR Lines
 
Figure 5-9. BACs and Load Points for Pressurizer Surge Line s (Nozzle Side at Pressurizer End)
 
Figure 5-10. BACs and Load Points for Pressurizer Surge Line (Nozzle Side at Hot Leg End)
 
Figure 5-11. BACs and Load Points for Pressurizer Surge Line s (Pipe Side at Pressurizer End)
 
Figure 5-12. BACs and Load Points for Pressurizer Surge Line s (Pipe Side at Hot Leg End)
Figure 5-13. 10 GPM BAC Curve for Pipe/Elbow of Accumulator Line s
 
Table 6-10for a partial through
-wall crack in the RHR Line, the crack growth in the depth direction is 0.0014 inch and the crack growth in the length direction is 0.0 004 inch. This is about 0.0 009% of the 43.96 inch circumference length, and compared to the crack growth of 0.1 2% (12.50% to 12.6 2% in Table 4
-1) in the depth direction, it is relatively small. For the Surge Line, the crack growth in the depth direction is 0.
0855 inch and the crack growth in the length direction is 0.0452 inch. This is 0.113% of the 40.03 inch circumference length, and compared to the 7.6% (12.50% to 20.10% in Table 4
-1) in the depth direction, it is relatively small. Overall, for the RHR and surge lines, the partial through-wall cracks tend to grow in the depth direction and through
-wall before extending significantly in the length (circumferentially) direction.
There is no growth in the accumulator line.
Table 6-1. Accumulator Line Operating Condition Transients(1)Table 6-2. RHR Line Operating Condition Transients [
32]
 
Table 6-4. Piping Loads for Accumulator and RHR Lines Table 6-5. Accumulator Line Maximum and Minimum Transient Stresses Table 6-6. RHR Line Maximum and Minimum Transient Stresses
 
Table 6-7. Stress Range for Accumulator Line
 
Table 6-8. Stress Range for RHR Line Table 6-9. Stress Range for Surge Line
 
Table 6-10. Results of Fatigue Crack Growth Analysis
 
tm pc-CRACK  for Windows Version 3.1-98348
 
                            (C) Copyright '84 - '98 Structural Integrity Associates, Inc.
3315 Almaden Expressway, Suite 24
 
San Jose, CA 95118-1557 Voice:  408-978-8200 Fax:    408-978-8964 
 
E-mail: pccrack@structint.com
 
Linear Elastic Fracture Mechanics
 
Date: Wed Apr 14 21:22:15 2010 Input Data and Results File: ACCL_TH.LFM
 
Title: Accumulator Line through wall crack stress intensity Load Cases:
 
Stress Coefficients Case ID                    C0          C1          C2          C3  Type
 
____________________________________________________________________________
tension                19.02            0            0            0  Coeff
 
        ------Through Wall Stresses for Load Cases With Stress Coeff-------
 
Wall        Case Depth      tension
_________________________
 
0.0000        19.02 0.2530        19.02
 
0.5060        19.02 0.7590        19.02 1.0120        19.02
 
1.2650        19.02 1.5180        19.02 1.7710        19.02
 
2.0240        19.02 2.2770        19.02 2.5300        19.02
 
Crack Model: Through-Wall Circ. Crack in Cylinder Under Tension And Bending
 
Crack Parameters:
Wall thickness:        1.0000 Outside diameter:    10.7500
 
Half crack length:    2.5300 Poisson ratio:        0.3000 Tension: Co = P/(2*Pi*Rm*t)
 
Max. bending: C1 = M/(Pi*t*Rm*Rm)
All other stress coefficients are neglected.
 
          --------------------Stress Intensity Factor--------------------
Crack        Case
 
Size      tension
_________________________
 
0.0506      7.58399 0.1012      10.7281 0.1518      13.1448
 
0.2024      15.1874 0.2530      16.9932 0.3036      18.6326
 
0.3542      20.148 0.4048      21.5668 0.4554      22.9085 0.5060      24.187
 
0.5566      25.4134 0.6072      26.5958 0.6578      27.7413
 
0.7084      28.8553 0.7590      29.9424 0.8096      31.0067
 
0.8602      32.0514 0.9108      33.0795 0.9614      34.0936
 
1.0120      35.096 1.0626      36.0886 1.1132      37.0733
 
1.1638      38.0516 1.2144      39.0252 1.2650      39.9954
 
1.3156      40.9634 1.3662      41.9304 1.4168      42.8974
 
1.4674      43.8655 1.5180      44.8356 1.5686      45.8085
 
1.6192      46.7852 1.6698      47.7663 1.7204      48.7526
 
1.7710      49.7448 1.8216      50.7435 1.8722      51.7494
 
1.9228      52.7631 1.9734      53.7852 2.0240      54.8161
 
2.0746      55.8564 2.1252      56.9067 2.1758      57.9673
 
2.2264      59.0389 2.2770      60.1217 2.3276      61.2163
 
2.3782      62.323 2.4288      63.4424 2.4794      64.5468
 
2.5300      65.6637
 
End of pc-CRACK Output tm pc-CRACK  for Windows Version 3.1-98348
 
                            (C) Copyright '84 - '98 Structural Integrity Associates, Inc.
3315 Almaden Expressway, Suite 24
 
San Jose, CA 95118-1557 Voice:  408-978-8200 Fax:    408-978-8964 
 
E-mail: pccrack@structint.com
 
Linear Elastic Fracture Mechanics
 
Date: Wed Apr 14 21:21:35 2010 Input Data and Results File: RHR_TH.LFM
 
Title: RHR Line through wall crack growth Load Cases:
 
Stress Coefficients Case ID                    C0          C1          C2          C3  Type
 
____________________________________________________________________________
tension                17.38            0            0            0  Coeff
 
        ------Through Wall Stresses for Load Cases With Stress Coeff-------
 
Wall        Case Depth      tension
_________________________
 
0.0000        17.38 0.3120        17.38
 
0.6240        17.38 0.9360        17.38 1.2480        17.38
 
1.5600        17.38 1.8720        17.38 2.1840        17.38
 
2.4960        17.38 2.8080        17.38 3.1200        17.38
 
Crack Model: Through-Wall Circ. Crack in Cylinder Under Tension And Bending
 
Crack Parameters:
Wall thickness:        1.2500 Outside diameter:    14.0000
 
Half crack length:    3.1200 Poisson ratio:        0.3000 Tension: Co = P/(2*Pi*Rm*t)
 
Max. bending: C1 = M/(Pi*t*Rm*Rm)
All other stress coefficients are neglected.
 
          --------------------Stress Intensity Factor--------------------
Crack        Case
 
Size      tension
_________________________
 
0.0624      7.69576 0.1248      10.886 0.1872      13.3379
 
0.2496      15.4099 0.3120      17.2411 0.3744      18.9032
 
0.4368      20.4389 0.4992      21.8763 0.5616      23.2347 0.6240      24.5286
 
0.6864      25.7689 0.7488      26.9641 0.8112      28.121
 
0.8736      29.2453 0.9360      30.3416 0.9984      31.4139
 
1.0608      32.4656 1.1232      33.4997 1.1856      34.5186
 
1.2480      35.5247 1.3104        36.52 1.3728      37.5063
 
1.4352      38.4852 1.4976      39.4582 1.5600      40.4267
 
1.6224      41.3919 1.6848      42.355 1.7472      43.317
 
1.8096      44.2789 1.8720      45.2416 1.9344      46.206
 
1.9968      47.1728 2.0592      48.1429 2.1216      49.117
 
2.1840      50.0957 2.2464      51.0797 2.3088      52.0695
 
2.3712      53.0659 2.4336      54.0692 2.4960      55.0801
 
2.5584      56.099 2.6208      57.1265 2.6832      58.1629
 
2.7456      59.2088 2.8080      60.2646 2.8704      61.3307
 
2.9328      62.4074 2.9952      63.4952 3.0576      64.5944
 
3.1200      65.6977
 
End of pc-CRACK Output
 
tm pc-CRACK  for Windows Version 3.1-98348
 
                            (C) Copyright '84 - '98 Structural Integrity Associates, Inc.
3315 Almaden Expressway, Suite 24
 
San Jose, CA 95118-1557 Voice:  408-978-8200 Fax:    408-978-8964 
 
E-mail: pccrack@structint.com
 
Linear Elastic Fracture Mechanics
 
Date: Wed Apr 14 21:20:14 2010 Input Data and Results File: SURGE_TH.LFM
 
Title: SURGE Line through wall crack growth Load Cases:
 
Stress Coefficients Case ID                    C0          C1          C2          C3  Type
 
____________________________________________________________________________
tension                24.73            0            0            0  Coeff
 
        ------Through Wall Stresses for Load Cases With Stress Coeff-------
 
Wall        Case Depth      tension
_________________________
 
0.0000        24.73 0.3300        24.73
 
0.6600        24.73 0.9900        24.73 1.3200        24.73
 
1.6500        24.73 1.9800        24.73 2.3100        24.73
 
2.6400        24.73 2.9700        24.73 3.3000        24.73
 
Crack Model: Through-Wall Circ. Crack in Cylinder Under Tension And Bending
 
Crack Parameters:
Wall thickness:        1.1250 Outside diameter:    12.7500
 
Half crack length:    3.3000 Poisson ratio:        0.3000 Tension: Co = P/(2*Pi*Rm*t)
 
Max. bending: C1 = M/(Pi*t*Rm*Rm)
All other stress coefficients are neglected.
 
          --------------------Stress Intensity Factor--------------------
Crack        Case
 
Size      tension
_________________________
 
0.0660      11.2621 0.1320      15.9321 0.1980      19.5234
 
0.2640      22.5608 0.3300      25.2483 0.3960      27.6912
 
0.4620      29.9521 0.5280      32.0723 0.5940      34.0806
 
0.6600      35.9982 0.7260      37.8413 0.7920      39.6225 0.8580      41.3522
 
0.9240      43.0387 0.9900      44.689 1.0560      46.3092
 
1.1220      47.9043 1.1880      49.4788 1.2540      51.0368
 
1.3200      52.5815 1.3860      54.1163 1.4520      55.6438
 
1.5180      57.1666 1.5840      58.6871 1.6500      60.2074
 
1.7160      61.7294 1.7820      63.2549 1.8480      64.7857
 
1.9140      66.3232 1.9800      67.8691 2.0460      69.4245
 
2.1120      70.991 2.1780      72.5696 2.2440      74.1615
 
2.3100      75.7679 2.3760      77.3898 2.4420      79.0281
 
2.5080      80.6839 2.5740      82.358 2.6400      84.0514
 
2.7060      85.7647 2.7720      87.499 2.8380      89.2371
 
2.9040      90.9721 2.9700      92.7252 3.0360      94.4966
 
3.1020      96.2866 3.1680      98.0955 3.2340      99.9235
 
3.3000      101.771
 
End of pc-CRACK Output
 
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251
 
L-2018-004 Enclosure 4 Attachment 13Enclosure 4 Non-proprietary Reference Documents and Redacted Versions of Proprietary Reference Documents (Public Version)  3 PWROG-17033- NP, 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, October 2017 (34 Total Pages, including cover sheets)
 
PWROG-17033-NP 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" Materials Committee PA-MSC-1498, Revision 0 October 2017 PR SSURIZED W TER RE CTOR OWN RS GROUP @Westinghouse PWROG-17033-NP 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" PA-MSC-1498, Revision 0 October 2017 ACKNOWLEDGEMENTS
 
LEGAL NOTICE
 
COPYRIGHT NOTICE
 
DISTRIBUTION NOTICE
 
PWR Owners Group United States Member Participation* for PA-MSC-1498, Revision 0 Utility Member Plant Site(s)
Participant Yes No PWR Owners Group United States Member Participation* for PA-MSC-1498, Revision 0 Utility Member Plant Site(s)
Participant Yes No
* 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.
PWR Owners Group International Member Participation* for PA-MSC-1498, Revision 0 Utility Member Plant Site(s)
Participant Yes No
* 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.
TABLE OF CONTENTS 1 BACKGROUND AND PURPOSE
 
===1.1 OPERATING===
EXPERIENCE AND APPLICATION OF LATEST FRACTURE TOUGHNESS DATABASE
 
&deg;&deg;
 
2 STABILITY ANALYSIS AND FRACTURE TOUGHNESS 2.1 FRACTURE TOUGHNESS DETERMINED IN WCAP-13045
 
===2.2 FRACTURE===
TOUGHNESS BASED ON NUREG/CR-4513, REVISION 2
 
3 FATIGUE CRACK GROWTH ANALYSIS
 
4 CONCLUSIONS
 
5 REFERENCES
 
See Appendix C
 
APPENDIX A: ASME SECTION XI CODE CASE N-481 APPENDIX B: ASME SECTION XI CODE IMPLEMENTATION OF N-481
 
APPENDIX C: CAST AUSTENITIC STAINLESS STEEL (CASS) PRESENTATION AT ASME CODE MEETING}}

Latest revision as of 17:45, 5 April 2019

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