Text

Application of Reactor Vessel Surveillance Data to Calvert Cliffs Unit 1
	ML20058G175
Person / Time
Site:	Calvert Cliffs
Issue date:	11/30/1993
From:	Bowman M, Stephanie Coffin, Gryczkowski G; BALTIMORE GAS & ELECTRIC CO.
To:	;
Shared Package
ML19311B212	List: ML20058G166; ML20058G175; ML20058G184;
References
	NUDOCS 9312090119
	Download: ML20058G175 (100)
	v • d • e

{{#Wiki_filter:. f I A.:? PLICATION OF R:EACTOR VESSEL ^ SU:1VEILLANCE DATA TO CALVERT CLIFFS UNIT 1 u 1 I B A LTIMCRE G AS AND ELECTRIC l V Baltimore Gas & Electric Company November 1993 Prepared by: M.E. Bowman S.M. Cotfin G.E. Gryczkowski R.O Hardies S. A. Henry D. A. Wright 9312090119 931129 PDR ADOCK 05000317 P pyg j

- -... - ~. -. ~ - ~- t i t I t t TA BLE OF CONTENTS j Section Pace .{ 'f List of Tables.... .. iii I List of Figures... . vi List of Acronyms and Definitions.. . ix l 1.0 OBJ ECTIVE. I 6

2.0 BACKGROUND

.2

3.0 INTRODUCTION

.3 4.0 CC-1 ORIGINAL SURVEILLANCE PROGRAM. .6 -l 5.0 McG-1 ORIGINAL SURVEILLANCE PROGRAM. .7 6.0 CC-f SUPPLEMENTAL SURVEILLANCE PROGRAM. . 11 - 7.0

SUMMARY

OF CC-1 RT AND RT VALU ES. .12 j g.r y3 i Appendix A CC-1 Original Surveillance Program. .. A-1 I . B-1 l Appendix B McG-1 Original Surveillance Program.. Appendix C Environmental Equivalency Between CC-1 Clad / Base Metal Interface and McG-1 Surveillance Capsule and Comparison Between CC-1 and Far-1 Surveillance Program Results. . C-1 l Appendix D CC-1 Supplemental Surveillance Program. . D-1 i: References. . Ref-l Enclosure Reactor Vessel Weld Materials for Calven Cliffs Unit i Supplemental Surveillance Program ii

I i i List of Tables Table 3-1 CC-1 Reactor Vessel Beltline Materials. .3 i Table 4-1 , RT and RT Values Based on Best Fit Chemistry Factor and Reduced yg7 y3 Margin for CC-1 Plate D-7206-3 and Girth Weld 9-203. ..6 l Table 5-1 RT and RT Values Based on Best Fit Chemistry Factor and Reduced sg7 ns Margin for CC-1 Axial Weld Seams 2-203-A,B.C. .10 i Table 7-1 CC-1 Reactor Vessel Beltline Material Chemistry and initial RT .12 yg.r. Table 7-2 CC-1 Reactor Vessel Beltline Material RT and RT Values. .13 yg7 ns Table 7-3 CC-1 Reactor Vessel Beltline Material Maximum Attainable Fluence. .14 Table A-1 Azimuthal Location of Surveillance Capsules. Location of Surveillance l Materials. . A-5 Table A-2 CC-1 Chemical Composition of Surveillance Materials. . A-7 l Table A-3 Material for Neutron Flux Monitors. . A-9 Table A-4 CC-1 Original Sun eillance Capsule Withdrawal Schedule. .~A-10 e Table A-5 Revision I (1982)of the CC-1 Surveillance Capsule withdrawal l Schedule. . A-10 i + Table A-6 Revision 2 (1992) of the CC-1 Surveillance Capsule Withdrawal Schedule. .A-10 l Table A-7 Measured USE for CC-1 Capsule 263* Surveillance Materials. .A-13 j Table A-8 Projected USE at 32 EFPY for CC-1 Reactor Vessel Beltline Materials.. .. A-13 l Table A-9 Calculated Versus Measured

RT for CC : Capsule 263* SRM.

.. A-14 sg7 1 Table A-10 Calculated Versus Measured.RT for C C-l Capsule 263* Surveillance sg7 Material. .A-14 j Table A-Il Measured USE for CC-1 Capsule 97* Surveillance Materials. . A-18 Table A-12 Projected USE at 32 EFPY for CC-1 Reactor Vessel Beltline Materials (Revision 1). . A-19 Table A-13 Calculated Versus Measured

RT for CC-1 Capsule 97" Surveillance sg7 Material..

. A-20 Table A-14 Calculated Versus Measured.RT for CC-1 Surveillance Material s g.r Using Best Fit Chemistry Factor. . A-23 iii i ,c-,- 7, y r

List of Tables (cont'd.) Table B-1 Material for Neutron Flus Monitors. . B-2 Table B-2 McG-1 Chemical Composition of Surveillance Materials. B-5 j i Table B-3 Current McG-1 Surveillance Capsule Withdrawal Schedule. . B-6 j Table B-4 Measured USE for meg-1 Capsule U Surveillance Materials. B-7 l I Table B-5 Calculated Versus Measured

RT tbr McG-1 Capsule U Surveillance Materia!...

..... B-S Table B-6 Measured USE tor McG-1 Capsule X Surveillance Materials.. ......B-ll 4 Table B-7 "alculated Versus Measured a RT for McG-1 Capsule X Surveillance g.r Material.. .B-12 i Table B-S Calculated Versus Measured

RT Ihr MCU~I S"fV#III""C" M3t*fI"I I

g.r Using Best Fit Chemistry Factor. .B-15 Table C-1 McG-1 and CC-1 Dimensions anJ Properties. . C-8 ' I Table C-2 McG-1 and CC-1 DPA Characteristics. . C-22 Table C-3 meg-1 and CC-1 Flux and Spectral Characteristics. ..C-22 f I Table C-4 McG-1 and CC-l DPA Characteristics with a 15% Bias Applied to Each Unit. .. C-24 Table C-5 McG-1 and CC-1 Flux and Spectral Characteristics with a 15% Bias Applied to Each Unit. . C-25 i i Table C-6 Far-1 and CC-1 Dimensions and Propenies. .C-37 Table C-7 meg-1. CC-1 and Far-1 DPA Characteristics. .,.... C-40 Table C-8 McG-1, CC-1 and Far-1 Flux and Spectral Characteristics. . C-46 i Table C-9 Far-1 Chemical Composition of Surveillance Weld Material.. . C-47 Table C-10 Far-1 Surveillance Capsule Withdrawal Schedule. .. C-47 l Table C-Il Far-1 Surveillance Program Results for Weld Wire Heat 33A277. . C-48 I Table C-12 Far-1 Surveillance Program Results for Weld Wire Heat 33A277 Using Best Fit Chemistry Factor. . C-48 Table C-13 CC-1 and Far-1 Surveillance Program Results for Weld Wire Heat 33A277 Using Best Fit Chemistry Factor . C-51 f i 1 l l 1 1 Y 6

I i List of Tables (cont'd.) j Table D-1 Material for Neutron Flux Monitors. . D-2 ~ Table D-2 CC-1 Supplemental Surveillance Capsule Withdrawal Schedule... . D-4 I i 4 s 4 i i k i e l l 4 l 4 l i ) I 4 I

l i

v

4 i List of Figures 'I Figure A-1 Location cf CC-1 Original Surveillance Program Capsules... . A-3 Figure A-2 CC-1 Original Surveillance Program Capsule. . A-4 Figure A-3 CC-1 CVN Impact Specimen Compartment Assembly.. . A -u l Figure A-4 CC-1 Tensile-Monitor Compartment Assembly. .A-8 Figure A-5 Comparison between RG 1.99 Calculation and CC-1 SRM Surveillance i Results... .A-15 Figure A-6 Comparison between PsG 1.99 Calculation and CC-1 Capsule 263* Surveillance Results for Base Metal. .A-16 l ? Figure A-7 Comparison between RG 1.99 Calculation and CC-1 Capsule 263" Surveillance Results for Weld Metal .A-17 i i Figure A-8 Comparison between RG 1.99 Calculation and CC-1 Capsule 97* Surveillance Results for Base Metal. .A-21 l Figure A-9 Comparison between RG 1.99 Calculation and CC-1 Capsule 97 f Surveillance Results for Weld Metal .A-22 Figure A-10 Comparison between RG 1.99 Calculation and CC-1 Surveillance Results 3 . Using Best Fit Chemistry Factor for Base Metal. .A-24 Figure A-11 Comparison between RG 1.99 Calculation and CC-1 Surveillance Results l Using Best Fit Chemistry Factor for Weld Metal. .A-25 j Figure B-1 Location of meg-1 Original Surveillance Program Capsules. . B-3 l Figure B-2 Illustration Showing Locations of Specimens. Therma! Monitors and Dosimeters in McG-1 Surveillanec Capsules. . B-4 Figure B-3 Comparison between RG 1.99 Calculation and McG-1 Capsule U Surveillance Results for Base Metal. B-9 I Figure B4 Comparison between RG 1.99 Calculation and McG-1 Capsule U Surveillance Results for Weld Metal . B-10 Figure B-5 Comparison between RG 1.99 Calculation and McG-1 Capsule X Surveillance Results for Base Metal. . B-13 i Figure B-6 Comparison between RG 1.99 Calculation and McG-1 Capsule X l Surveillance Results for Weld Metal . B-14 Figure B-7 Comparison between RG 1.99 Calculation and McG-1 Surveillance Results Using Best Fit Chemistry Factor for Base Metal. . B-16 Figure B-8 Comparison between RG 1,99 Calculation and McG-1 Surveillance Results Using Best Fit Chemistry Factor for Weld Metal. . B-17 vi 1 m,

i List of Figures (cont'd.) l t Figure C-1 Absolute Flux versus Energy for CC-1 Capsule - Clad / Base Mett'. Interface - 1/4T Location...

-10 i

Figure C-2 Normalized Flux versus Energy for meg-1 Capsule X and CC 1 . C-I l j Capsule 97 a Figure C-3 Normalized Flux versus Energy for McG-1 Capsule X and CC-1 l Clad! Base Metal Interface. .C-l' i Figure C-4 Normalized Flux versus Energy for McG-1 Capsule X and CC-1 1/4T l Location. ...C-13 Figure C-5 Normalized Flux versus Energy for McG-1 Capsule X and CC-1 Cll Clad / Base Metal Interface. ..C-14 i Figure C-6 Normalized Flux versus Energy for meg-1 Capsule X and CC-1 Cll l 1/4T Location.. . C-15 i Figure C-7 Absolute Flux versus Energy for McG-1 Capsule X and CC-1 Capsule 97 .C-16 Figure C-8 Absolute Flux versus Energy for McG-1 Capsule X and CC-1 Clad / Base Metal Interf ace. .C-17 Figure C-9 Absolute Flux versus Energy for meg-1 Capsule X and CC-1 1/4T Location. .C-18 Figure C-10 Absolute Flux versus Energy for McG-1 Capsule X and CC-1 Cil Clad / Base Metal Interface. .C-19 Figure C-l l Absolute Flux versus Energy for McG-1 Capsule X and CC-1 Cll 1/4T i Location. . C-20 Figure C-12 DPA Versus Fluence for CC-1 Capsule 97" and Clad / Base Metal Interface. McG-1 Capsules and Far-1 Capsules. . C-27 Figure C-13 CC-1 Inlet Temperature Control Program. . C-30 Figure C-14 CC-1 EFPD versus Inlet Temperature. .C-31 J Figure C-15 McG-1 Temperature Control Program. . C-33 Figure C-16 McG-1 versus CC-1 Inlet Temperature Control Program.. . C-34 Figure C-17 Normalized Flux versus Energy for CC-1 Capsule 97a and Far-i Capsules . C-41 .m Figure C-18 Absolute Flux versus Energy for CC-1 Capsule 97 a and Far-1 Capsules. . C-42 vii

l j i t t I List of Figures (cont'd.) Figure C-19 Normalized Flux versus Energy for McG-1 Capsule U and Far-1 Capsules.... . C-43 Figure C-20 Absolute Flux versus Energy for McG-1 Capsule U and Far-1 Capsules. j .C-44 Figure C-21 Comparison between RG 1.99 Calculation and Far-1 Surveillance Results for Weld Wire Heat 33A277. . C-49 1 Figure C-22 Comparison between RG 1.99 Calculation and Far-1 Surveillance f Results Using Best Fit Chemistry Factor nor l 3 Weld Wire Heat 33A277 .. C-50 t Figure C-23 Comparison between RG 1.99 Calculation and Far-1 and CC-1 Surveillance Results Using Best Fit Chemistry Factor for l Weld Wire Heat 33A277 .C-52 j Figure D-1 CC-1 Supplemental 263 Surveillance Capsule. .. D-3 i a i i i i ? 4 i 1 l l viii

? i List of Acronyms and Definitions 10 CFR 50 Title 10 of the Code of Federal Regulations, Pan 50 i l 1/4T 1/4 Thickness of the Reactor Vessel Wall ASME American Society of Mechanical Engineers i ASTM American Society for Testing and Materials J B&W Babcock & Wilcox j n BG&E Baltimore Ga', & Electric BM-T Base Metal - Transverse Orientation i BM-L Base Metal - Longitudinal Orientation t l CC-1 Calvert Cliffs Nuclear Power Plant Unit 1 C-E Combustion Engineering CT Compact Tension CVN Charpy V-notch DOT Discrete Ordinate Transport Code dpa Displacements Per Atom i EFPD Effective Full Power Days EFPY Effective Full Power Years ENDF Evaluated Nuclear Data File EOC End Of Cycle i' l EPRI Electric Power Research Institute Far-1 Farley Unit 1 FSAR - Final Safety Analysis Report i l IIAZ Heat-Affected-Zone !!SST Heavy Section Steel Technology l i i i McG-1 McGuire Unit t .i NDTT Nil-Ductility Transition Temperature ^ NRC Nuclear Regulatory Commission IX i ? ~.

List of Acronyms and Definitions (cont'd.) NRR Nuclear Reactor Regulation NSSS Nuclear Steam Supply System PTS Pressurized Thermal Shock PWR - Pressurized Water Reactor RG Regulatory Guide RT.nr Reference Temperature Nil-Ductility Transition 3 RT Reference Temperature Pressurized Thermal Shock g3 SRM Standard Reference Material T T Reactor Vessel Inlet emperature c USE Upper Shelf Energy W Westinghouse Electric Corporation i l l J x 1

.-.~ f t i 1.0 OBJECTIVE l This document demonstrates that signincant improvement in the calculation of the embrittlement of several Calvert Cliffs Unit 1 (CC-1) reactor vessel beltline plates and welds may be obtained based on . data obtained from three reactor vessel surveillance programs. Baltimore Gas & Electrie (BG&E) will I use the surveillance data described in the following sections to modify the nil-ductility transition l reference temperature (RT,) and the Pressurized Thermal Shock reference temperature (RTy3) values for one bel *.line plate and two beltline welds, The more accurate calculations will increase the l margin of safety against brittle fracture without incurring additional costs and man-Rem dose associated j i with physical modi 6 cations to the plant. i i I l l i a 1 5 -I } 1 i e i t I 6 h I .l a I r i l 1 l

i

2.0 BACKGROUND

a i Nuclear reactor pressure vessels are designed with fracture toughness sufficiert to provide adequate j t margins of safety against brittle fracture throughout the operating life of the plant. Thus the original j construction employed thick section, low alloy steel base and weld materials which were uAerently { tough as characterized by the initial nil-ductility transition temperature (NDTT). Particular attention was given to the vessel beltline, the region that surrounds the effective height of the reactor core. This region is exposed to a relatively high level of neutron irradiation which, over time, will reduce the toughness of (i.e., embrittle) the base and weld materials. Reactor vessel surveillance programs - provide the means to measure irradiation-induced changes in the toughness properties of the beltline e materials. Surveillance capsules containing Charpy V-notch (CVN) impact specimens machined from l representative beltline base and weld material are removed periodically from the reactor vessel and { tested. This provides a direct measurement of the change in the toughness as a function of accumulated [ fluence. Neutron irradiation embrittlement of the reactor vessel beltline must be addressed for both normal operation and for limiting transients. Conservative heat-up and cool-down rates which limit pressure and temperature are adjusted to account for the irradiation-induced shift in RTgg.,. as required by Title l 10 of the Coce of Federal Regulations, Part 50 (10 CFR 50), Appendix G. The irradiation-induced shift in RT values of the beltline plates and welds may be calculated using Regulatory Guide 1.99 yg (RG 1.99), Regulatory Position 1.1. A limit is placed on the level of neutron embrittlement to ensure l vessel integrity will be maintained in the event of a postulated transient such as pressurized thermal ] shock (PTS). The measured level of embrittlement for PTS conditions, RTm, is calculated for all the j reactor vessel beltline plates and welds using the method described in 10 CFR 50.61. The values of RT.3 must remain below 270*F for beltline plates and axial welds and 300*F for beltline girth welds. g Regulatory Position 2.1 of RG 1.99 allows the use of surveillance results to more accurately calculate the values of the RTyg.r. Paragraph (b)(3) of 10 CFR 50.61 encourages the use of surveillance data to calculate the value of the RT .3 g ".Any information that is believed to improve the accuracy of the RT value significantly ns ] shall be reported to the Director, Office of Nuclear Roetor Regulation.." 4 10 CFR 50, Appendix H provides the means for a set of reactors that have similar design and operating a features to share reactor vessel surveillance data. 3 2 c

l I t

3.0 INTRODUCTION

j The RT and RT values for the CC-1 reactor vessel materials have been reported [1,19]. The Nor ns chemical composition, chemistry factor and initial RT values are summarized in table 3-1. For the i yg.r CC-1 reactor vessel beltline materials, additional information from surveillance programs exists that .j significantly improves the accuracy of the calculated values of RT and RT yg.r ns-Table 3-1 CC-1 Reactor Vessel Beltline Materials j l Weld Wire / Plate Chemistry Initial i ID Ilent Number (s) Cu (w/o) Ni (w/o) Factor RTsu'r ( F) { i 2-203-A,B,C* 12008/20291 0.21 0.88 210 -50 3-203-A,B,C 21935 0.21 0.69 179 -56 l 9-203** 33A277 0.23 0.23 121 -80 C-4351-2 0.11 0.55 74 20 D-7206-1 f ~ D-7206-2 C-4441-2 0.I2 0.64 84 -30 i 'i D-7206-3** C-4441-1 0.12 0.64 84 10 l l D-7207-1 C-4420-1 0.13 '0.54 90 10 D-7207-2 B-8489-2 0.I1 0.56 74 -10 1 ^ D-7207-3 B-8489-1 0.I1 0.53 74 -20 monitored in Duke Power Company's McGuire Unit I reactor vessel surveillance program monitored in the CC-1 reactor vessel surveillance program t l CC-1 Original Surseillance Program Two capsules from the CC-1 reactor vessel have been removed and analyzed as part of the CC-1 original surveillance program. The data from these capsules provide information on the embrittlement behavior of a CC-1 beltline plate (D-7206-3) and girth weld (9-203). The data indicate the plate material embrittles in a manner that is within la, of the RG 1.99, Regulatory Position 1.1 calculation; ] however, the weld material are much less embrittled compared with the RG 1.99, Regulatory Position 1.1 calculation. Using the surveillance data in accordance with RG 1.99, Regulatory Position 2.! provides a more accurate calculation of the embrittlement behavior for the materials monitored in the CC-1 reactor vessel surveillance program. Therefore, we propose using the surveillance data in 1 3 1 I

t -l t i accordance with RG 1.99, Regulatory Position 2.1 to modify the RT, and RT values for the ] y3 beltlinc plate D-7206-3 and girth weld 9-203. Details of the CC-1 curveillance program and results-to- ) I date are provided in Appendix A. j Surveillance data for axial weld seams 2-203-A,B,C are not available through the CC-1 original I surveillance program. However, in addition to fabricating vessels for its Nuclear Steam Supply j Systems (NSSS), Combustion Engineer.ng (C-E) fabricated vessels for Westinghouse Electric Corporation (E) designed NSSS. Therefore, surveillance materials from these W NSSS vessels l represent a potential source of data for C-E NSSS vessels on specific heats and types of vessel beltime j materials. The CC-1 reactor vessel was designed and fabricated by C-E. Duke Power Company's l McGuire Unit 1 (McG-1)is a W NSSS with a C-E fabricated reactor vessel. These vessels were l fabricated during the same period by C-E in Chattanooga, Tennessee. C-E manufactured weld seams in f each vessel using the same submerged tandem are weld process, the same heats of filler wire, the same j type of weld flux, and the same weld procedure. The weld selected by W for inclusion in the McG-1 l reactor vessel surveillance program is identical to the CC-1 axial weld seams 2-203-A,B,C. Use of the McG-1 surveillance data for these weld seams will permit a more accurate assessment of the RT, and RT values for CC-1. l y3 t McG-1 Original Surveillance Program i Two capsules from the McG-1 reactor vessel have been removed and analyzed as part of the McG-1 l original surveillance program. The data from these capsules provide information on the embrittlement l behavior of CC-1 beltline weld seams 2-203-A,B,C. The data indicate the weld material is much less embrittled compared'with the RG 1.99, Regulatory Position 1.1 calculation. Using the surveillance - data in accordance with RG 1.99, Regulatory Position 2.1 provides a more accurate calculation of the embrittlement behavior for the weld material monitored in the McG-1 reactor vessel surveillance i program. Therefore, we propose using the McG-1 surveillance data in accordance with RG 1.99, Regulatory Position 2.1 to modify the RT, and RT values for the CC-1 axial weld seams 2-203-y3 A,B,C. Details of the McG-1 surveillance program and results-to-date are provided in Appendix B. The technical justification for applying the McG-1 surveillance data to calculate the embrittlement state of the CC-1 reactor vessel is provided in Appendix C and in the Enclosure to this Attachment, " Reactor l Vessel Weld Materials for Calven Cliffs Unit i Supplememal Surveillance Program.' The results from another E NSSS reactor vessel surveillance program are also discussed in Appendix C. Alabama Power Company's Joseph M. Farley Unit 1 (Far-1) surveillance weld material is identical { (same weld wire heat. Oux typa Oi'r lot) to the surveillance weld material in the CC-1 original i surveillance program. The measured RT,.g.r shifts as a function of fluence level are very similar i f 4 .l i - - +

i .l .l between the two programs c'espite the differences in neutron environment and temperature. These i results indicate that differences in neutron environment and temperature do not noticeably affect the embrittlement behavior of a material. t CC-1 Supplemental Surveillance Program F Archive weld material from the McG-1 surveillance program was inserted in a surveillance capsule e installed in the CC-1 vessel in 1988. This material will provide supplemental surveillance data for CC-i i 1 beltline weld seams 2-203-A,B,C at two lluence levels. The first supplemental surveillance capsule j removal will occur when the capsule has accumulated a lluence approximately equivalent to the fluence l received by the first McG-1 capsule. This will provide additional assurance that use of the McG-1 surveillance data provides the most accurate calculation of embrittlement of CC-1 weld seams 2-203-A,B,C. Details ofISis program are provided in Appendix D. j i The following three sections summarize the results from the CC-1 original surveillance program, the McG-1 original surveillance program and the CC-1 supplemental surveillance program. Section 7 j s provides the more accurate projections for RT.g.r nd RT values using the surveillance data obtained s m from the CC-1 and McG-1 original surveillance programs. .i i e l } I t I 7 1 e i f 5

l ) I l 4.0 CC-1 ORIGINAL SURVEILLANCE PROGRAM ) The Calvert Cliffs reactor vessels were designed to the 1965 edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code including the winter 1967 Addenda. The original Calvert Cliffs surveillance program meets American Society for Testing and Materials (ASTM) E185-70 and 10 CFR 50, Appendix H requirements 12.3J and is described fully in Appendix j A. l Two reactor vessel surveillance capsules have been removed from CC-1 and the results are discussed in Appendix A. The results meet all pertinent RG 1.99 credibility criteria. RG 1.99, Regulatory Position l 2.1 is used to determine the best fit chemistry factor for the base metal and weld monitored in the .i program. The new chemistry factor and reduced margin term are used to calculate a more accurate RT,g7 and RTy3, table 4-1. 3 Table 4-1 l RT and RT Ts Values Based on Best Fit Chemistry Factor and Reduced Margin for yg.r I CC-1 Plate D-7206-3 and Girth Weld 9-203 i l Best Mt . R T,,.rs RT,,73 Chemistry Fluence Initial ID Factor Factor'

RT RT Margin RT g.r s3.r s37 D-7206-3 84 1.18 99 10 17 126

] 9-203 73 1.I8 86 -80 28 34 i 2 Calculated based on end-of-cycle (EOC) 10 vessel wall 11uence = 1.96 x 10" n/em I i 6 )

4 5.0 McG-1 ORIGINAL SURVEILLANCE PROGRAM 5.I' Bases for the Application of McG-1 Surveillance Data to CC-1 Weld Scams 2-203-A,B,C Duke Power Company's McG-1 has weld material identical to CC-1 axial weld seams 2-203-A.B.C in its reactor vessel surveillance program. We propose using the McG-1 surveillance data in accordance R with RG 1.99, Regulatory Position 2.1 to modify the RT and RT v lues for the CC-1 axial weld g.r m seams 2-203-A,B,C. To justlfy the use of McG-1 surveillance data for CC-1, in the following paragraphs we show the integrated surveillance program criteria of 10 CFR 50. Appendix H are met. Paragraph (II)(C) of 10 CFR 50. Appendix H states the following: An integrated surveillance program may be considered for a set of reactors that have similar design and operating features. The representative materials chosen for surveillance from each reactor in the set may be irradiated in one or more of the reactors, but there must be an adequate dosimetry program for each reactor. No reduction in the requirements for number of materials to be irradiated, specimen types. or number of specimens per reactor is perraitted, but the amount of testing may be reduced if the initial results agree with calculations. Integrated surveillance programs must ne approved by the Director. Office ut Nuclear Reactor Regulation, on a case-by-case basis. Criteria for approval include the following considerations:

1. The design and operating features of the reactors in the set must be suf ficiently similar to permit accurate comparisons of the predicted amount of radiation damage as a function of total power.

output.

2. There must be adequate arrangement for data sharing between plants.

3. There must be a contingency plan to assure that the surveillance program for each reactor will not be jeopardized by operation at reduced power level or by an extended outage of another reactor from which data are expected.

4. There must be substantial advantages to be gained. such as reduced power outages or reduced exposure to radiation as a direct result of not requiring surveillance capsules in all reactors in the 8

set. 7

Although the application of the AfcG-1 surveillance data to CC-1 weld seams 2-203-A,B,C is not part of an integrated surveillance program, the above criteria and considerations are satisfied. A detailed discussion of these criteria and technical considerations follows: CC-1 is a C-E NSSS and hicG-1 is a }y NSSS. Both are pressurized water reactors (PWRs) operating at about 550*F and 2250 psi nominal inlet temperature and pressure with low enrichment fuel (approximately 2-5% enrichment). Operating and design features of demonstrated relevancy to neutron radiation damage to re. actor vessel materials include neutron spectra, Oux and irradiation temperature, and are discussed comprehensively in Appendix C. The results indicate that irradiation environment differences between CC-1 and hicG-1 are not suf6cient to produce an observable difference in embrittlement response. Both vessels

original surveillance programs will continue as currently approved by the NRC in accordance with 10 CFR 50, Appendix H. These programs meet NRC requirements for adequate dosimetry.

Neither vessel will be impacted in terms of the number of materials to be irradiated, specimen types, or number of specimens per reactor. Both vessels' original surveillance programs will continue as currently approved in accordance with 10 CFR 50. Appendix H. The intention of tilis document is not to request approval for an integrated surveillance program, rather tojustify the use of hicG-1 surveillance data to CC-1 weld seams 2-203-A,B,C by showing the criteria for an integrated surveillance program are met:

1. The similar design and operating features of the two reactors as those features affect embrittlement response are discussed comprehensively in Appendix C.

2. The data necessary to evaluate the irradiation response of the CC-1 controlling weld is developed in the hicG-1 reactor vessel surveillance program. The data are available in the Public Document Room. BG&E and Duke Power Company personnel have been working together since 1987. The relationship will continue [4).

3. Even if the hicG-1 surveillance program were disrupted by an extended outage or period oflow power operation, the lead factor for the hicG-1 capsules permits fast accumulation of fluence and thus this would not likely present a problem with respect to CC-1 data application. In any case, sufficient data is currently available to characterize the behavior of CC-1 weld seams 2-203-A.B,C.

8

l

4. If BG&E did nothing further to reduce Gux, CC-1 weld seams 2 203-A,B,C would exceed the

] PTS screening criteria in 2004 as calculated according to 10 CFR 50.61. However, credible surveillance data demonstrates the CC-1 weld seams 2-203-A,B,C are embrittling much less than i calculated by the correlation prescribea in 10 CFR 50.61, j l BG&E has implemented significant t!ux reduction through fuel management and has evaluated other l options including annealing, shielding, Regulatory Guide 1.154 analyses, and radical fuel management changes. Annealing has been performed abroad for power reactors but has not been - attempted for U.S. designs. BG&E does not presently consider it to be demonstrated or cost-j effective for U.S. vessels. BG&E's evaluation of shielding found it to be costly, less effective than-j fuel management and to incur signi6 cant man-Rem dose for implementation. BG&E had been r pursuing the RG-1.154 analysis via the Combustion Engineering Owners Group based on favorable NRC results for CC-1 per such an analysis documented in [5]. However, concerns identined in { SECY 91-333 " Additional Requirements for Yankee Rowe Pressure Vessel Issues" has caused j BG&E to suspend pursuing this approach. l Use of plant-speci6c surveillance data has already been demonstrated 126] and provides the necessary gain in safety at the lowest cost. As a contingency. radical fuel management changes are also being pursued to further reduce neutron Gux. However, these adversely affect safety by reducing fuel thermal margins and are substantially more costly. The following additional criteria were developed and described in the Combustion Engineering Owners j Group topical report CEN-405-P, " Application of Reactor Vessel Surveillance Data for Embrittlement l Management [6]." This approach combines the credibility criteria of RG 1.99 with the concept of { integrated surveillance programs defined in 10 CFR 50, Appendix H. The topical report was originally -f submitted to the Nuclear Regulatory Commission (NRC) on December 6.1991. After incorporating i changes based on questions and comments received from the NRC, CEN-405-P. Revision 2 was l ~ submitted for review on August 6,1993. The NRC was asked to generically review this topical report l and issue an Safety Evaluation Report allowing use of the methodology contained. l The additional criteria include the following: 3 ( credibility of CC-1 surveillance program data. f credibility of McG-1 surveillance program data. j similarity of the irradiation environments, and traceability and similarity of material. } t + 9 r

I The CC-1 and McG-1 surveillance programs meet these additional criteria as discussed in Appendices A through C and in the Enclosure to this Attachment, " Reactor Vessel Weld Materials for Calvert Cliffs Unit 1 Supplemental Surveillance Program." t 5.2 Application of 31cG-1 Surveillance Data to CC-1 Weld Seams 2-203-A,B,C I I The McG-1 reactor vessel was designed to the 1971 edition of the ASME Boiler and Pressure Vessel Code including the s6mmer 1971 Addenda. The McG-1 surveillance program meets ASTM E185-73 j and 10 CFR 50, Appendix H requirements [3] and is described fully in Appendix B. l Two capsules from McG-1 have been removed and analyzed. As discussed in Appendix B, the results meet all the pertinent RG 1.99 credibility criteria. RG 1.99, Regulatory Position 2.1 is used to I determine the best fit chemistry factor for the weld material monitored in the program. The new j chemistry factor and reduced margin term are used to calculate a more accurate RT and RTm "' I g7 the CC-1 axial weld seams 2-203-A.B,C, table 5-1. Table 5-1 l RT and RT Values Based on Best Fit Chemistry Factor and Reduced Margin for CC-1 Nor p73 Axial Weld Seams 2-203-A,B,C Best Fit a R T.rs RT p rrs - Chemistry Fluence Initial ID Factor Factor

RT RT Margin -

RT 397 yp7 nut. 2-203-A,B,C 170 1.18 201 -50 28 179-Calculated based on EOC 10 vessel wall Buence = 1.96 x 10" n/cm2 ).. Y 9 10 ,n.-. ,..,.e ..,,c.,

_ ~ i i 6.01 CC-1 SUPPLEMENTAL SURVEILLANCE PROGRAM. i BG&E has established a supplemental surveillance program for the CC-1 reactor vessel to provide additional surveillance data for axial weld seams 2-203-A.B,C. Currently this program consists of one f surveillance capsule installed in CC-1 in 1988 containing CVN impact specimens fabricated from McG-1 1 archive weld material and dosimetry. i The capsule contains two sets of dosimetry and 48 CVN impact specimens (24 of the McG-1 archive f weld material and 24 of a different weld material). The capsule is constructed so that it may be i withdrawn and separated into sections. Twelve specimens of each material and one set of dosimeters f will be tested when the capsule has accumulated fluence approximately equivalent to the fluence l accumulated by the first McG-1 capsule. The remaining 24 specimens will be reinserted in CC-1. After further neutron exposure to reach an intermediate fluence level, these specimens will be i withdrawn and tested. The resultant data will be used along with the McG-1 surveillance capsule 4 results to recalculate the best fit chemistry factor in accordance with RG 1.99, Regulatory Position 2.1. i The supplemental surveillance program is limited by the low flux and lead factor of the surveillance capsule location. This prevents the supplemental material from reaching end-of-life vessel fluence. j However, data from the program will be used to verify the embrittlement observed in the McG-1 l surveillance program. This will support the application of the McG-1 surveillance data to calculate the embrittlement of the CC-1 axial weld seams 2-203-A.B,C. Details of the CC-1 Supplemental Surveillance Program are provided in Appendix D. l i J 4 a 1 11 ,v.

r 1 ) 7.0 SU51M ARY OF CC-1 RT AND RT,,rs VALUES l yg7 } This document demonstrates that significant improvement in the accuracy of embrittlement calculation of the CC-1 reactor vessel beltline materials is obtained based on surveillance data obtained from three j surveillance programs. The fol'ov ing tables provide the current calculations of RT and RT for l s g.r y3 4 the CC-1 reactor vessel beltline maierials. i l Based on the results from the CC-1 and McG-1 original surveillance programs, the chemistry factor I used to calculate the values of RT.g.r and RT were modified in accordance with RG L99, j 3 ns Regulatory Position 2.1. Table 7-1 summarizes the chemical composition, revised chemistry factor f (where applicable) and initial RT.g.r for the CC-1 beltline materials. j a 3 l j Table 7-1 CC-1 Reactor Vessel Beltline Staterial Chemistry and initial RT3 g.r l Chemistry initial l ID Cu (w/o) Ni (w/o) Factor RT3g3. (* F) 2-203-A,B,C 0.21 0.88 170' -50 i 3-203-A.B.C 0.21 0.69 179 -56 l 9-203 0.23 0.23 73* -80 D-7206-1 0.11 0.55 74 20 i l D-7206-2 0.I2 0.64 S4 -30 l D-7206-3 0.12 0.64 84* 10 D-7207-1 0.13 0.54 90 10 D-7207-2 0.11 0.56 74 -10 D-7207-3 0.I1 0.53 74 ; i i Revised based on surveillance program results per RG 1.99. Regulatory Position 2.1 1 { i t i i 12 -v .m-e,--,.m-,

3-P l In addition to the revised chemistry factor. RG 1.99, Regulatory Position 2.1 allows the reduction of i the margin term used to calculate the values of RT and RTns. In cases where the initial RT has g7 yg.r been determined and credible surveillance data is available, the margin term is cut in half (6]. Table 7-2 summarizes the values of RT.g.,. and RT based on the current EOC 10 reactor vessel wall 11uence 3 ns using the revised chemistry factor and margin term. The controlling material for CC 1 is axial weld seams 3-203-A,B C since these weld seams are calculated by RG 1.99. Regulatory Position 1.1 to have the highest level of embrittlement. f Table 7-2 CC-1 Reactor Vessel Beltline Material RT and RT,,73 Values g7 i i a RT,,73 RT,,7s j Chemistry Fluence Initial ID Factor Factor

RT RT Margin RT

{ wr g7 g7 i 2-203-A,B C 170 1.18 201 -50 28' 179 3-203-A,D C 179 211 -56 66 221 9-203 73 86 -80 28' 34 c D-7206-1 74 87 20 34 141 ) D-7206-2 84 99 -30 34 103 l D-7206-3 84 99 10 17' 126-t l D-7207-1 90 106 10 34 150 D-7207-2 74 87 -10 34 111 D-7207-3 74 87 -20 34 101 I Calculated based on EOC 10 vessel wall fluence = 1.96 x 10" n/cm 2 i Margin term cut in half as per [6] l a 1 1 W 13 n-, ,.m n.. m g

r i The fluence at which the 10 CFR 50.61 screening criteria will be met can be calculated using the i revised RT values. Table 7-3 provides the maximum attainable fluence before the 10 CFR 50.61 ns screening criteria is reached for all the CC-1 reactor vessel beltline materials. None of the CC-1 reactor I vessel beltline materials will reach this fluence before the current end-of-license or before the end of a 20-year renewal period. l; i Table 7-3 l CC-1 Reactor Vessel Beltline Material Maximum Attainable Fluence { 4 Limiting Limiting Screening Initial Chemistry Fluence Fluence 2 ID Criteria RT Margin Factor Factor (x 10 n/cm ) wr 2-203-A,B.C 270 -50 28 170 1.718 > 7.00 3 l t 3-203-A.B.C 270 -56 66 179 1.453 > 6.57 3 9-203 300 -80 28 73 4.822 > 7.00 s a j D-7206-1 270 20 34 74 2.919 j D-7206-2 270 -30 34 84 3.167 D-7206-3 270 10 17 84 2.893 } D-7207-1 270 10 34 90 2.511 i l D-7207-2 270 -10 34 74 3.324 } D-7207-3 270 -20 34 74 3.459 1 r i a i s 4 l -J i 4 0 l i e 14

Y J t e l . I APPENDIX A I i CC-1 ORIGINAL SURVEILLANCE PROGRAAI l l 1 1 A-1 m=.

APPENDIX A CC-1 ORIGINAL SURVEILLANCE PROGRAM Contents and Location iae CC-1 original sdrveillance program consists of six surveillarce capsules attached to the reactor vessel inside wall; each capsule contains mechanic.d specimens, dosimetry and thermal monitors [7,8). The mechanical specimens (CVN and tensile) were fabricated from material considered representative of the CC-1 re tetor vessel beltline. A pre-irradiation (baseline) evaluation of the strength and toughness of the surveillance materials was performed [9]. 1 ASTM E185-70 recommended the surveillance program material be representative of the reactor vessel beltline materials. The ASTM E185-70 criterion suggested using the plate with th. highest NDTT, es determined by the drop-weight test. as the source for base metal and heat-affected-zone (HAZ) materials. Two of the Unit I plates (D-7206-1 and D-7206-3) had an NDTT of 0*F. Therefore, these J twe plates were evaluated against a second selection criteria: the highest temperature at the 30 ft-lb CVN energy level. Based on this criteria. plate D-7206-3 was selected. The surveillance weld selected was the same weld wire heat / flux combination (33A277/1092) used in girth weld 9-203 of the reactor vessel beitline. This too was considered representative material and the selection of the beltline girth weld m sterial was the general practice for C-E surveillance programs. Details of the baselit e evaluation results can be found in 19,10]. The surveillance materials are contained in capsules positioned adjacent to the reactor vessel inside wall so the irradiation conditions (temperature, neutron spectrum and flux) for the capsule are very similar to those of the reactor vessel. The capsules are bisected by the midplane of the core and are positioned in capsule holders azimuthally at locations near the regions of maximum flux, figure A-1. The six surveillance capsules contain four CVN specimea compartments and three tensile specimen compartments, figure A-2. Table A-1 provides the placement of CVN and tensile specimens within each capsule compartment 7y material and specimen type. Specific identification and coding of individual test specisnens is prouded in [7]. Table A-2 provides the chemical composition of the plate and weld surveillanc : materials. Each Charpy impact compartment has 12 impact test specimens in a 4H x 3W x 1D array with the notches facing the reactor core, figure A-3. All of the original survaillance capsules contain base metal, weld metal and H AZ CVN specimens. Two of the original capsules (104" and 263* azimuthal positions) also contain CVN specimens fabricated from Standard Reference Ma'erial (SRM). The SRM CVN specimens were made from the Heavy Section Steel Technology (HSST) 01 reference plate. A-2 - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ =

180 '---~~- ' Outlet Nozzle ] '/ s / l l h C sule /' ~~~ \\ i, ei f Core Shrotal s ' Cal sule Capsule % Core Support llarrel Reactor Vessel 0 , Ca; si e --i - -d- - - Capsule ' ~~ ' Capsule 277 j k 83 if ,/ 's Capsule N \\ \\s / 284 i 's ? 'O ~ ~j l 5 i + - Core f;"arre [/[/ Reactor i 8 Vessel \\ / \\ ~ - - - - - -,i O Elevation Enlarged Plan View View Figure A-1 I,ocation of CC-1 Original Surveillance Prograin Capsules A-3

r Ai ~1.ock Assembly 4 k s.' l > Wedge Coupling Assembly l l Tensile-Monitor Companment 1 i N / > CVN Companments 2 & 3 s \\ N Tensile-Monitor Companment 4 9 / N w N\\ > CVN Companments 5 & 6 N t I / I N Tensile-Monitor Companment 7 \\ N T Figure A-2 4 CC-1 Original Surseillance Program Capsule i A-4 --y

I Table A-1 i Azimuthal Location of Surveillance Capsules, Location of Surveillance Materials Azimuthal Position: Compartment 83* 97" 104* 263* 277* 284* d 1 HAZ HAZ HAZ HAZ HAZ HAZ (3) (3) (3) (3) (3) (3)- l Tensile Tensile Tensile Tensile Tensile Tensile j 2 HAZ HAZ HAZ HAZ HAZ HAZ l (12) (12) (12) (12) (12) (12) CVN CVN CVN CVN CVN CVN l l 3 BM-T BM-T SRM SRM BM-T BM-T (12) (12) (12) (12) (12) (12) CVN CVN CVN CVN CVN CVN 4 BM-L BM-L BM-L BM-L BM-L BM-L I (3) (3) (3) (3) (3) (3) Tensile Tensile Tensile Tensile Tensile Tensile I 5 BM-L BM-L BM-L BM-L BM-L BM-L (12) (12) (12) (12) (12) (12) CVN CVN CVN CVN CVN CVN 6 W e!d Weld Weld Weld Weld W eld (12) (12) (12) (12) (12) (12) j i CN N CVN CVN CVN CVN CVN 7 Weld Weld Weld Weld Weld Weld (3) (3) (3) (3) (3) (3) Tensile Tensile Tensile Tensile Tensile Tensile (# of Specimena) Base Metal - Transverse Orientation BM-T = Base Metal - Longitudinal Orientation BM-L- = 4 A-5

j j wWedge Coupling -- End Cap / Charpy Impact Specimens '#,\\ Spacers ? ,s N i N ,/ ~i % / Rectangular Tubing t Wedge Coupling -- End Cap g l Figure A-3 CC-1 CVN Impact Specimen Compartment Assunbly A-6

i i J Table A-2 I ' CC-1 Chemical Composition of Surveillance Materials i ~ Element Plate Material Weld Material l 4 Carbon 0.26 0.15 Manganese 1.29 1.05 i Phosphorus 0.011 0.014 i Sulfur 0.016 0.013 Silicon 0.24 0.20 1 Nickel 0.64

0. I 8 Chromium 0.08 0.06 Molybdenum 044 0.55 Copper 0.12 0.24 Columbium 0.01 0.01 Boron 0.0004 0.0001 Cobalt 0.00S 0.003 Aluminum 0.022 0.002 i

Tungsten 0.01 0.01 i Titanium 0.01 0.01 Arsenic 0.01 0.01 !7 Tin 0.005 0.002 Zirconium 0.001 0.001 { Vanadium 0.001 1 0.003 i t i 4 The changes in the impact properties of the reference material provide data for correlating the results of. i j 2 CC-l's surveillance program'with the results from experimental irradiations and other reactor u surveillance piograms which use the same SRM specimens 18]. Each tensile specimen compartment contains three tensile specimens, a set of flux monitors, and a set of-I temperature monitors, tigure A-4. One set of nine flux monitors was installed in each tensile specimen compartment, table A-3 [8). Flux i monitor chemistry is provided in reference 17]. i The temperature monitors are alloy wires with various melting temperatures sealed inside quartz tubes. A stainless steel weight sitting on top of each wire drops to the bottom of the tube when the melting point of the alloy is reached. The alloy assemblies melt at 536,558,580, and 590"F. This provides an i estimate of the minimum and maximum temperatures to which the specimens have been exposed. l i l A-7 t }

Wedge Coupling - End Cap ^: Stainless Steel Tubing ~ Cadmium Shield Flux Spectmm Monitor Threshold Detector Cadmium Shielded Flux Monitor Housing [ Stainless Steel Tubing Flux Spectrrn Monitor - y i Threshold Detector l v Temperature Monitor Quanz Tubing \\ = Temperature Monitor j Weight Low Melting Alloy Hbusing i b Tensile Specimen {Q~ Split Spacer Jg* Tensile Specimen Housing -N Rectangular Tubing 4 / Wedge Coupling - End Cap Figure A-4 CC-1 Tensile 41onitor Compartment Assembly A-8

~ ,1 i Table 4-3 i Staterial for Neutron Flux 31onitors l i l i l Threshold Energy i Staterial (MeV) llatf-Life 1. ( Sulfur 2.9 14.3 days l l Uranium

0.7 28 years j

l. Iron 4.0 314 davs Nickel ** 5.0 71 dais l Copper ** 7.0 5.3 yeIsrs Titanium 8.0 84 days i Cobalt Thermal 5.3 years f

cadmium shielded and bare
- cadmium shielded A.2 Installation And Withdrawal Schedules l

The original CC-1 withdrawal schedule is in table A-4 During the first CC-1 capsule withdrawal, capsule 263* was inadvertently removed instead of capsule 97* Since these two locations are approximately isofluent, the material property changes were expected to be similar and we informed j '5 NRC of our intent to update the Final Safety Analysis Report (FSAR) withdrawal schedule 111). NRC agreed with our assessment [12] and the withdrawal schedule was modified in 1982. table A-5. I In 1991, BG&E submitted a request to the NRC to amend the CC-1 withdrawal schedule betore the 1992 refueling outage as required by 10 CFR 50 Appendix H [13]. BG&E proposed to pull the 97* capsule in lieu of the 104* capsule to provide specific dosimetry information by choosing a location approximately isofluent with the 263* ex-vessel dosimetry program. Also, SRM material should be tested at widely separated fluence values. Pulling the 104* capsule would have removed the last SRM-capsule too soon. NRC approved this change [14] and the schedule was modified, table A-6. l A-9 I i

r i 1 2 i Tahle A-4 r i CC-1 Original Surveillance Capsule Withdrawal Schedule Projected l Azimuthal Withdrawal Capsule Fluence 2 Position Year (x 10 n/cm ) t 97 5 (1979) 0.36 104 14 (1988) 0.88 l 284 23 (1997) 1.5 l 263 30 (2004) 2.2 i 277 35 (2009) 2.5 83 40 (2014) 3.1 i I Table A-5 'i Resision 1 (1982) of the CC-1 Surveillance Capsule Withdrawal Schedule t Actual Azimuthal Lead Withdrawal Capsule Fluence .f Position Factor Year EOC EFPY (x 10 n/cm ) 'l 2 263 1.43* 5 (1979)* 3* 2.94* 0.6 104 14 (1988) 284 23 (1997) } 97 30 (2004) j 277 35 (2009) 1 83 40 (2014) l Table A-6 Revision 2 (1992) of the CC-I Surveillance Capsule Withdrawal Schedule l 1 Actual Azimuthal Lead Withdrawal Capsule Fluence Position Factor Year EOC EFPY (x 10 n/cm') 263 1.43* 5 (1979)* 3* 2.94* 0.62* 97 1.34* 18 (1992)* 10* 11.07* 2.64* i 284 23 (1997) 104 30 (2004) 277 35 (2009) 83 40 (2014) effective full power years [ EFPY = actual value l = t f A-10 I m. r

i A.3. Projected Use Of Data i The original surveillance program monitors the changes in beltline material strength and toughness due l to neutron embrittlement. The program also provides dosimetry to benchmark the CC-111uence model. i Capsule fluence is used to compare measured values of

RT with RG 1.99 calculations. Results yg.r from the CC-1 original surveii!ance program must meet the following credibility criteria of 10 CFR 50.

Appendix H, E185 and RG 1.99 as described in [6]. { t 1. Temperature l l ASTS1 E185-82 requires the maximum exposure temperature of the surveillance j program materials (indicated by the temperature monitors) remain within 25*F of the expected capsule exposure temperature (indicated by operating temperature). The CC-1 j operating temperature is nominally 550 F. Therefore, the 536 F monitor must melt I while the 580"F and 590"F monitors must not melt. J 2. Upper-Shelf Energy (USE) ] The determination of the USE shall be unambiguous. The measured USE for the surveillance materials shall be greater than 50 ft-lbs. Using figure 2 from RG 1.99, the projected USE for all the beltline materials shall remain above 50 ft-lbs based on the 1/4 thickness of the reactor vessel wall (1/4T) end-of-life fluence. 3. Standard Reference Slaterial (SRN1) l l The measured 30 ft-Ib shift shall not exceed the 4RT calculated using RG 1.99 by i gg7 more than 2a, (34*F for base metal). 4. Repe ",ig The capsule withdrawal and test results must be submitted to NRC within one year of capsule withdrawal unless an extension is granted by the Director. Nuclear Reactor Regulation (NRR) as required by 10 CFR 50. Appendix H. 5. Calculated Versus Sleasured 30 ft-lbs Shift The determination of the 30 ft-lbs temperature shall be unambiguous. The measured 30 ft-lb shift shall not exceed the a RT calculated using RG 1.99, Regulatory Position yg7 1.1 by more than 2o, (34 F for base metal,56*F for welds). A-lI

i 6. Dosimetry Measurements and Fluence Calculations l 1 e The dosimetry measurements should be used to obtain the fluence at the capsule. Results should also be used to ensure the CC-1 fluence model is accurate, i i 7. Best Fit Chemistry Factor { A minimum of two sets of data shall be used to perform an analysis per RG 1.99. { Regulatory Position 2.1. The CC-1 surveillance data measurements shall be within 10, j of the 4 RT calculated using RG 1.99, Regulatory Position 2.1 (17*F for base yg.r metal,28*F for welds). If the fluence range is equal to or greater than two orders of l magnitude, the measurements shall not exceed 20, (56*F for welds 34*F for base metal). i A.4 Discussion Of Results l l Capsule 263* 'i Capsule 263* was removed from CC-1 in ' pril 1979 at the EOC 3 atter 2.94 EFPY of reactor l A operation. The capsule contents were analyzed by Battelle Columbus Laboratories 115] and j results reported to the NRC [16]. l l The data obtained from capsule 263* met all the original surveillance program credibility f criteria: i i l. Temperature The 536*F monitor melted completely, the 558*F monitor was slumped, the 580 F and 590"F monitors did not melt. 2. Upper-Shelf Energy (USE) j i The determination of the USE was unambiguous. The measured USE is above 50 ti-lbs, table A-7. Based on a projected 1/4T tluence of 2.6 x 10 n!cm at 32 EFPY. the 2 \\ USE of all the beltline materials will remain above 50 ft-lbs, table A-8. i -12 i A

i 1 i Table A-7 Measured USE for CC-1 Capsule 263* Surveillance Materials i M aterial Measured USE BM-L 115 Weld 119 Table A-8 Projected USE at 32 EFPY for CC-1 Reactor Vessel Beltline Materials initial USE Copper Calculated Decrease Calculated USE i Material (ft-lbs) - (w/o) (%) (ft-ths) 2-203-A,B,C i10 0.21 44 62 3-203-A,B,C TBD* 0.21 44 TBD 9-203 160 0.23 48 83 D-7206-1 90 0.1I 26 67 D-7206-2 81 0.I2 27 59 D-7206-3 112 0.12 27 82 D-7207-1 77 0.13 28 55 D-7207-2 90 0.1 I 26 67 D-7207-3 81 0.I1 26 60 To Be Determined [17] 3. Standard Reference Material (SRM) The measured 30 ft-lb shift of 88 ft-lbs is less than the RG 1.99. Regulatory Position 1.1 calculation by 31 *F, table A-9 and Ogure A-5. This satisfies the criterion that the measured shitt be no greater than 2o, above the RG 1.99, Regulatory Position 1.1 shift calculation (34*F for base metal). A-13

l t l l Table A-9 l Calculated Versus hicasured a RTsur "' I CC-1 Capsule 263* SR31 l Calculated 31easured Calculated - 51easured l Chemistry Fluence 30 ft-lbs shift 30 ft-ths shift 30 ft-lbs shift 31aterial Factor Factor (*F) (* F) - ("F) -i -t SRh1 137 0.866 119 88 31 i .t l 4. Reporting l The capsule was withdrawn in April 1979 and reported in February 1981. The requirement for the results to be reported within one year did not exist until after the l capsule analysis was ;ompleted. t 1 5. Calculated Versus hieasured 30 ft-lbs Shift i I The determination of the 30 ft-lbs temperature was unambiguous. The measured 30 tt-lb shift for the base metal is less than the RG 1.99, Regulatory Position 1.1 calculation by 13*F, table A-10 and figure A-6. The measured 30 ft-lb shift for the weld metal is less than the RG 1.99, Regulatory Position 1.1 calculation by 45"F, table A-10 and figure A-7. These results satisfy the criterion that the measured shifts be no greater than 2a, above the RG 1.99, Regulatory Position 1.1 shift calculation. Table A-10 Calculated Versus hicasured a RT for 337 CC-1 Capsule 263* Surveillance 51aterial Calculated hicasured Calculated - 51easured Chemistry Fluence 30 ft-ths shift 30 ft-lbs shift 30 ft-lbs shift 51sterial Factor Factor (* F) (*F) (* F) BM-L - 84 0.866 73 60 13 Weld 120 0.866 104 59 45 A-14

300 Cci AI (IISsy g1) ntis,

acto, 137 C,{,ulepg,,

Pluegce, $i]"2 x toi,"'ev Ef0 ~

o.gg p,'hl",*l'ac

~ ~ Ate ! my^'*r

88 't-lbs

~~ ~~~ Q 10 0. gym ~~ - 102~* 2E 120 ~ 5" ~~ E N ALCULATI C o ?.: U E 1.99 RG ~ ~ ~ ~ ~ J1. l R 0= r i i i i t-i i i i 1 i i l .1 1 6 bg fluence (x 10E19 n/cm2) Figure A-5 Comparison fictween RG 1.99 Calculation and CC-1 SRM Surveillance Results A-15

300 CC-1 Base Metal Chemistry Factot = 84 Capsule 263* s 2 Fluence = 0.62. x 10" n/cm 80 - Fluence Factor = 0.866 Measured ART = 60 It-lbs myr t;' 18 0 - u &u z O 12 0.. 34F ~~~ 1.99 4 RG TION gg ' ALCULA C 1.99 JL4E RG / ~ ~ JLG 1.99.: O I i i l'y L n i .1 1 6 log fluence (x 10E19 n/cm2) Figure A.6 Comparison Between RG 1.99 Calculation and CC-1 Capsule 263* Surveillance Results for liase Metal A-16 1 - _ _._ _ _ _. _ _ _ _____. _. _. -. _ _, _ _ _ _ _, _

300 CC-1 Weld Metal Chemistry Factor = 120 Capsule 263* Fluence = 0.62 x 10" n/cm' 240 - Fluence Factor = 0.866 ^!easu,cq ' R7'%r

59 0-Ibs

~~~~~~ ~~ $0 ~ %k D 22 2.1@E. _ D ff 120 ~~ ~ R L-G ~~- TION UI. A CALC 1.99 Q 00 ~ RG 99 __5J _____ F 1 R_G. 0~~ i i I i i'q 1 i l i .i 1 G bg fluence (x 10E19 n/cm2] Figure A-7 Comparison lictween 1(G 1.99 Calculation and CC-1 Capsule 263* Surveillance Itesults for Weld Metal ' A - 17

6. Dosimetry 31easurements and Fluence Calculations The dosimetry from the capsule was used to calculate the fluence at the capsule. A i fluence model for CC-1 was developed. and the dosimetry results from capsule 263* agreed with the model. ] i 7. Best Fit Chemistry Factor -l l 4 l Not applicable since only one data set is available. I i I Cancule 97* Capsule 97* was removed from CC-1 in April 1992 at the EOC 10 after 11.07 EFPY of reactor I operation. Babcock & Wilcox (B&W) analyzed the capsule contents and C-E provided the 4 fluence analysis [18]. The results were reported to the NRC in June 1993 [191 I I The data obtained from capsule 97* met all the original surveillance program credibility criteria: 1. Temperature The 536*F and 558*F monitors mehed completely while the 580*F and 590*F l I l monitors did not melt. i 2. Upper-Shelf Energy (USE) J t The determination of the USE was unambiguous. The measured USE remained above 2 50 ft-Ibs, table A-11. Based on a projected 1/4T tluence of 1.97 x 10* n/cm at 32 EFPY, the USE of all the beltline materials will remain above 50 ft-lbs, table A-12, Table A-Il 'l Steasured USE for CC-1 Capsule 97* Surveillance Staterials Staterial 31easured USE BM-L 102 BM-T 84-i Weld 106 ( i A-18 { i f

- -=. -. - _. i Table A-12 l Projected USE at 32 EFPY for CC-1 Reactor Vessel lleltline Materials (Resision I) Initial USE Copper Calculated Decrease Calculated USE Material (ft-lbs) (w /o) (%) - (ft-lbs) i 2-203-A,B,C 110 0.21 42 64 l 3-203-A,B,C TBD* 0.21 42 TBD l 9-203 160 0.23 44 90 l \\ D-7206-1 90 0.1 I 24 68 j D-7206-2 81 0.I2 25 61 { D-7206-3 112 0.12 25 84 l D-7207-I 77 0.13 26 57 r D-7207-2 90 0.1I 24 68 l 4 l D-7207-3 81 0.1I 24 62 i To Be Determined [17] l 1 3. Standard Reference Material (SRM) I I Not applicable since there was no SRM in this capsule. l l -i 4. Reporting The capsule was withdrawn on Apri! 23,1992. DG&E requested a 60 day extension of the reporting due date [20]. NRC approved the extension 121] and the final results were i reported to the NRC on June 22,1992 [19). 4 5. Calculated Versus Measured 30 ft-lbs Shift The determination of the 30 ft-lbs temperature was unambiguous. The measured 30 ft-lb shift for the base metal is greater than the RG 1.99, Regulatory Position 1.1 calculation by 2 to 5'F, table A-13 and tigure A-8. The measured 30 ft-lb shift for the i weld metal is less than the RG 1.99. Regulatory Position 1.1 calculation by 58'F, table \\ i ) i e A-19 i . ~

l i I A-13 and figure A-9. These results satisfy the criterion that the measured shifts be no j greater than 2a, above the RG 1.99. Regulatory Position 1.1 shift calculation. Table A-13 Calculated Versus aleasured a RT,.r "' I CC-1 Capsule 97* Suneillance Staterial -t Calculated afeasured Calculated - Sleasured Chemistry Fluence 30 ft-lbs shift 30 ft-lbs shift - 30 ft-lbs shift Staterial Factor Factor ('F) (*F) (' F) - BM-L 84 1.260 106 108 -2 BM-T 84 1.260 106 111 -5 Weld 120 1.260 151 93 58 i i 1 6. Dosimetry afeasurements and Fluence Calculations The dosimetry from the capsule was used to ca!culate the fluence at the capsule. A i fluence model for CC-1 was developed. and the dosimetry results from capsule 97* l i agreed with the model. i l 4 l f 4 I i 1 l i 6 1 l A-20

300 CC-1 Ilase hietal Chemistry Factor = 84 Capsule 97* 2 Fluence = 2.64 x 10 n/cm 240 - Fluence Factor = I.260 hicasured ART = 108 (Uht-L) and lII (Ilht-T) ti-Ibs yg7 t;' 18 0 - m 1 tr G

8e 120 ~

24F 4 8Q 1.99-- gg. ATION CALCUL 1.99 RG "9 l:29.= 3.9E 0-I i l' I gm I i i T I .I 1 6 bg fluence (x 10E19 n/cm2] Figure A-8 Comparison Iletween RG 1.99 Calculation and CC-I Capsule 97* Surveillance Results for Itase Metal A-21 .. ~. -.. _ _..,

300 CC-1 Welti Metal 'Orist,,,,actog.,12n C '#"le 97, se t PtyCIIce, 24 k lo leni S$O ~ Pty' Hee pacto,

I,gg y; easyd < /tj"r
93 %s NO u

} 56F 4 4 J 99 120 ~ ~ ~M -g N TI ULA CALC 1.99 60, RG o .99 3c r 1 i n i i i i i i i ry# / e 0 ~ / ,f log fluence (x 10E19 n/cm2) Figure A-9 Contparison lictween RG 1.99 Calculation and CC-l Capsule 97" Surveillance Results for Weld Aletal u -22 A

r i i ii 7. Best Fit Chemistry Factor i i The best fit chemistry factor was calculated using RG 1.99, Regulatory Position 2.1. } The measured 30 ft-lb shift for the base metal at the low fluence value is 13*F less than the RG 1.99, Regulatory Position 2.1 calculation while the high fluence values are 2 and 5*F greater, table A-14 and tigure A-10. The measured 30 ft-lb shift for the weld j metal at both tluence values are within 4*F of the RG 1.99, Regulatory Position 2.1 shitt calculation. table A-14 and figure A-11. These results satisfy the criterion that the measured shifts be within la, of the RG 1.99, Regulatory Position 1.1 shift calculation. l t Table A-14 1 Calculated Versus hicasured RTg.g. for l CC-1 Surveillance Staterial e Using Best fit Chemistry Factor Best Fit Calculated Steasured Calculated - Sleasured Chemistry Fluence 30 ft-lbs Shift 30 ft-lbs Shift 30 ft-lbs Shift l Staterial Factor Factor (* F) (* F) (* F) i Bht-L 84 0.866 73 60 13 Bht-L 84 1.260 106 108 -2 Bht-T 84 1.260 106 111 -5 l Weld 73 0.S66 63 60 3* Weld 73 1.260 92 94 -2*

Because the copper and nickel content of the surveillance weld differs from that of the vessel weld, I

the measured values of a RT'S7 are adjusted by multiplying them by the ratio of the chemistry facj for the vessel weld to that of t e surveillance weld. .j -i I A-23 )

300 CC-1 Dase Metal Best Fit Cliemistry Factor = 84 Capsules 263* and 97* 240 - C 18 0 - v U C 4 e 20 - ~~~ IlG l.99 + ~17F ~ - D-W ~~ TION 60 ~ CALCULA R ' 1.99 7F . ~ J1. 9~1~.9~9 1 0-- i i i i 1-i i i i i i i i- .1 1 0 bg fluente (x 10E19 n/cm2) Figure A-10 Comparison lietween RG 1.99 Calculation anci CC-1 Surveillance Results Using liest lit Clicinistry Factor for Ilase Nietal A-24 ~ ~ ~ .. - _ _ ~., -. _. - -.

E r 300 CC-1 Weld Metal liest Fit Cliemistry I actor = 73 Capsules 263* and 97* 240 - tL' 00 - v C$ e b 12 0 - no 1.99 4.zor 60 - C ALCUl. AT ION RG 1.99 !39 W9:.1E 0-i N $) r I f ..I I 6 log fluence (x 10E19 n/cm2) Fi;;ure A.II Comparison fictween RG 1.99 Calculation nuti CC-f Surveillance Results Using flest Fit Cliemistry Factor for Welti Metal A-25 . ~

t t i i i i 4 5 f a . i i n APPENDIX B r 31cG-1 ORIGINAL SURVEILLANCE PROGRAAI i ? l t i 4 4 1 a B-1 ---e n ---,x .,c

i i i APPENDIX B SicG-1 ORIGINAL SURVEILLANCE PROGRAM f 11.1 Contents and Location j The McG-1 surveillance program consists of six surveillance capsules located in guide baskets welded - I to the outside of the neutron shield pads [22]. A plan view showing the location and dimensional i spacings of the capsules with relation to the weld seams, neutron shield pads, and vessel core is shown j in figure B-1. l The six capsules each contain CVN, tensile, bend bar and %T compact tension (CT) specimens; ] dosimetry and thermal monitors. Figure B-2 illustrates the McG-1 surveillance capsule design, 5 Specific identification and coding of individual test specimens is provided in [22]. Each of the three i tensile specimen compartments contain three tensile specimens. Each of the six CT compartments has l two CT specimens. There i; one bend bar specimen in each surveillance capsule. l The ten CVN impact compartments contain 6 impact tes: specimens each with the notches tacing each other. Three of the CVN compartments each have a set of flux monitors and a temperature monitor. Table B-1 details the dosimetry included in the surveillance capsule [22]. The temperature monitors are j wires fabricated from eutectic alloys of various melting temperatures sealed inside Pyrex tubes. The j alloy assemblies melt at 579'F and 590* F to provide an estimate of the maximum temperature to which j the specimens have been exposed. Tanle B-2 provides the chemical composition of the plate and weld ] surveillance materials. j i Tahle B-1 i i Material for Neutron Flux Monitors Threshold Energy j Material (MeV) Italf-Life Copper 4.7 5.3 years i Iron 1.0 312 days Nickel 1.0 71 days Uranium

0.4 30.1 years Neptunium 0.08 30.1 years Cobalt-Aluminum" Thermal 5.3 years cadmium shielded cadmium shielded and bare Il-2

\\ 0 Reactor Vessel 7 l Core Barrel 1 l Neutron Pad v l 1 i U 56 56* 270 - x*,>N, - 90 / ' 58.5 /- // N \\ Y l W x i I N! i i I 1 180" l I l Figure B-1 1;ntion of McG-1 Original Suncillance Program Capsules B-3

I _J Vi!SSEL Wall. CORE n t '..psule t I Np " 2 l12 B ('sniqutt C4wnp4Lt Ctenpm,1 O gnpat t i h wuncitt Ownpact Caenpett 3r flerat il ar 4 tmde 'I rnsu mi Tcusim Otupy O urpy Charry 'I cusu m Tensum G:irpy Cloq.y lita l 'I nmie (harpy Charpy Charpy Garry Garpy Tentim Tuisum ~l en sile I Ji Ji di Cu --,,- 4-.- Al.15% C Cu -o,- Al.15% Gi c, _,,- - +-- Al.15% Co 1:e --o-le - -o-Ie --s. 579*F ~ * ^3' 3 5% C"(CI) 590*1: 4--.- Al.15% Co (Cd) 579'l: 'e-Al.15% Co (Cd) ~' l Mantor '_ _t-Ni M< entor '__t - NI Mminor '__t Ni t i l Center Region d Vessel i To Top of Vessel To llottom of Vessel Figure 11-2 litustration Showing Locations of Speciniens, Tliennal Monitors anti Dosinieters in McG-1 Surveillance Capsules IM

( Table B-2 McG-1 Chemical Composition of Surveillance 31aterials Element Plate Material Weld Material Carbon 0.21 0.10 Manganese 1.26 1.36 Phosphorus 0.010 0.011 Sulfur 0.016 0.008 Silicon 0.23 0.24 Nickel 0.60 0.88 Chromium 0.068 0.04 Molybdenum 0.57 0.55 Copper 0.087 0.21 ' Columbium < 0.001 < 0.01 Boron < 0.003 < 0.001 l Cobalt 0.016 0.014 Tungsten <0.001 < 0.01 Titanium 0.005 < 0.01 j Arsenic 0.008 0.009 Tin 0.007 0.007 Zirconium < 0.003 < 0.001 I Nitrogen 0.003 0.008 Antimony < 0.001 0.002 i Lead 0.001 < 0.001 Vanadium 0.003 0.04 'l H.2 Installation And Withdrawal Schedules BG&E and Duke Powec began working together regarding reactor vessel surveillance programs in 1987. Based on mutual surveillance program needs. Duke Power revised the surveillance capsule withdrawal schedule; the current schedule is shown in table B-3. The current schedule will prevent the remaining McG-1 surveillance capsules from accumulating higher exposure than would be applicable to the McG-1 vessel. Also, the current schedule supports potent al CC-1 life extension needs. We expect i capsule W to accumulate a fluence equivalent to 48 EFPY for CC-1. B.3 Projected Use Of Data i The McG-1 surveillance program results will supplement the CC-1 original surveillance program by providing material property data on the CC-1 axial weld seams 2-203-A,B.C. Results from the McG-1 original surveillance program must meet the following credibility criteria of 10 CFR 50, Appendix H. E185 and RG 1.99 as described in 161 in order to apply them to CC-1 weld 2-203-A,B.C 16]. B-5 J

_,..~. l i Table B-3 Current McG-1 Surveillance Capsule Withdrawal Schedule l Projected l Azimuthal Lead Withdrawal Capsule Fluence .l ID Position Factor Year EOC EFPY (x 10 n/cm') t U 56 5.33* 1 (1984)* 1* 1.06* ' O.46 X 236 5.31' 5 (1988)* 5* 4.33* 1.380* V 58.5 4.76 10 (1993)* 8' 7.27* 2.0 Y 238.5 4.76 14 (1997) 11 10.12 2.86 Z 304 5.31 10 (1993)* 8* 7.27* 2.0 W 124 5.31 Standby f actual values original surveillance capsule reoort documented a capsule fluence of 0.414 x 10" n/cm w j 2 reanalyzed the fluence results and revised the capsule fluence to 0.4 x 10" n/cm We expect this { 2 revision to be officially documented in 1994. 1. Temperature l i The irradiation temperature indicated by the McG-1 surveillance program thermal monitors thsll be within 25*F of the CC-1 reactor vessel wall temperature (based on the I CC-1 cold leg temperature of 550 F). Therefore, the 579"F and the 590*F thermal monitors mus: not melt. g j 2. Upper-Shelf Energy (USE) l The determination of the USE shall be unambiguous. The measured USE shall remain -l f above 50 ft-lbs. i 3. Calculated Versus Measured 30 ft-Ibs Shift i i i The determination of the 30 ft-lbs temperature shall be unambiguous. The measured 30 ft-lb shift shall not exceed the ART,g.r calculated using RG 1.99, Regulatory Position 3 1.1 by more than 2o, (34*F for base metal,56*F for welds). 4. Dosimetry Measurements and Fluence Calculations i The dosimetry measurements should be used to obtain the fluence at the capsule. Results should also be used to ensure McG-111uence model is accurate. i B6

i P t 5. Best Fit Chemistry Factor A minimum of two sets of data shall be used to perform an analysis per RG 1.99, Regulatory Position 2.1. The McG-1 surveillance data measurements shall be within in, of the

RT calculated using RG 1.99, Regulatory Position 2.1 (17'F for base g7 metal,28'F for welds). If the fluence range is equal to or greater than two orders of t

magnitude, the measurements shall be within 20, (34*F for base metal,56*F for welds). [ 5 B.4 Discussion Of Results l i Caneule U t Capsule U was removed in 1984 at the EOC 1 after 1.06 EFPY, and its contents were analyzed [ by }l' [23] and compared to the pre-irradiation results reported in [22]. { The data obtained from capsule U met all the McG-1 original surveillance program credibility criteria: i 1. Temperature 1 The 579"F and the 590*F monitors did not melt. l 2. Upper-Shelf Energy (USE) The determination of the USE was unambiguous. The USE was above 50 ft-lbs, table B-4 I i Table B-4 f i Steasured USE for McG-1 Capsule U Surveillance Materials j Material Measured USE BM-L 133 BM-T 100 l W eld 75 l i i f B-7

i t i i t i 3. Calculated Versus Measured 30 ft-lbs Shift I t The determination of the 30 ft-Ibs temperature was unambiguous. The measured 30 ft- { lb shift for the base metal is within 4*F of the RG 1.99,' Regulatory Position 1.1 calculation, table B-5 and figure B-3. The measured 30 ft-lb shift for the weld metal is 5'F less than the RG 1.99, Regulatory Position 1.1 calculation, table B-5 and figure B-- l

4. These results satisfy the criterion that the measured shifts be no greater than 2a,

] above the RG 1.99, Regulatory Position 1.1 shift calculation. t Table B-5 l l Calculated Versus Measured a RTg.r "I I I McG-1 Capsule U Surveillance Material j Calculated Measured Calculated - Measured Chemistry Fluence 30 ft-lbs shift 30 ft-lbs shift 30 ft-lbs shift l Material Factor Factor (*F) (" F) (* F) i l BM-L 58 0.784 46 45 1 BM-T 58 0.784 46 50 -4 Weld 210 0.784 165 160 5 i 4. Dosimetry Measurements and Fluence Calculations l The dosimetry from the capsule was used to calculate the fluence at the capsule. A l i fluence model for McG-1 was developed and the dosimetry results from capsule U j agreed with the model. 5. Best Fit Chemistry Factor Not applicable since only one data set is available. j i 1 -l B-8 i

M-McG-1 Ilase Metal Chemistry Factor = 58 Capsule U Fluence = 0.46 x 10" n/cm' 240 - Fluence Factor = o.784 Measured a RT = 45 (UM-L) and 50 (IIM-T) It-lbs gg7 c' 10 0 - m c: 12 2e !?U - r1G jf,2,4_2,4,F,,,,,_, 9 C AI.CUL ATION RG 1.99 M 1<.93__21F _,,,, r-,._ ________________ o-I t ,~, tn, t I I i ,7 I 6 bg fluence (x 10E19 n/cm2) Figure 11-3 Comparison lietween ItG 1.99 Calculation aint McG-1 Capsule U Surveillance flesults for Ilase SIctal 11-9 l

300 McG-1 Weld Metal ,/ Chemistry Factor = 210 ,/ Capsule U / Fluence = 0.% x 10" n/cm' ,/ Fluence Factor = 0.784 ,/ NO - Measured ART = IM ft-lbs gg7 j ,s' ,s' ,s' / / ,/ ,/ / t;' 18 0 - ag/ / v $ -b

g g 9,V

,/ gG# c:N Ti / $sO ,/ a c )b / T s v 12 0 - op / gS / 9 G / g 5f,o / Y 39,k)l 9 60 - ,/ / 0-i i i i i i i-i i i i i i .1 1 6 i bg fluence [x.10E19 n/cm2) l agure 11-4 Coniparison lietween RG I.99 Calculation and NicG-1 Capsule U Surveillance Itesults for Weld Sietal m , ' Il-10 _... _.

1 i l i Carnule X j l Capsule X was removed in 1988 at EOC 5 after 4.33 EFPY, and its contents were analyzed by ? IV [24] and compared to the pre-irradiation results reported in [22l. i -i The data obtained from capsule X met all of the McG-1 original surveillance program credibility criteria: l 1. Temperature - j 1 The 579"F and the 590*F monitors did not melt. 2. Upper-Shelf Energy (USE) The determination of the USE was unambiguous. The USE remained above 50 ti-lbs, table B-6. I Table B-6 l l 31easured USE for SicG-1 Capsule X Surveillance 31aterials i ] 31aterial Nieasured USE l BM-L 133 i BM-T 102 l Weld 83 i ? l 3. Calculated Versus Steasured 30 ft-lbs Shift l The determination of the 30 ft-lbs temperature was unambiguous. The measured 30 A-j i Ib shift for the base metal in the transverse direction is 2*F greater than the RG 1.99, i Regulatory Position 1.1 calculation, table B-7 and 6gure B-5. The m.asured 30 ft-lb shitt for the base metal in the longitudinal direction is 18*F greater than the RG 1.99. Regulatory Position 1.1 calculation, table B-7 and figure B-5. The measured 30 ft-lb shift for the weld metal is 64*F less than the RG 1.99, Regulatory Position 1.1 calculation, table B-7 and 6gure B-6. These results satisfy the criterion that the measured shifts be no greater than 2o, above the RG 1.99, Regulatory Position 1.1 } shift calculanon. a s l B-11 ] 1

i k 1 i t Table 11-7

f Calculated Versus Steasured a RT for g.r McG-1 Capsule X Surveillance Material l

k Calculated Measured Calculated - Measured Chemistry Fluence 30 ft-lbs shift 30 ft-lbs shift 30 f:-lhs shift. Material Factor Factor (* F) (*I9 (* F) BM-L 58 1.090 63 45 18 BM-T 58 1.090 63 65 -2 i Weld 210 1.090 229 165 64 i t 4. Dosimetry Measurements and fluence Calculations The dosimetry from the capsule was used to calculate the fluence at the capsule. A l fluence model for McG-1 was developed, and the dosimetry results from capsule U agreed with the model. t' J 1 i i e I t t i 1 i 2 4 P B-12

la t e i 6 N e s a i t i r o I f s lt u s e R e cn la i l u ev r ~ u S X o __ )2 e l m u s c / p a n ,I 5 C s 9 - b I 1 1 1 1 E 1 3 f 0 e G 1 1 r c ) m T u i B x g N M F d t ( i I e I c n ( n a 5 6 e n u l i o m f t a g a N b lu ) L O T 4 lc F I I M a 'm I( I e, A 3 C t L 9 c / 5 g U n0 4 9 9 9 8

0. =

t_ C a. 5 '0 L l l = 1 i y G 9t A t a R x= g g C 9 r g teo 8 rT J Mt 3 n 3 oR 9 c J, e a g 9 t eF c a e 1 g w s a n c a yX = F ! t l r t I t eeer G s l c c u l i 1 ims nns u R Ge ee a n p ch al i e o uu MCCFrM s i ra p m o C 0 0 0 0 0 0 0 4 2 2 6 1 3 2 cm 5Cl on8 i 9t

300 / McG-1 Weld Metal / Chemistry Factor = 210 Capsule X ,/ Fluence = 1.38 x 10" n/cm' ,/ 240 _ Fiuence Factor = i.090 / Measured a RT = 165 ll-lbs,,/ ,/ - Nor s / / ,/ ,/ / / t/ e' r 20 - s9 s.?. 3 v 3 -o -E / / ,/ s &c 'r . o ,/ o %s c / u 12 0 - op-s3 S gG ,/ st ' 9,9) ' g 9 00 - / 0-i i i i i i r, i i i i .1 1 6 bg fluence [x 10E10 n/cm2) Figure 11-6 Comparison lietween RG 1.99 Calculation and AlcG-1 Capsule X Surveillance Results for Weld Stetal 11-1 4

5. Best Fit Chemistry Factor The best fit chemistry factor was calculated using RG 1.99, Regulatory Position 2.1. The measured 30 ft-lb shifts for the base metal at the low fluence value were 3 and 8'F greater than the RG 1.99, Regulatory Position 2.1 calculation. The measured 30 ft-lb shift for the base metal (longitudinal orientation) at the high fluence value was 14*F less than the RG 1.99, Regulatory Position 2.1 calculation. The measured 30 ft-Ib shift for the base metal (transverse orientation) at the high fluence value was 6*F greater than the RG 1.99, Regulatory Position 2.1 calculation, table B-8 and ngure B-7. The measured 30 ft-lb shift for the weld metal at the low fluence value is 27'F greater than the RG 1.99, Regulatory Position 2.1 calculation while the high fluence value is 20*F less, table B-8 and figure B-8. All of the measured shifts satisfy the criterion that the measured shifts be within lo, of the RG 1.99, Regulatory Position 2.1 shift calculation. Table B-8 Calculated Versus Measured ART.g for 3 McG-1 Surveillance Material Using Best Fit Chemistry Factor Calculated Measured Calculated - Me:asured Best Fit Fluence 30 ft-lbs Shift 30 ft-ths Shift 30 ft-lbs Shift. Material CF Factor (*F) (*F) (*F) BM-L 54 0.784 42 45 -3 BM-T 54 0.784 42 50 -8 Bht-L 54 1.090 59 45 14 Bhi-T 54 1.090 59 65 -6 W eld 170 0.784 133 160 -27 Weld 170 1.090 185 165 20 1 B-15

F l l l l 3M - McG-1 Hase Metal Ilest Fit Chemistry Factor

54 Capsules U and X 240 -

t; 10 0 - m y C t2 .s m = 12 0 - 1.99 + 17F 60 - 4-RG J CUI. ATI RG 1.99 C AI. RG 1.99 17F .g. 1 i 'l I g g I i .1 I 6 i log fluence (x 10E19 n/cm2) Figure 11-7 Comparison Iletween ItG 1.99 Calculation and NicG-1 Surveillance Itesults Using Ilest Fit Cliemistry Factor for llase hietal Il-16

500 McG-1 Weld Metal llest Fit Chemistry Factor = 170 ~, ,<'p/~ Capsules O and X 240 - / c 10 0 - O m '/y ,,. / ye: t2 p,f, 2 o 12 0 - u .9.s go '#o,- c o @ 0?', 2,%,$/ g6 00 -.- On i i i i iiri i i i i .1 1 6 bg fluence (x 10E19 n/cm2) Figure 11-8 Comparison Iletween RG 1.99 Calculation ami McG-1 Surveillance Results Using liest Fit Chemistry Factor for Weld Metal 11-17

? l l 1-b i' APPENDIX C ENVIRONAIENTAL EQUIVALENCY BETWEEN CC-1 CLAD / BASE METAL INTERFACE AND AlcG-1 SURVEILLANCE CAPSULE AND CONIPARISON BETWEEN CC-1 AND FAR-1 SURVEILLANCE PROGRA31 RESULTS 4 I l l l l l C-1

~_ P Z APPENDIX C ENVIRON 31 ENTAL EQUIVALENCY BETWEEN CC-1 CLAD / BASE SIETAL INTERFACE AND 31cG-1 SURVEILLANCE CAPSULE AND COMPARISON BETWEEN CC-1 AND FAR-1 SURVEILLANCE PROGRAM RESULTS C.1 Environmental Equivalency Between CC-1 Clad /hase Metal Interface And i McG-1 Surveillance Capsule A data base consisting of surveillance capsule results from C-E and.\\V designed NSSS for plates and. l welds processed in the C-E plant in Chattanooga was analyzed. Design differences affecting irradiation i environments will be evidenced by a bias in the statistical results. There was no bias in the data as a i function of the copper, nickel, fluence or flux, and therefore, no apparent difference in the effect of l irradiation environment [6]. 2 Reference [6] also shows, for fluences greater than 1 x 10" n/cm. most surveillance weld results for C-E fabricated vessels show less RT shift than calculated using RG 1.99. Those that are greater than y g.r the RG 1.99 calculation are within about 10 degrees of the calculation and/or have chemical compositions considerably different from the CC-1 limiting material. Reference [6] also indicates that the most conservative RG 1.99 calculations to date are for the CC-1 and McG-1 surveillance welds. Despite this evidence that different NSSS designs do not cause discernable effects on embrittlement, l BG&E has placed specific additional requirements on the demonstration of similar radiation i environments between the CC-1 and McG-1 reactors. Similarity must be established between fluence l analysis methods, neutron spectra, flux and irradiation temperature before the results of the McG-1 surveillance program can be used to determine the CC-1 values of RT and RT f r CC-1 weld yg.r m seams 2-203-A,B,C. I C.l.1 Fluence Analysis Methods i It is necessary to employ methods which permit accurate comparison of the calculated amount of radiation damage as a function of total power output and environmental conditions. The radiation damage analysis method for the different re. actors should be accurate and yield comparable results for j comparable environments. C-2 1

f C.I.2 Neutron Spectrum i Differences in neutron energy distribution, which are a function of reactor geometry, materials, ana j fuel management, are es aluated in three ways: graphical evaluation of the spectral amplitudes, damage as measured by displacements per atom (dpa) at the McG-1 reactor surveillance capsule and at the CC-1 reactor clad / base metal interface, and the thermal-to-fast flux ratios. The geometry, materials, and fuel management of the CC-1 and McG-1 reactor vessels must be sufficiently similar so that differences in neutron energy spectra between the vessels are no larger than those encountered within a single vessel. l The damage suffered by the McG-1 reactor's surveillance capsule should be equal to or greater than the j damage suffered by the CC-1 reactor's clad / base metal interface. Because some experimental results suggest that thermal-to-fast neutron flux ratios have to exceed 9:1 before embrittlement effects of t engineering significance can be attributed to thermal neutrons, the thermal-to-fast flux ratios at the McG-1 reactor surveillance capsule and the CC-1 reactor's clad / base metal interface must be less than 9:1. l i C.I.3 Neutron Flux l i The neutron fluxes for the McG-1 reactor's surveillance capsule and for the CC-1 reactor's clad / base -l J metal inte. face are to be obtained and compared. Based on extensive published research, it has been. documented that differences within a factor of ten will result in comparable irradiation behavior. Smaller differences are preferable for irradiation periods of ten or more years. To be conservative, any I i flux differences that may exist should yield a greater damage rate for the McG-1 reactor surveillance l capsule. C.I.4 Irradiation Temperature Reactor coolant inlet temperature at full power is to be determined for each cycle and for each vessel. Vessel operating history is to be reviewed to the extent possible to identify any changes from normal i operating temperature for any significant periods of time. Nominal inlet temperatures shall be between 525 and 590*F. The McG-1 surveillance capsule temperature shall be within 25aF of the CC-1 clad / base metal interface. I s [ 8 o C-3

.~ i i t The above criteria and considerations are satisfied. A detailed discussion of these criteria and . l I technical considerations follows: } C.I.1 Fluence Analysis 51ethods .l I It is necessary to employ methods which permit accurate comparison of the calculated amount of radiation damage as a function of total power output and environmental conditions. The radiation t damage analysis method for the different reactors should be accurate and yield comparable results for comparable environments, j The methods for aralyzing the CC-1 and 51cG-1 capsules are documented in [15,18,23,24]. An j additional Electric Power Research Institute (EPRI) calculation of NicG-1 capsule U is contained in 4 1 [25]. The calculational methods employed in these analysis are similar, accurate and yield comparable i resalts. k The hicG-1 capsule U and CC-1 capsules 263* and 97* fluences were calculated with the two-dimensional discreet ordinates transport code (DOT) 4.2 in R-O and R-Z geometries, in which two-dimensional fluxes were synthesized into a single three-dimensional flux distribution. The hicG-1 l capsule X fluences employed a two-dimensional DOT 4.2 R-G calculation and used an axial shape i factor derived from the axial power distribution to determine the flt$ence at any three-dimensional f location. The latter method was benchmarked by W to yield similar results to the former. DOT 4.2 uses a multigroup approximation to the energy variable in the transport equation solution, a P3 Legendre expansion of the flux distribution and scattering terms, an S, approximation to the direction variables and scattering quadrature, a finite difference approximation to the spatial variables, and the ] Evaluated Nuclear Database File (ENDF/B-IV) transport cross section set. The R-O DOT simulations l employ eighth-core symmetry at the core midplane and explicitly model the surveillance capsule structures. i While the CC-1 method uses actual as-built dimensions to model the system, the hieG-1 method credits design dimensions with very tight tolerances, which results in very small uncertainties in the geometry and dimenaions The neutron source distribution in the reactor fuel was derived from three-dimensional reactor-core diffusion theory depletion calculations. Note that while CC-1 used a pin-by-pin level of detail for each fuel assembly source. McG-1 used an assembly average level of detail which was l expanded to pin-by-pin via empirically determined assembly shape matrices. However, the results should be insensitive to these small differences in the modeling of the power distribution. The cycle specific neutron source distributions utilized in these analyses included not only spatial variations of i fissior rates within the reactor core but also the effects of plutonium build-up on neutron yield-per- { i C-4

l i' fission and fission spectrum. This was accomplished during the generation of the pin-by-pin source term in McG-1 and as an assembly-dependent post-calculation correction factor in CC-1. l i Benchmarking of the calculated neutron energy-dependent fluxes is accomplished via comparison with the surveillance capsule dosimetry results. The neutron flux dosimeters in the McG-1 and CC-1 i surveillance capsules both employ bare and cadmium-shielded cobalt / aluminum dosimeters to measure thermal neutron activity and copper, iron, nickel, and depleted uranium dosimeters to measure fast 3 neutron activity. McG-1 uses neptunium and CC-1 uses titanium for additional neutron flux measurements. The use of these passive monitors does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the i integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. The effective dosimeter reaction cross sections were calculated using ENDF/B-V based libraries via the l SAND computer code. The neutron energy spectrum used by SAND was based on the spectrum .f calculated with the DOT 4.2 model of the surveillance capsule. The input data to SAND also included j l the saturated activities calculated from the measured dosimetry activities and from the core power history from dosimeter insertion to extraction. Measurements of individual dosimeters are corrected for { impurity contributions, and McG-1 also corrects for photofission. CC-1 does not correct for I photofission due to its negligible contribution to the total number of fissions [18]. j Similarity of the fluence calculational methodologies permits accurate comparison of the calculated amount of radiation damage as a function of total power output and environmental conditions. ) C.I.2 Neutron Spectrum { Differences in neutron energy distributions are evaluated in three ways: geometry of the reactor vessel, ) damage as ;neasured by dpa at the McG-1 reactor surveillance capsule and at the CC-1 reactor clad / base metal interface, and the thermal--to-fast flux ratios. The geometry of the McG-1 and CC-1 reactor vessels must be sufficiently similar so that differences in neutron energy spectra between the vessels are no larger than those encountered within a single vessel. The damage suffered by the McG-1 reactor's surveillance capsule should be equal to or greater than the damage suffered by the CC-1 reactor's clad / base metal interface. Because some experimental results suggest that thermal-to-fast neutron flux ratios have to exceed 9:1 before embrittlement effects of engineering significance can be attributed to thermal neutrons, the thermal-to-fast flux ratios at the McG-1 reactor surveillance capsule and the CC-1 reactor's clad / base metal interface must be less than 9:1. C-5 i

-~ .- ~ _ -. i i ~l l 1 Many references [6,26-28] in the literature state that PWR energy spectrum differences are of minor importance because the spectra generally encountered are too similar to cause a detectable change in irradiation behavior. This will be shown to be true for CC-1 and McG-1. I Neutron energy distribution comparisons between CC-1 and McG-1 are of three forms. The first is a. j direct graphical comparison of the McG-1 capsule X spectra with CC-! reactor vessel locations; j Additional comparisons between McG-1 capsule X and capsule U spectra show that the two are basically equivalent except for the absolute flux level, so that additional comparisons between the McG-I capsule U spectrum and CC-1 reactor vessel spectra are not necessary. The second is a comparison of the damage parameter dpa between the McG-1 capsules and CC-1 reactor vessel locations. The third is a comparison of thermal-to-fast flux ratios which is included as an indicator of spectral effects. The dpa and thermal-to-fast results will then be biased to simulate a " worst case" scenario, and the biased results re-analyzed. j i C.I.2.1 Graphical Neutron Spectrum Comparisons f Although the neutron energy distribution in and near the cores oflight water moderated reactors i are very similar to the fission spectrum, especially for neutrons in excess of 1 MeV, the neutron distributions near power reactor pressure vessels are not [29]. The relative neutron energy spectrum is a function of the geometry and materials of the reactor internals depending on the 'i proximity to the reactor core and on the media through which the neutrons must pass [6,26]. .As shown in table C-1, the materials and dimensions of the two reactors differ and thus will produce some variation in neutron spectra 118,23-25.30-33]. However, a surveillance program i -i for a single reactor must also contend with the variation in spectrum between the surveillance l capsule and vessel wall and through the reactor vessel wall. The difference in spectra between CC-i and McG-1 are no larger than that encountered within the individual CC-1 and McG-1 l programs. To verify that the differences in neutron energy spectra between CC-1 and McG-1 are no larger i than those encountered within the reactor vessel of a single plant, plots of absolute flux and flux normalized to unity versus energy were generated at various CC-1 and McG-1 locations. Figure C-1 depicts the absolute flux versus energy spectra for the cycle 1-10 average CC-1 surveillance capsule 97*, clad / base metal interface, and 1/4 T location. Figures C-2 through C-1 show the flux normalized to unity versus energy spectra for the cyc'e 1-5 average McG-1 j 1 capsule X and the cycle 1-10 average CC-1 capsule 97", clad / base metal interface and 1/4 T j location, respectively. Figures C-5 and C-6 present the flux normalized to unity versus energy spectra for the cycle 1-5 average McG-1 capsule X and the cycle 11 average CC-1 clad / base l i C-6 i

f i i p metal interface and 1/4 T location, respectively. Figures C-7 through C-Il depict the absolute flux to energy spectra corresponding to the normalized fluxes in figures C-4 through C-8. Comparison of figure C-3 with the normalized and absolute spectra of figures C-2 through C-6 i shows that the spectral differences between the McG-1 and CC-1 surveillance programs and .l within the CC-1 survei!!ance program alone are small. In addition, the cycle 'l-10 average CC-f I absolute flux versus energy data for the clad / base metal interface and 1/4T location presented l t in figures C-8 and C-9 are in general less than that presented for the McG-1 capsule X. Since j the cycle 11 average CC-1 absolute flux versus energy data for the clad / base metal interface and l 1/4T location presented in figures C-10 and C-11 are in general less than that presented for the l cycle 1-10 average CC-1 absolute flux versus energy, the cycle 11 average CC-1 flux versus-j energy data are also in general less than that presented for the McG-1 capsule X. Thus the j application of the McG-1 surveillance capsule results to the CC-1 clad / base metal interface would be conservative. i 4 t in summary, the differences between the CC-l and McG-1 spectral fluxes are of the same j magnitude as differences within the CC-1 environment. In addition. the CC-1 spectral tiuxes in the reactor vessel are less than those derived for the McG-1 capsule X. I -i 4 I } t 1 i r 5 l l t i f C-7 j I

-I i i t Table C-1 l AlcG-1 and CC-1 Dimensions and Properties - l l [ McGuire Unit 1 Calvert ClitTs Unit I j Westinghouse PWR Combustion Engineering PW R i MATERIALS: l Baffle Type 304 stainless steel Type 304 stainless steel Core Barrel Type 304 stainless steel Type 304 stainless steel l j Thermal Shield Type 304 stainless steel na Liner Type 304 stainless steel Type 304 stainless steel l r Pressure Vessel SA 533 Grade B Class 1 SA 533 Grade B Class 1 l ? OPERATING PARAMETERS-Thermal Output 3411 M Wt 2700 MWt .l { Maximum overpower 18 % 12 S Nominal pressure 2280 psia 2250 psia ? 6 Total coolant flow 144.4 x 10'lbm/hr 143.8 x 10 Ihm/hr { Core coolant flow 135.8 x 10'lbmihr 138.5 x10'lbm/hr Inlet Temperature 55 8. l *F 548*F Average Temperature 588. I 'F 572'F Outlet Temperature 618.2*F 596*F Hot Zero Power Temperature 557.0* F 532'F Number assemblies 193 217 Fuel Design 17X17 RCC Canless 14X14 Guardian Enrichment 2-4 % 2-4.5 % C-8

= . - ~. - . } l 1 t Table C-1 (cont'd.) { i McG-1 and CC-1 Dimensions and Properties f I McGuire t' nit I Calvert Clifts Unit 1 Westinghouse PWR Comhustion Engineering PWR l GEOMETRY: j Core equivalent Outer Diameter 132.7 inch 136 inch h Active fuel length 365.76 cm = 144 inch 136.7 inch ) k Total assembly length 414.02 cm = 163 inch 157 inch Pellet Outer Diameter 0.3225 inch 0.3765 inch l 1 Clad Inner Diameter 0.329 inch 0.3840 inch l Clad Outer Diameter 0.374 inch 0.440 inch j i Fuel rod pitch 0.496 inch 0.580 inch t Assembly pitch 8.426 inch 7.980 inch Baffle Inner Radius 161.33 cm 176.81 cm .l Baffle Outer Radius 164.19 cm 179.35 cm Reflector Inner Radius 104.19 cm 179.35 cm Reflector Outer Radius 1S7.96 cm 187.96 cm Core barrel Inner Radius 187.96 cm 187.96 cm i f Core barre! Outer Radius 193.68 cm 192.40 crn Thermal shield Inner Radius 193.68 cm na l j Thermal Shield Outer Radius 200.665 cm na Downcomer Inner Radius 193.68 cm 192.40 cm Downcomer Outer Radius 219.30 cm 220.589 cm l Liner Inner Radius 219.30 cm 220.589 cm Liner Outer Radius 219.71 cm 221.383 cm i 1 i Pressure Vesse! Inner Radius 219,71 cm 221.383 cm j Pressure Vessel Outer Radius 241.62 cm 243.291 cm Capsule inner Radius 205.73 cm 215.237 cm j Capsule Centerline 207.31 cm 117.1585 cm Capsule Outer Radius 208.90 cm 219.0S0 cm j C-9 i - - = -

S i i ~ i 7 ~ > l. 0

~

r" 1 n I o i 9 t a h 0 c 6 o L T 1 4 n / 1 e 5 c [ i 0 a f 1 r i e u. t n I P)7' t 0 r i la 4 te i M u F 1 J ) a n e s r' V B , 'i' E d / 3 r: in 1 l a 0 ( 7 C [ 9 o 1 Y C ~ n EM LB i G 0 r le 1 e U/ R u -C u s T 2 SD E g p o.' PA4 i 0 a i N F C n u 1 AL/ n E 1 u CC1 C i C 1 1 1 1 r CCC i 0 o i f n CCC i 1 y n g g i e r C i E n 0 f_ f in u 0 s n 1 sr n e 7 v x I_ 1 u ] f_ 0 l 1 F L t e 1 u ~ l ~ os ~ , 2 b A 0 a n i 1 i g 7 7 r. 0 9 8 7 0 1 0 0 0 0 1 1 0 0 1 1 1 1 = 1 1 mw[soNZv X3b 4 ) i

1 7 0 1l i in f 1 {, 1' n 6 i 0 79 n n 1 e l u sp u 5 0 C ~ E in f n 1 n C i C g d 4 u 0 a i, i, n 1 X7 n L i ) lu e E9 V s I L E p UL E i 3 I C SU i 0 ( 2 n PS n 1 1 Y C ~ AP n G I CA i G e l c r f C R e a .C i 1 2 g E r 0 F o i 1 in N f G - n 1 y CC MC E g r 1 e t i n 1 i i 0 s n u [I e u i 1 s n r n e i v x [ [J in 0 l 0 f L n 1 d % p i 1 i e o i ~ s I i l a S n r o R to 0 N n 1 o i e 2 '0 in n 1 o

i r,

~.

r 2 1 2 2 3 2 4 2 5 2 6 0 0 0 0 0 0 1 1 1 1 1 1 m$ ^6 xD IL o d ~~J_ h $ Z i

t r t ~ i t e Il iI 7 1 l l 0 i t t h 1 l 1 i e l 1 t ca i f i re 6 tn M t 0 t I i it 1 l s a t t i e F t M e 5 s ~ J t 0 a e n t i I s 1 / i i t e l a 1 I C 1u l 4 1 ]I l 0 C i t 1 X i C t J s ) d E t n ,(' LM i V a UB 3 E X '~j S / f 0 ( i e i 3 l Pb-t P D t 1 A A t Y C u i s CL e G p 2 C i R C 1 e a r u C 1 I 2 E g 1 G - li 0 1 i t N F G a CC t 1 a c e E M ~ MC i r o t f ~ t 0 y 1 t es g t 1 re i n e E i s t u 1", t 0 r 0 s L i v i e i Hs t 1 t t x F u t l r_ e 1 F F d g r e T i 0 i s r t i 1 l n a t m i r i 2 o N l 0 n i t n i 1 - :. i 1 2 2 2 3 2 4 2 5 7 0 0 0 0 0 1 1 1 1 1 fN2OxZ x D>- OJN]43mOz u il

i t I l 7 l 1l !t 0 i i 1 l 1 3 ) i ~ 1 t l i n t 6 C, t 0 t n i i o l 1 i i t ta \\ t o I c l d 5 _ I.I l 0 T i t 1 / n 4 e 1 f i I -C ~ i C 4 f 0 d ru t 1 n L i a r. X t ) X E i V f L i E le U 3 u s,p s I. S 0 ( p tu 4 l PT t 1 u s Y A4 i C C C/ t G 3 1 e 1 1 i R r u G C - 1 2 E g c 1 { 1 0 1 1 N F 1 M i ~- G - 1 1 1 1 y r '- CC 1 E g MC i r e ~ ~ - ~ f 0 s n n E 1 t u i r l 1 s f t v i r e n, t xu 0 F lu l 0 F L a 1 d r e NF t z s i i l a J i 1 m J f ro ] t 0 N vi s i t 1 t i i 2 l 0 a i ti i 1 t ~ 1 2 2 2 3 2 4 2 5 p 0 0 0 0 0 1 1 1 1 1 nWI N3oxZ xDJ_L OJN 43 DOZ _ i_ 2 c 1

10-1 7 ~ F lT I T dCT 2 L y- ] ~ = Q1 10-2 f-q[ f 3 2 l o i_,r j tr N J Z 2 1_ v I] X 10-3 D __1 l U_ l O 2 bJ N 10-4 5 i 4 MCG-1 CAPSULE X h 2 CC-1 CYCLE 11 CLAD /BM h 10-5 \\"'"''"" 2 10-2 10-1 100 1 01 1 02' 103 1 04 1 05 1 06 1 07 ENERGY (EV) Figure C-5 Normalised Flux versus Energy for McG-1 Cupsule X and CC-1 Cil Clad /Ilase Metal interface C-14

10-1 l [ r ~ P ,~ 't\\ 7h t y 2 rl ct f -{ '_l , F, g, ~ A N 10-2 'L 1 g [ N 1 2 tF 1 5 L x 10-3 r D }l J tt C_3 2 lij N 10-4 ] i ~ MCG-1 CAPSULE X 2 CC-1 CYCLE 11 1/4T m 2 h 10-5 t t t ttit t t i r iitti t i t t titti e i i t itit! I i t s tilil I t enfint I t t s t riil e t I ( t rif f I t I t itief I t i i t tiff 1 Ii 7 10-2 10-1 1 00 1 01 1 02 103 1 04 1 05 106 1 07 ENERGY (EV) l'igure C-6 Normalired Flux versus Energy for McG-1 Capsule X and CC-f Cil I/4T location C-15

l' l c t t t 7 kI{ l 0 Ll i 1 i t t t ~L t t 1 i P s it o i t 1 7 t 9 i t e b t s i lu l_ 5 p f l 0 a t C n t 1 i 1 l C ~ ~ e C 1 i 4 d t 0 n i t a H t X t 1 i L i ) le i T[f I E p V u s ~ 3 L X7 l 0 ( 7 I e a E9 i 1 C t t V C 1 1 L E 1 6 UL i G e G I r t R i c C SU t i r 2 E g S L PS i 0 F r i AP i 1 e N o t t CA t E 2 f i y C I g 1 t r 1 en G - 1 CC l 0 E e s ti u I MC l s e 1 l i r e i v j 0 x 1 0 lu i F n L s 1 t e i t t I i lu o i 1 i s f li 0 A l 1 tt t 1 n t i t 2 0 r e i t 1 j : - r!. - r-t ei : j i: 2 9 2 8 2 7 2 6 1 0 0 0 0 1 1 0 0 1 1 1 1 1 1 n1 N2ONZ X' J_L J.L ( 9

lI 7 0 a 4 ,I. i e i n 1 h a c fr 0 te 0 ~ a n u I J i 1 i l t i t U a ,f M e rl i 0 5 e a s fI i d a n, 1 t I / d a ~ C l m 4 0 1 f a, -C 1 o j C ) l q V u u u 3 E s 0 ( X o X n 8 o L M Y C le 1 E n u U B G e p 1 s 7 r J S / R aC C u 2 E a P D 0 ig 1 N F a A A n T CL u E cG 1 C i M 1 go 1 r o G - f 1 0 CC y an 1 g MC n r e n ? r I 0 su _J a, 0 sr 1 e 1 n i v ri ux f 1 lF 0 te an i u i 1 lo i s i bA , 2 a, 0 n, 1 r r - ~

~

7-, 2 0 2 9 2 8 2 7 2 6 0 0 0 0_ 10 1 1 1 1 1 nygyO\\zv XgJ_LL

-L ~- T 2 k L ,_j-] f trq 1 010 i f nS 2 ~ J[ L Fg]g J b{l' Ur r i 1 09 f~ d m FJp 2 lf O 2 h 1 08 Z X MCG-1 CAPSULE X D 2 CC-1 1/4 T J Le 1 07 2 1 06 m i l t t til it t I tiltl i t i t ilfil iI i f till! I i i t illtI i1 1 I Ill!I t i i f fil!I I t I 1111tl t l i lif til l I i l i f fiI I i1 7 10-2 10-1 1 00 1 01 1 02 1 03 1 04 1 05 1 06 107 ENERGY (EV) Figure C-9 Absolute Flux versus Eucrgy for McG-1 Cupsule X und CC-11/4T location C-18 .. ~. -.... ~..

e-i i i 7 i 0 i i i 1 e i i i f i ca LT i t i r 1 e 6 n L 0 l I 5 i 1 a i i t r i M e i c i e 5 s L t> 0 a l I E d i / i 1 a l_ i l i C i I L 4 I g l / i ) 0 C d 'r M i 1 l - B i C l i V C i D i E d uI A 3 n XL i 0 ( 0 a j d C ii 1 1 l X E i Y L1 i C U1 G le o e u i S E i R r s u p c ] P L 0 2 E g a C d AC 1 N F i f CY E 1 C G ^ 1 c 1 M G - 1 0 r f i r CC 1 f d o MC i y g e 5 i n g1 s 0 E 0 d s u 1 t re v 1 xu O l " 0 F [ 1 te u l o s ' 2 b A d 0 1 r : : r.: : - 7... ' ' 2 0 2 9 2 a 2 7 2 6 2 0 o 0 0 1 0 1 t 1 1 1 ^$msO\\zv bJ _u l l I llll

[_r_r, 1 010 = J LonJ J rt ^ - 1{t W 1 09 ^F t' [ F = l ~ y-j'lr 1 r, - _r-i f Jo 3 fJ ,d 1 08 N, 7 v X MCG-1 CAPSULE X 3 1 07 i CC-1 CYCLE 11'1/4 T b_ 1 06 l 1 05 10-2 10-1 100 1 01 1 02 1 03 1 04 105 106 1 07 ENERGY (EV) Figiere C-Il Absolute Flux versus Eucrgy for McG-1 Capsule X ami CC-1 Cil I/4T Isication C-20

C.1.2.2 Displacement Per Atom Comparisons The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neuuon exposure has traditionally been accepted for the implementation of trend curve data to assess vessel condition. However, an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and vessel wall j positions could lead to an improvement in the uncertainties associated with vessel condition [34). ASTM E 853-87 recommends reporting dpa as an energy dependent damage function [35]. The irradiation damage parameter dpa uses the full neutron flux energy spectrum and is defined as the number of times an atom is displaced from its lattice. The energy dependent dpa function used for this evaluation is specified in ASTM E 693-79 [36). Table C-2 tabulates the McG-1 and CC-1 dpa characteristics for a simple and weighted average scheme as defined in ASTM E 693-79. Note that the simple and weighted dpa rates are very similar, which indicates that the energy group structure is sufficiently fine. The cycle 1-10 average CC-1 dpa rates for the clad / base metal and 1/4T location are 45% and 26% of the McG-1 capsule X dpa rate, while the cycle 11 average CC-1 dpa rates for the clad / base metal f interface and 1/4T location are 18% and 12% of the McG-1 capsule X dpa rates, respectively. 3 1 Thus the CC-1 dpa rates are very conservative compared to the most conservative McG-1 dpa l l rate, in addition, the McG-1 capsule X dpa value is 88% of that at the CC-1 clad / base metal interface and 151% of that at the 1/4T location even though McG-1 has operated for half as-l many cycles. Thus the McG-1 surveillance capsule property changes should conservatively l represent the reactor vessel weld property changes at CC-1. j C. I.2.3 Thermal-To-Fast Neutron Flux Ratio Comparisons l Another indication of spectral effects is the thermal-to-fast neutron flux ratios. Variations in transition temperatures are in direct proportion to the fast neutron fluence and do not appear to l be caused by differences in thermal-to-fast neutron flux ratios [37]. Some experimental results, - however, suggest that thermal-to-fast neutron flux ratios will have to exceed 9:1 before i embrittlement effects of engineering significance can be attributed to thermal neutrons [38). l Table C-3 lists the McG-1 and CC-1 exposures, fluxes, fluences, and thermal-to-fast neutron ? flux ratios for various capsules and at various locations. Note that all of the thermal-to-fast I neutron flux ratios are significantly less than 9 and fbr the CC-1 reactor vessel locations are less than unity. [ l 6 I t C-21

r l Table C-2 51cG-1 and CC-1 DPA Characteristics Simple Average Weighted Aserage Simple Average Weighted Aserage 3 4 4 Cycle (x 10~ dpa/sec) (x 10 " dpa/sec) (x 10* dpa) (x 10 dpa) SicG-1 Capsule U 1 2.56 2.57 8.59 8.64 j l,- McG-1 Capsule U (EPRI) 1 2.39 2.53 8.03 8.51 AlcG-1 Capsule X l-5 1.93 1.94 26.4 26.5 CC-1 Capsule 97* 1-10 1.16 1.21 40.5 42.4 CC-1 Clad / Base Metal Interface 1-10 0.873 0.866 30.5 30.' ( i CC-1 1/4T 1-10 0.532 0.504 18.6 '17.6 l CC 1 Clad / Base Metal Interface 11 0.354 0.350 na na i CC-11/4T 11 0.237 0.225 na na i Table C-3 McG-1 and CC-1 Flux and Spectral Characteristics t Exposure Flux Fluence Thennal/ Fast CycJe (EFPY) (x 10 n/cm*-sec) (x 10 n/cm ) Ratio 2 McG-1 Capsule U 1 1.0654 12.3 0.414 0.49 3 r McG-1 Capsule U I (EPRI) 1 1.0654 12.3 0.414 3.78 a McG-1 Capsule X l-5 4.3316 10.I 1.38 0.40 P CC-1 Capsule 97* l-10 11.0700 7.56 2.64 2.92 CC-1 Clad / Base Metal Interface 1-10 11.0700 5.63 1.97 0.92 .{ CC-1 1/4T l-10 11.0700 3.08 1.08 0.05 CC-1 Clad / Base Metal Interface 11 11.0700 2.27 1.97 0.86 11.0700 1.37 1.08 0.05 CC-1 1/4T I1 l l 1 C-22

I 4 J The data used in this analysis are directly from surveillance capsule reports and includes no uncertainties for model differences or measurement errors. Since the surveillance capsule ~ i results display the same characteristics, one can assume that measurement or calculational procedure differences are minimal. For comparison, a very conservad uncertainty of 15% (1 sigma) [39-41] may be applied to both the McG-1 and CC-1 dpa rate spectral results, but in opposite directions, and to the thermal-to-fast neutron ratios, tables C-4 and C-5. Decreasing the McG-1 dpa rates by 15% and increasing the CC-1 dpa rates by 15%, the McG-1 capsules U and X weighted average dpa rates become 2.18 x 10"" and 1.65 x 10"" dpa/see, while the cycle 1-10 average CC-1 clad / base metal interface and 1/4T dpa rates become 1.00 x i I 10" and 0.580 x 10"" dpa/sec, respectively. The cycle 11 average CC-1 clad / base metal interface and 1/4T dpa rates become 0.403 x 10* and 0.259 x 10'"' dpa/sec, respectively, table C-4. The McG-1 capsules still suffer a greater dpa rate than the CC-1 reactor vessel. The results indicate that all thermal-to-fast ratios are still considerably less than the threshold of l l 9:1 [38]. The McG-1 capsule U and X thermal-to-fast reios as calculated by w [23,24] become { 0.56 and 0.46, while the cycle 1-10 average CC-1 capsule clad / base metal interface, and 1/4T i j ratios become 3 36,1.06, and 0.06, respectively, Role C-5. The cycle 11 average CC-1 f clad / base metal interface and 1/4 T ratios become 0.99 and 0.06. respectively, table C-5. t The CC-1 material changes due to spectral effects as measured by graphical, dpa, and thermal-l to-fast flux ratio methods are negligible or less severe compared to McG-1 results. j 4 i i 4 f f i f i ? a f I i l l I i b i i C-23 l . _ ~

1 l I i Table C-4 l SicG-1 and CC-1 DPA Characteristics With a 157c Bias Applied to Each Unil { Simple Average Weighted Aserage Simple Aserage Weighted Average 6 Cycle (x 10" dpa/sec) h 10-'" dpa/sec) (x 10 dpa) . (x 10* dpa) 4 l SicG-1 Capsule U 1 2.18 2.18 7.30 7.34 j -15 % -15% -15 % -15 % i McG-1 Capsule U 1 2.03 2.15 6.83 7.23 (EPHI) -15 % -155 -15% - 15 % AfcG-1 Capsule X l-5 1.64 1.65 22.4 22.5 -15 % -15 5 -15 % -15 % CC-1 Capsule 97* l-10 1.33 1.39 46.6 48.S + 15 % + 15 % + 15 % + 15 % CC.I Clad / Base 1-10 1.00 1.00 35.1 34.7 Interface + 15 9 + 15 % + 15 % + 15 % f CC-1 1/4T l-10 0.612 0.580 21.4 20.2 + 15 5 + 15 % + 15 % + 15 % l l CC-1 Clad /Hase 11 0.407 0.403 na na Sletal Interface + 15 % + 15 % [ CC-1 1/4T 11 0.273 0.259 na na i, + 15 % + 15 % i ) ~ I I l t i i ? i 1 ? 1 C-24 l 8

7 Table C-5 31cG-1 and CC-1 Flus and Spectral Characteristics i With a 15'7c 111as Applied to Each Unit l l Espmure Flus Fluence Thermal / Fast l 2 2 Cycle (EFPY) (x 10 nicm -sec) (x 10 n/cm ) Ratio i McG-1 Capsule U 1 1.0654 10.5 0.352 0.56 i -15% -15 % + 15 % a 3fcG-1 Capsule U 1 1.0654 10.5 0.352 4.35 (EPRI) -15 % -15 % + 15 % i McG-1 Capsule X 1-5 4.3316 S.59 1.17 0.46 l -15 % -15 % + 15 % 4 CC-1 Capsule 97* 1-10 11.0700 8.69 3.04 3.36 l + 15 % + 15 % +15% CC-1 Clad / Base 1-10 11.0700 6.47 2.27 1.06 l Metal Interface + 15 9 + 15 % + 15 % i CC-1 1/4T I-10 11.0700 3.54 1.24 0.06 + 15 % + 15 % + 15 % l t CC-1 Clad / Base 11 11.0700 2.61 2.27 0.99 1 Metal Interface + 15 % + 15 % + 15 % I CC-1 Il4T 11 11.0700 1.58 1.24 0.06 i + 15 % + 15 % + 15 % r I i i .I ~h i l l i C-25 i i

-e

-a

l 1 J l 1 C.I.3 Neutron Flux i t l Neutron flux for each surveillance capsule is to be obtained and compared. Based on extensive j published research, it has been documented that differences within a factor of ten between the fast neutron flux of the McG-1 capsule and CC-1 clad / base metal interface will result in comparable j irradiation behavior., Smaller differences are preferable for irradiation periods of ten or more years. l In order to interpret the neutron irradiation-induced material property changes observed in the surveillance capsules, the neutron environment (flux and neutron energy spectrum) to which the I surveillance capsules were exposed must be deduced from measurements and analysis. To relate the surveillance capsule material propeny changes to the condition of the reactor vessel, an analysis must be performed to relate the neutron environment at various positions within the reactor vessel to that j experienced by the surveillance capsules. Thus the neutron fluxes at the surveillance capsule and within f the reactor vessel wall are critical to determining accurate irradiation-induced material property changes. 1 Neutron flux effects are probable; however, their effects are small, if not negligible, for the range of ) fast fluxes typical of PWR surveillance capsules. Neutron flux differences within a factor of ten will j result in comparable irradiation behavior [6,26 '.:v,37,41-45]. i l Table C-3 shows that the McG-1 fluxes at the capsules exceed those at the CC-1 capsules and in the r CC-1 reactor vessel.. The fluence of both units exceeds 1 x 10" n/cm2 Thus if a flux effect on transition temperature shift does exist, it should be more severe for the McG-1 capsule than for the CC-j 1 clad / wall interface. j i Radiation embrittlement, as measured in dpa, is directly proportional to the fluence with the - i proponionality constant being a function of flux. This phenomena can be observed in Ogure C-12 for the CC-1 capsule, CC-1 clad / base metal interface and McG-1 capsule data (figure C-12 also shows Far-1 I which is discussed later). Note the McG-1 capsule dpa is greater than the CC-1 capsule dpa which is l greater than the CC-1 base / clad metal interface dpa. The McG-1 capsule flux is greater than the CC capsule flux which is greater than the CC-1 wall flux. The data used in this analysis is directly from surveillance capstJe reports and includes no uncertainties for model differences or measurement errors. Since the surveillance capsule results display the same - characteristics, one can assume that measurement or calculation procedure differences are minimal. For comparison, a very conservative uncertainty of 15% (1 sigma) 139-41] may be applied to both the McG-I and CC-1 flux results but in opposite directions. C-26

0.07 0.06 - MCG-1 CAPSULES ~ o FAR-1 CAPSULES -a-CC--1 CAPSULE 97 0.05 .O-CC-1 CLAD /BM y, /,. / _,s/ '/ 0.04 _o. - N,W' <( Q-0.03 / 4/x-o e 0.02 /j-/y- / / 0.01 y- / 0.00 i - 0.01 O.0 0.5 1.0 1.5 2.0 2.5 3.0 FLUENCE (XE19 N/CM2) Figure C-12 DPA versus Fluence for CC-I Capsule 97* and Clad /Ilase Metal Interface, McG-1 Capsules and Far-i Capsules C-27

~ r i Applying a -15% uncenainty to the McG-1 capsules U and X fluxes yields 10.5 x 10'" and 8.59 x 10'" 2 n/cm -sec, while applying a +15% uncertainty to the cycle 1-10 average CC-1 capsule, clad / base metal i interface, and 1/4T fluxes yields 8.69 x 10'",6.47 x 10', and 3.54 x 10' n/cm -sec, respectively. l 2 Applying a +15% uncertainty to the cycle 11 average CC-1 clad / base metal interface and 1/4T fluxes yields 2.61 x 10 and 1.58 x 10 n/cm -sec, respectively. Note that the decreased McG-1 flux is still 2 greater than or equal to the increased CC-1 flux. Thus if a flux effect on embrittlement does exist, it l

,hould still be more severe for the high flux McG-1 capsules than for the low flux CC-1 reactor vessel locations even with unreasonably large uncertainties applied, table C-5.

i C.I.4 Irradiation Temperature Reactor coolant inlet temperature at full power is to be determined for each cycle and for each vessel. Vessel operating history is to be reviewed to the extent possible to identify any changes from normal f operating temperature for any significant periods of time. Nominal inlet temperatures shall be between l 525 and 590"F. The McG-1 surveillance capsule temperature shall be within 25*F of the CC, clad / base metal interface. 'Ihis section compares the inlet temperature operating history for CC-1 and McG-1 and reviews power and test reactor irradiations which studied irradiation temperature effects. The surveillance capsule results are assessed for temperature-induced anomalies. C.1.4.1 CC-1 Inlet Temperature Operating IIistory l t BG&E reviewed temperature data over the operating history of CC-1 and determined that L essentially no power operation occurred with inlet temperature below 525'F [2L in addition, l 4 BG&E reviewed CC-1 operating practices and records and determined the following: t From initial startup in 1974 to September 1977 CC-1 operated on a temperature control program where reactor vessel inlet temperature (T,) was increased linearly from 532 to f i 544.5*F as reactor power was increased from 0 to 100% power (2560 MWTh at that time). -l i Operators adjusted reactor coolant system boron concentration periodically as fuel depleted l to compensate for the decrease in core reactivity and thereby maintained T, within about l one degree of the program value. Approximately 800 Effective Full Power Days (EFPD) .l of operation occurred during cycle I and 2 using this inlet-temperature program. About 40 l EFPD (based on 2700 MWTh full power) occurred during this period at powers for which the program T, was less than 540*F. 4 1 I I C-28 i

i Since September 8,1977, when maximum allowed power was raised from 2560 to 2700 MWTh, CC-1 has operated in accordance with a similar temperature program where T, is 548'F at full power. Figure C-13 shows the inlet, average and outlet temperatures for this .{ control program. l CC-1 is a base-load plant and does not typically operate at reduced power or conduct end-l of-cycle coastdowns of inlet temperature and/or power. l Short periods of operation at reduced power occur during plant startup and shutdown. In addition, post-refueling core-physics testing typically results in several days of operation at reduced power. Minor periods of reduced power operation have occurred during testing and due to operational problems. Minor coastdowns did occur at the end of several cycles. l Figure C-14 shows our estimate of the number of EFPD accumulated in various inlet temperature ranges. A conservative EFPD-weighted average inlet temperature is 545*F. Although gamma heating of the CC-1 vessel beltline would also occur, it is conservative to i discount it in this case. Any irradiation temperature effects on the vessel would also be found in surveillance capsule l results. However, the RT shift of the weld material is substantially less than calculated by sor RG 1.99 for both CC-1 surveillance capsules 115,18]. Base metal for both capsules and SRM RT shifts for the first capsule were in close agreement with RG 1.99. There were no SRM yg.r specimens in the second capsule. Surveillance capsule results did not indicate any adverse departure from the expected capsule behavior, -l d i [ l r a C-29

j 0 1 s j 0 F 1 p 7'- to g> h 0 T 9 1 / FE r E g- / 0 ,3 8 / v g /, a T g 0 a ,)' G 7 m ar y _r lo 6% d g 0) o c, 1 r t, P c ( r T l o e r r w e t 4 w 3 on o P 0 o O ,~ h - 5 P C eC 1 r 0 ) u 3 F- / S i r e t C r a ( S g e t y 9 v* 0 S i p e n l r e i 4N u c, e /, i le T t t a 0 r ^ n e I 1 p y l 3 C m-g C e-T- tn 8,jg 0 6 a 2 loo C '..q 0 r r 1 o tcae R O 0 0 0 0 0 0 0 0 0 0 0 0 9 8 7 6 5 4 3 2 1 0 6 5 5 5 5 5 5 5 5 5 5 1

4 i 1 I Ef fective Full Power Days j 3500'I_ _ _ l l l 1 j 3000 I 3 l 2500 l l i ,l 2000 i 4 i l 1500 j 10C0 i i { 500 i j J /] l 0 < 540 540 - 545 > 545 Inlet Temperature (F) i i I Figure C-14 CC-I EFI'D versus Inlet Temperature 1 l i C.31

C. I.4.2 McG-1 Inlet Temperature Operating IIistory McG-1 operates in accordance with the temperature program shown in figure C-15. This l program is the same as that in the original McG-1 FSAR [33]. Figure C-16 compares the inlet temperature control programs for CC-1 and McG-1. The inlet temperature at all McG-1 power levels is within about 10*F of the CC-1 full power inlet temperature. _l The McG-1 automatic control rod drive system adjusts control rod positions to maintain the f inlet temperature within 2*F of the program value. McG-1 safety analyses assume inlet j temperature is maintained within this range [46]. These are consistent with descriptions in the l original McG-1 FSAR. i i i Since the inlet temperature program for McG-1 is essentially constant at all power levels, a i review of the McGuire power history is not necessary. Any operation at reduced temperature i (eg. end-of-cycle coastdown) would result in conservative surveillance capsule results for the purposes of this program. j l Because of the flat inlet temperature control program, control rod drive control system, and the l t safety analysis assumptions BG&E concludes that the preponderance of irradiation of the McG-I surveillance capsules has occurred at about 558"F. C.I 4.3 McG-1 Capsule Irradiation Temperature f The 579*F temperature monitors included in the McG-1 surveillance capsules did not melt or i exhibit irradiation assisted creep indicating that the temperature of the capsule specimens did j ] not approach this value [23,24]. To minimize the. effects of gamma heating of the McG-1 capsules, the capsules were hydraulically compressed and autoclaved 122,26). This maximizes contact between capsule specimens and the capsule wall to maximize heat transfer. B&W concluded that the B&W capsule design maintains capsules within 9"F above the temperature of the entering coolant water. B&W also concluded that thinner W cladding and f smaller capsule cross section would cause the W designed capsules to have even less sensif to gamma heating and greater response to the coolant inlet temperature [26]. j ^ f I We conclude thit the McG-1 capsule specimen irradiation temperature did not exceed 567*F, the full-power c >olant inlet temperature plus 9"F, and that it was probably several degrees less. i C-32 4 i

l. l l l Reactor Coolant Temperature (F) l l 620 ) 610 600 590 -j 580 s ~ 570 s [ 560 3 3 3 3 3 a c s 3 550 Tcold O Tavg O Thot 540 ^ 530 O 10 20 30 40 50 60 70 80 90 10 0 NSSS Power (%) Figiere C-15 McG-1 Teniperature Control l'rogram C-33 ...... ~., _. _ _ _ _. - _.. _. _. - _ -. - _ _. . - -. _ _. _ _ _ _. _, ~ _

Reactor Coolant inlet Temperature (F) 570 1 '?' CC-1 Curren t McG - 1 O CC-J initial 560 s n n ( } ( '} $) V-w- 550 / .) s 540 gg' 530 O 10 20 30 40 50 60 70 80 90 100 NSSS Power (%) Figure C-16 McG-1 versus CC-1 Inlet Tennperature Control Prograin C-34

'l t ? The preponderance of the irradiation of the CC-1 controlling weld is conservatively estimated to be have occurrred within 22*F of the irradiation of the McG-1 surveillance materials. The use of RG 1.99 i to calculate embrittlement is valid for operating temperatures between 525*F and 590*F and when the surveillance capsule is within 25*F of the reactor vessel inlet temperature. Both CC-1 and McG-1 operate between 525 and 590*F. The McG-1 surveillance capsule is within 25*F of the CC-1 reactor vessel inlet temperature. i BG&E has demonstrated the similarity between the CC-1 clad /hase metal interface and the SIcG 1 surveillance capsules in terms of fluence analysis methods, neutron spectra, neutron flux and irradiation temperature. This justifies the application of the McG-1 surveillance data to the CC-1 ~ reactor vessel weld seams 2-203-A,B,C. f I I 4 1 4 i I 1 I a i i -l ) 4 1 0 C-35

= ) i C.2 Comparison Between CC-1 And Far-1 Surveillance Program Results i 1 BG&E also compared the surveillance program results from Far-1 with the surseillance program - a results from CC-1. This provides additional support that differences between NSSS designs do not noticeably affect how a material emhrittles. The results of the Far-1 surveillance program are i discussed below: t In a Request for Additional information concerning the Calvert Cliffs response to the 1991 version of 10 CFR 50.61, the NRC required BG&E to review other plants surveillance program results containing welds manufactured with wire heats used in the construction of Calvert Cliffs reactor vessel beltline l welds [47]. Two nuclear power plants. McG-1 and Southern Alabama Power Company's Far-1 nuclear reactor, have surveillance programs that contain weld wire heats used in the manufacture of the CC-1 j beltline welds. l Far-1 is a L' NSSS plant with a reactor vessel fabricated by C-E. The weld material monitored by the j Far-1 surveillance program was fabricated using the identical weld wire heat, flux type and tlux lot as l the weld monitored in the CC-1 surveillance program. These welds were fabricated using weld wire heat 33A277 with flux type 0091 (tlux lot 3922) [48-50). Results from Far-1 will be used to support i j the application of the' McG-1 surveillance program data to calculate the embrittlement of the CC-1 axial weld seams 2-203-A,B C by providing evidence that W and C-E NSSS plants are sufficiently similar to produce equivalent embrittlement in materials. 1 Far-1 is a W NSSS plant, similar in design to McG-1, tables C-1 and C-6 [48-51J. Because the results from the Far-1 program compare well with the results from the CC-1 program, this provides strong j y evidence that vessels of different NSSS vendors experience similar 5 i i i l I I i .i i r^ C-36 -)

Table C-6 Far-1 and CC-1 Dimensions and Properties Farlev Unit I Cahert ClitTs Unit I l Westinghouse PWR Combustion Engineering PWR j t MATERIALS: Baffle Type 304 stainless steel Type 304 stainless steel. Core Barre! Type 304 stainless steel Type 304 stainless steel Thermal Shield Type 304 stainless steel na Liner Type 30SL weld overlay Type 304 stamless steel l Pressure Vessel SA 533 Grade B Class 1 SA 533 Grade B Class 1 OPERATING PARAMETERS: .i Thermal Output 2652 MWt 2700 MW l i Maximum overpower 18 % 12 % l Nominal pressure 2250 psia 2250 psia l t i Total coolant flow 99.3 x 10'lbmihr 143.8 x 10' Ibm /hr l Core coolant flow 92.2 x 10* lbm7hr 138.5 x 10' lbm/hr f Inlet Temperature 543.5

F 548*F Average Temperature 577.2
F 572'F Outlet Temperature 606. 8'F 596*F l

5 llot Zero Power Temperature 547.0*F 532*F i Number assemblies 157 217 9 i Fuel Design 17X17 Vantage 5 14X14 Guardian s Enrichment 2-5 % 2-1.5 % 4 j l t i 'f C-37

l i i Talsle C-6 (cont'd.) ( Far-1 and CC-1 Dimensions and Properties i Farley Unit I Calvert Clifts Unit 1 j Westinghouse PWR Combustion Engineering PWR i 1 GEO11ETRY: } Core equiv Outer Diameter 120 0 inch 136 inch Active fuellength 365.76 cm = 144 inch 136.7 inch l t Total assembly length 396.24 cm = 156 mch 157 inch i Pellet Outer Diameter 0.3088 inch 0.3765 inch Clad Inner Diameter 0.315 inch 0.3840 inch Clad Outer Diameter 0.360 inch 0.440 inch I Fuel rod pitch 0.496 inch 0.580 inch j Assembly pitch S.500 inch 7.980 inch Baffle Inner Radius 161.4 cm 176.81 cm i i Baffle Outer Radius 164.2 cm 179.35 cm l Reflector Inner Radius 164.2 cm 179.35 cm i Reflector Outer Radius 182.5 cm 187.96 cm Core barrel Inner Radius 182.5 cm 187.96 cm Core barrel Outer Radius 187.6 cm 192.40 cm Thermal shield Inner Radius 187.6 cm na Thermal Shield Outer Radius 192.7 cm na Downcomer Inner Radius 192.7 cm 192.40 cm Downcomer Outer Radius 5/32-inch min weld overlay 0.589 cm Liner Inner Radius 5/32-inch mm weld overlay 220.589 cm Liner Outer Radius 199.39 cm 221.383 cm Pressure Vessel Inner Radius 199.39 cm 221.383 cm Pressure Vessel Outer Radius 219.393 cm 243.291 cm Capsule Inner Radius 184.619 cm 215.237 cm Capsule Centerline 186.207 cm 217.1585 cm Capsule Outer Radius 187.795 cm 219.080 cm C-38

embrittlement, supporting the application of McG-1 surveillance program results to CC-1. The remainder of this sec. tion demonstrates the environmental equivalency between CC-1, McG-1 and Far-1. C.2.1 Fluence Analysis Methods Since both the Far-1 and McG-1 surveillance programs are analyzed by.W. there is no difference in the I computational method [48-50]. It was established earlier that the McG-1 computational method is equivalent to the CC-1 method; therefore, the Far-1 method must also be equivalent to CC-1. C.2.2 Neutron Spectrum ( Neutron energy distribution comparisons between CC-1 and Far-1 are examined in three ways: the Orst j is a direct graphical comparison of the Far-1 capsule spectra with the CC-1 capsule 97* and McG-1 capsule U. The second is a comparison of the damage parameter dpa between the Far-1 capsules and CC-1 capsule 97*. Finally, a comparison of thermal-to-fast Oux ratios is included as an indicator of 7 spectral effects. i l To verify that no large spectral differences exist between the CC-and Far-1 surveillance capsules, plots of absolute flux' and flux normalized to unity versus energy were generated for the CC-1 capsule 1 j 97*, the McG-1 capsule U, and the Far-1 capsules Y. U, and X [18,23.48-50]. Figures C-17 and C-18 depict the Oux normalized to unity and absolute Cux for the CC-197* capsule versus the Far-1 capsules Y. U, and X. Figures C-19 and C-20 depict the Oux normalized to unity and absolute flux for the ) McG-1 capsule U and the Far-1 capsules Y. U, and X, Note that all of the spectral differences are [ i minor and that the spread is much less than that found within the environment of a single plant. i The irradiation damage parameter dpa uses the full neutron flux energy spectrum and provides some i indication of spectral effects. Table C-7 tabulates the McG-1 CC-1 and Far-1 dpa characteristics for a l simple and weighted average scheme as defined in ASTM E 693-79 [18,23-24,36,48-50). The simple ~ i and weighted dpa rates are very similar, which indicates that the energy group structure is sufficiently l ) fine. The dpa rate for Far-1 capsule X of 3.14 x 10* dpa/see is 2.6 times the CC-1 capsule rate of { l.21 x 10* dpa/sec. Thus the CC-1 dpa rates are much less than the minimum Far-1 dpa rate. The i integrated dpa for Far-i capsule X of 6.07 x 10'2 dpa exceeds that for CC-1 of 4.24 x 10'2 dpa. Thus the Far-1 surveillance capsule property changes should bound those of CC-1 if dpa is a useful measurement of embrittlement. j i l i s C-39

Tahle C-7 McG-1, CC-1 and Far-1 DPA Characteristics Simple Aserage Weighted Aserage Simple Average Weighted Aterage Cycle (x 10" dpaisec) (x 10" dpa/sec) (x 10-' dpa) (x 10 dpa) McG-1 Capsule U 1 2.56 2.57 S.59 8.64 McG-1 Capsule U (EPRI) 1 2.39 2.53 S.03 8.51 McG-1 Capsule X l-5 1.93 1.94 26.4 26.5 CC-1 Capsule 97' l-10 1.16 1.21 40.5 42.4 CC-1 Clad /Itase i Metal Interface 1-10 0.873 0.S66 30.5 30.2 l i CC-1 1/4T l-10 0.532 0.504 I8.6 17.6 i CC-1 Clad / Base Metal Interface 11 0.354 0.350 na na CC-1 1/4T 11 0.237 0.225 na na i Far-1 Capsule Y I 3.30 3.24 11.7 11.5 l I Far-1 Capsule U 1-4 3.90 3.90 37.1 37.2 Far-1 Capsule X l-7 3.11 3.14 60.1 60.7 ] i i I I i 1 a l I i l l C-40 I

t;r i ~ 7 'lg .q 0 l 1 (: 1 - 1' %[ s L o s l e . W-p lu s .P: i 0 a 5 C =. i I - r 1 a F d i n 4 a ~[ i 0 7 1 9 7 YUX e 9 F EEE n ) lu t V ELLL i E sp 3 LUUU a I 0 ( 7 C U SSS 1 S PPP Y 1 l ~ C PAAA G C i. ACCC e C 4 C i R r C. u r 1 1 1 2 E g o i 0 1 f i N F y i RRR 1 CAAA i E g r e CFFF n i E 1 V s 0 u tu s o 1 r e v x b: u l 0 F f 0 =~ t " 1 e n_ ~ m t i i l - -- [J ' 0 N a 1 r o _ j 1 i ' 2 i 0 n n i 1 i^ i: - - g5 - i 2 2 3 2 4 2 5 2 6 2 1 0 0 0 0 0 0 1 1 1 1 1 1 ^O o N z v X D 1_U o J NJ_< h O z 1 L L

I l l 2 1 011 3.p. tra ~~ ~ ~ -l W 2 , w lf b.-.3g F4 L .[d.'- L ~ Q3 m 1 f- 'l Di1f JJ N 9 l- ~h I~ r, 1 ~ fY 1 010 L7 y J N p I l b 1 09 r 4 x D

1

._1 2 ~ g .y 1 08 r w F- ~ t J _1 2 CC-1 CAPSULE 97 O m i-FAR-1 CAPSULE Y w 1 07 FAR-1 CAPSULE U FAR-1 CAPSULE X 2 106 7 -i i r na i i ina ,,,,,,,a 10-2 10-1 1 00 1 01 1 02 1 03 1 04 1 05 j os 107 f ENERGY (EV) Figure C-18 Alisolute Flux tersus Energy for CC-1 Capsule 97" and Far-1 Capsules C-42

I 10-1 r _fD$ 3 - d%g .=:- l ir (f) d l 2 tg N --1 r ~ 10-2 : . 4_: f. y Y O '-~ 'k N Z 3 ~ _f v 2 X 3 10-3 7 _J ] t Q 3 Ld 2 N . -[: ] 10-4 i MCG-1 CAPSULE U "I h ~ FAR-1 CAPSULE U FAR-1 CAPSULE Y O 2 FAR-1 CAPSULE X Z 10-5 I r i s ti!! I t t e rtttt t i t e ritti i t t i titil e i t e t tiil i e i ! 6 !i! - t t s t reit! t i e t teiit i e s t rarr! t i r e i ti!! t a# 10-2 10-1 1 00 1 01 1 02 103 1 04 1 05 1 06 1 07 ENERGY (EV) Figure C-19 Norinalized Flux versus Energy McG-1 Cupsule U and Far-i Cupsules C-43

a fG,]4 1 011 L ~$, g

j 1

t -- A y a'~... F,.. t t. ' = - w 3 3 g _., qdR:4.bN l 2 - E =

L-:

j~p 1-y 1 010 g J R i LH Z 3 1 v 2 1 X D 1 09 _J b_ b.1 3 1-3 2 J5 los y MCG-1 CAPSULE U W FAR-1 CAPSULE Y Q [ FAR-1 CAPSULE U 3 FAR-1 CAPSULE X 1 07

t 1111 1 e e a l t ti!!

I t s t a rtil s e e s ittil t t s t a r til i e a t t eill t t t a ittif t t i ttrif a trient t

t t t tiil I it j-10-2 10-1 1 00 1 01 1 02 103 1 04 1 05 106 1 07 l

ENERGY (EV) Figure C-20 l Alisolute Flux versus Energy for McG-1 Capsule U and Far-I Capsules C-44

i t Another indication of spectial effects is the thermal-to-fast neutron flux ratios. Variations in transition j temperatures are in direct proportion to the fast neutron fluence and do not appear to be caused by i 2-differences in thermal-to-fast neutron flux ratios l37]. As discussed earlier, some experimental results suggest that thermal-to-fast neutron flux ratios in the range of 9:1 will have to be exceeded before j embrittlement effects of engineering significance can be attributed to thermal neutrons [38). Table C-8 lists the McG-1, Far-1 and CC-1 thermal-to-fast neutron flux ratios [18,23,24,31,32,48-50]. Note that I all of the thermal-to-fast neutron flux ratios are significantly less than 9. Thus differences due to spectral effects are negligible. ( C.2.3 Neutron Flux i Since the Far-1 fluxes at the capsules exceed those at the CC-1 capsules and since the fluence of both l units exceeds 1 x 10" n/cm, the embrittlement would be more pronounced at the high fluxes of Far-1 2 than the lower fluxes of CC-1. Thus if a flux effect on embrittlement does exist, it should be more j severe for the Far-1 capsule than for the CC-1 capsule. l Radiation embrittlement, as measured in dpa, is directly proportional to the fluence with the i proportionality constant being a function of flux. This phenomena can be observed in figure C-12 for the CC-1 and Far-1 capsules. Note that the Far-1 capsule dpa and flux exceed the CC-1 cap >ule dpa and flux. C.2.4 Irradiation Temperature l l The use of RG 1.99 to calculate embrittlement is valid for operating temperatures between 525"F and 590*F and when the surveillance capsule is.<ithin 25*F of the reactor vessel inlet temperature. Both I CC-1 and Far-1 operate within 525 and 590*F. The Far-1 surveillance capsule is within 25"F of the f CC-1 surveillance capsule, table C-6. l i i k 4 f I f 3 a i (, 3 l C-45

Table C-8 hicG-1, CC-1 and Far-1 Flux and Spectral Characteristics Expmure Flus Fluence Thermal / Fast 2 2 j Cycle (EFPY) (x 10" n/cm -sec) (x 10" n/cm ) Ratio 51cG-1 Capsule U 1 1.0654 12.3 0.414 0.49 31cG-1 Capsule U (EPRI) 1 1.0654 12.3 0.414 3.78 McG-1 Capsule X 15 4.3316 10.I 1.38 0.40 CC-1 Capsule 97* l-10 11.0700 7.56 2.64 2.92 CC-1 Clad / Base Metal Interface 1-10 11.0700 5.63 1.97 0.92 CC-1 1/4T l-10 11.0700 3.08 1.08 0.05 CC-1 Clad / Base 31etal Interface 11 11.0700 2.27 1.97 0.86 CC-1 1/4T I1 11.0700 1.37 1.08 0.05 Far-1 Capsule Y I 1.1300 16.4 0.583 0.50 Far-1 Capsule U 1-4 3.0200 17.3 1.65 0.18 Far-1 Capsule X l'-7 6.1200 14.5 2.80 0.70 C-46

~ _.. - _ -. -. _. 5 t 1 C.2.5 Far-1 Original Surveillance Program } I The surveillance program at Far-1 consists of six surveillance capsules located in the reactor vessel l between neutron shielding pads and the vessel wall. The capsules contain tensile and CVN specimens as well as dosimeters and thermal monitors [48-50]. Table C-9 provides the chemical composition of l the surveillance weld material. l The current withdrawal schedule for the Far-1 surveillance capsules is shown in table C-10. i i Table C-9 1 Far-1 Chemical Composition of Surveillance Weld Material Element Weld Material Carbon 0.13 Manganese 1.06 Phosphorus 0.016 Sulfur 0.009 'i Silicon 0.27 i Nickel 0.19 Chromium 0.063 l Molybdenum 0.50 Copper 0.14 l 1 Cobalt 0.018 i Tin 0.005 l Nitrogen 0.005 j Aluminum 0.009 Vanadium 0.003 + l l Table C-10 l t i Far-1 Surseillance Capsule Withdrawal Schedule j Removal Time Fluence Capsule EFPY Year (x 10" n/cm ) 1 2 Y 1.13* 1979 0.583* U 3.02" 1983 1.65* X 6.12' 1986 2.80* W 12 4.46 V 21 7.80 Z Standby

actual data l

l C-47 I

t t The Far-1 surveillance program data has been examined and the measured 30 ft-lb shifts are no greater f than 2a, (56*F for weld metal) of RG 1.99, Regulatory Position 1.1 calculations, table C-11 and figure. C-21. Table C-11 Far-1 Surveillance Program Results for Weld Wire IIcat 33A277 i Calculated Measured Calculated - Measured Chemistry Fluence 30 ft-ths Shift 30 ft-Ibs Shift 30 ft-lbs Shift Capsule Factor Factor ('F) (*F) (* F) i Y 78 0.849 66 80 -14 i U 78 1.138 89 80 9 X 78 1.274 99 100 -l A best-fit chemistry factor for the Far-1 results was developed using RG 1.99 Revision 2, Regulatory [ Position 2.1; the results are shown in table C-12 and figure C-22. The measured 30 ft-lb shifts for the weld metal were within lo, (28*F for weld metal) of the RG 1.99, Regulatory Position 2.1 shift calculation. 6 Table C-12 Far-1 Surveillance Program Results for Weld Wire IIcat 33A277 Using Best Fit Chemistry Factor Best Fit Calculated Measured Calculated - Measured Chemistry Fluence 30 ft-ths Shift 30 ft-lbs Shift 30 ft-lbs Shift Capsule Factor Factor ( F) (* F) (* F) Y 79 0.849 67 80 -13 U 79 1.138 90 80 10 X 79 1.274 101 100 1 i i i t I C-48

300 Far-1 Weld Metal Chemistry Factor = 78 Capsules Y and U and X Fluence = 0.583 and 1.65 and 2.80 x 10 n/cm2 NO - Fluence Factor = 0.849 and 1.138 and 1.274 Measured a RT = 80 and 80 and 100 It-lbs yg.r e 10 0 - m .g cg a g,19. 53f g u 12 0 - B B 60 - ) g,339._qsE 0= i i i i i i i ri i i i i .1 1 6 log fluence [x 10E19 n/an2] Figure C-21 Comparisim lictween itG 1.99 Calculation and Far-1 Surveillance Ilesults for Weld Wire IIcal 33A277 C-49

300 Far-1 Weld Metal liest Fit Chemistry Factor = 79 Capsules Y and U and X 240 - c 10 0 - v W te d [EO ~ g 5 .w _L.n s_7.gc______ 60 ~ C ALCUL ATION 1.99 RG 89-E8f Eh 0-I I i '~l'r, I I I .1 I 6 bg fluence (x 10E19 n/cm2] Figure C-22 Comparison lietween ItG 1.99 Calculation and Far-l Surveillance Itesults Using liest Fil Cliemistry Factor for Weld Wire IIcal 33A277 C-50

A best-fit chemistry factor using the Far-1 and CC-1 surveillance program results was developed using j RG 1.99, Regulatory Position 2.1; the results are shown in table C-13 and figure C-23.. The measured 30 ft-lb shifts for the weld metal are within lo, (28*F for weld metal) of the best fit curve. ' The similarity _in results between Far-1 and CC-1 is evidence that different NSSS vendors OV and C-E) will i produce similar embrittlement of the same material. j i 4 Table C-13 I CC-1 and Far-1 Surseillance Program Results for Weld Wire IIcat 33A277 Using the Best Fit j Chemistry Factor l Best Fit Calculated Steasured Calculated - 51easured l Chemistry Fluence 30 ft-lbs Shift 30 fi-lbs Shift 30 ft-lbs Shift l Capsule Factor Factor (* F) (*F) (* F) i j Y 77 0.849 65 80 -15 U 77 1.138 88 80 8 X 7.7 1.274 98 100 -2 i 263* 77 0.866 67 59 8 97* 77 1.260 97 93 4 ) I in conclusion, reference [6] demonstrated there is no bias in surveillance program results as a i I function of copper content, nickel content or fluence between C-E and E NSSS designs. BG&E j has shown that differences in fluence analysis methods, neutron spectrum, neutron flux and f irradiation temperature between CC-1 and SicG-1 are negligible. Lastly, the surveillance program i results on ueld wire 33A277 from a E NSSS (Far-1) and a C-E NSSS (CC-1)is a specific example i demonstrating that small environmental differences do not affect embrittlement response. l i l 1 I 1 C-51 i

300 Far-1 and CC-l Weld Metal Best I it Chemistry Factor = 77 Capsules Y and U and X and 263" and 97* 240 - c 10 0 - v

ts c:

bi2 sie 12 0 - g _ _BG _1_.9_9_+_ _20_F_ _ _ _ _ _ _ _ _ 4 4 O OO - C At.CUL AT ION IL_,l.99 C _B9 LD.=3D __ _ 0= I 1 I ' r r~7 I i .1 1 6 log fluence (x 10E10 n/cm2) Figure C-23 Comparison Iletween RG l.99 Calculation and Far-i aitti CC-1 Surveillance Results Using flest Fit Chemistry Factor for Weld Wire IIcat 33A277 C-52

I APPENDIX D CC-1 SUPPLEMENTAL SURVEILLANCE PROGILOI i i 1 9 i l l L f D-1 f

t i i t APPENDIX D l i CC-1 SUPPLEMENTAL SURVEILLANCE PROGRAM i 6 D.1 Contents and Location l The supplemental capsule is installed in the 263* position left vacant by withdrawal of an original t surveillance capsule in 1979. I t The material for the supplemental capsule was obtained from Duke Power Company's McG-1 surveillance weld archive block. The McG-1 weld was fabricated by C-E using identical processes as i the CC-1 controlling material. Additional material for the supplemental program was obtained from a ~ nozzle drop-out which has a weld chemistry similar to the McG-1 weld. The McG-1 and nozzle drop-out weld material chemistry and traceability is documented in 152,53]. The supplemental capsule contains three sets of nine flux monitors and two sets each of twelve CVN impact specimens from the McG-1 surveillance and nozzle drop-out welds. The materials are evenly distributed between the top and bottom, figure D-l. This allows the capsule to be halved, reconstituted and reirradiated so embrittlement data can be obtained at two different fluence levels on both materials. ) i Each CVN impact compartment has 12 impact test specimens in a 4H x 3W x ID array with the notches facing the core. Each flux monitor compartment contains one set of nine flux monitors, table D-1. One j set of nine flux monitors was installed in each tensile specimen compartment. In contrast to the original l surveillance program, niobium was used in place of sulfur to provide a low threshold energy and long halflife. Flux monitor chemistry is provided in references 152,53]. Table D-1 Material for Neutron Flux Monitors i i Threshold Energy Material (MeV) IIalf-Life Niobium 0.03 16.0 years Uranium

0.'

30.2 years Iron 4.0 314 days Nickel" 5.0 71 day's Copper" 7.0 5.3 years t Titanium 8.0 84 days l Cobalt Thermal 5.3 years 1 cadmium shielded and bare " cadmium shielded i D-2 i

e ? 4 Mes -lark Assembly 3 s.d } Wedge Coupling Assemb A l / -N Dux Monitor Compartment l N\\ N _ Nozzle Drop-Out CVN Compartment N / N i -McG 1 CVN Compartment .\\ a N. 4 a f Rux Monitor Compartment i / N. _ Nozzle Drop Out j CVN Companment N \\ s a---McG 1 CVN Compartment N! N I I Rux Monitor Compartment .N N N Fi;:ure D-1 CC-1 Supplemental 263* Surveillance Capsule D-3 i .I

1 i l D.2 Installation and Withdrawal Schedules The supplemental capsule was installed during the 1988 outage (EOC 9). When removed, it will be sectioned in two with one half removed for testing and the second half reconstituted for reinstallation in - l the 263* position, table D-2. [ Il Table D-2 i CC-1 Supplemental Surveillance Capsule Withdrawal Schedule l Capsule Azimuthal Lead Withdrawal Projected Capsule Fluence

i 2

ID Position Factor Year EOC (x 10' n/cm ) t Si 263* 1.30 2000 14 0.57 S2 263* 1.30 2012 20 1.43 Projected capsule fluences based on continuation of the current cycle 11 fuel management. D.3 Projected Use Of Data The supplemental suiveillance program is limited by the low flux and lead factor of the surveillance-capsule location. This prevents the supplemental material from reaching end-of-life fluence before the l current end-of-license data. However, the data obtained t' rom the McG-1 weld will be used to verify the j embrittlement observed at low to intermediate fluence levels from the McG-1 surveillance program. This will support the application of the McG-1 surveillance data to calculate the embrittlement of the CC-1 controlling weld (2-203-A,B,C). The results from the supplemental surveillance program must meet the following criteria in order to support the application of the McG-1 surveillance program data ^ to CC-1: 1 1. Temperature { ASTM E185-82 requires the maximum exposure temperature of the surveillance !i program materials (indicated by temperature monitors) remain within 25"F of the { expected capsule exposure temperature (indicated by operating temperature). The CC-1 [ f operating temperature is nominally 550*F. Therefore, the 536*F monitor must melt while the 580*F and 590*F monitors must not melt. [ 2. Upper-Shelf Energy (USL') I Determination of the USE shall be unambiguous. The measured USE shall remain above 50 ft-lbs. j i i D-4 l

'1 l'! t 3. Calculated Versus hicaspicd 30 ft-lbs Shift The determination of the 30 ft-lb temperature shall be unambiguous. The measured 30 l ft-lb shift shall be no greater than 2o, (56*F for welds) of the RG 1.99, Regulatory. l Position 1.1 shift calculation. In addition, the measured 30 ft-lb shift shall be within j lo, (28*F for welds) of the RG 1.99. Regulatory Position 2.1 shift calculation usmg j the data from this program and the McG-1 surveillance program. f l 4. Dosimetry 31easurements i The dosimetry measurements must be used to obtain the fluence at the capsule. Results 3 should also be used to ensure CC-111uence model is accurate. l r D.4 Discussion o'r Results Results are not expected until EOC 14 (approximately 2000). i I r -i h i t i i L i i D-5

I e 5 l .1, } i 5 l .i, I l [ i i i 1 l i REFERENCES 4 I I l i i i I i Ref-1

i -1 l 1 REFERENCES i 1. Letter from G.C. Creel (BG&E) to NRC Document Control Desk, dated Decernber 13, 1991. j " Response to the 1991 Pressurized Thermal Shock Rule." i 2. Letter from G.C. Creel (BG&E) to NRC Document Control Desk, dated June 30,1992, " Response to Generic Letter 92-01, Reactor Vessel Structural Integrity,10 CFR 50.54(f)." j i 3. SECY 93-048," Status of Reactor Pressure Vessel Issues Including Compliance with 10 CFR Part 50, Appendices G and H (WITS 9100165)," February,1993. 1 4. Letter from R.E. Denton (BG&E) to T.C. McMeekin (Duke Power Company), dated December r 21,1992, "McGuire Unit 1 Reactor Vessel Surveillance Capsule." S. " Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit i Nuclear Power Plant," NUREGICR-4022, ORNL/TM-9408, September 1985. 6. " Application of Reactor Vessel Surveillance Data for Embrittlement Management," CEN-405-P Revision 2, Enclosure 1 to CEOG 93-252, Raymond F. Burski, Chairman CEOG to USNRC. June 9,1993. .l i 7. " Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Calvert Cliffs - Unit 1 Reactor Vessel Materials," Combustion Engineering, CENPD-34, February 4,1972. I 8. " Program for Irradiation Surveillance of Calvert Cliffs Reactor Vessel Materials," B-i NLM-010, Combustion Engineering, July 15. 1970. l l 9. " Testing and Evaluation of Calven Cliffs. Units I and 2 Reactor Vessel Materials, Irradiation Surveillance Program Baseline Samples for the Baltimore Gas & Electric j Company," Combustion Engineering, TR-ESS-001 January 31,1975. l l 10. " Test Plan for the Testing of Calvert Cliffs Units 1 & 2 Reactor Vessel Materials Irradiation Surveillance Program Baseline Samples," Combustion Engineering, Test Plan No. 8067-TP-l NLM-001, August 1973. 11. Letter from R.F. Ash (BG&E) to R. A. Clark (NRC), dated January 20,1982, " Surveillance Capsule 263." 12. Letter from R.A. Clark (NRC) to A.E. Lundvall, Jr. (BG&E), dated February 2,1982, " Surveillance Capsule 263." l 13. Letter from G.C. Creel (BG&E) to NRC Document Control Desk,' dated January 24,1992, l " Modification of the Withdrawal Schedule for Reactor Vessel Material Specimens for l Calvert Cliffs Unit 1." i 14. Letter from D.G Mcdonald (NRC) to G.C. Creel (BG&E), dated March 11, 1992, l " Withdrawal Schedule Change for Reactor Vessel Material Specimens " 1 15. Perrin, J.S., et al., "Calvert Cliffs Unit No.1 Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule 263," December 15, 1980. Ref-2

. ~ _ ~. i l 16. Letter from R.F. Ash (BG&E) to R.A. Clark (NRC), dated February 4,1981, " Reactor l Pressure Vessel Surveillance Program: Capsule 263." i 17. Letter from R.E. Denton (BG&E) to NRC Document Control Desk, dated October 8,1993, " Response to NRC's Request for Additional Information Regarding Baltimore Gas and Electric Company's Response to Generic Letter 92-01 and Capsule Report BAW-2160 Data Clarification." i 18. Lowe, Jr., A L. and G.P. Cavanaugh, " Analysis of Capsule 97* Baltimore Gas & Electric .-t Company Calvert Cliffs Nuclear Power Plant Unit No.1," B&W Nuclear Service Company, BAW-2160, June 1993. 19. Letter from R.E. Denton (BG&E) to NRC Document Control Desk, dated June 22,1993, " Analysis of the Calvert Cliffs Unit No.1 Reactor Vessel Surveillance Capsule Withdrawn from the 97* Location." 20. Letter from R.E. Denton (BG&E) to NRC Document Control Desk, dated March 22,1993, i " Reactor Vessel Capsule Report." j i 21. Letter from D.G. Mcdonald (NRC) to R.E. Denton (BG&E), dated April 13,1993, " Reactor Vessel Surveillance Capsule Report." -l I 22. " Duke Power Company William B. McGuire Unit No.1 Reactor Vessel Radiation Surveillance Program. Westinghouse Electric Corporation. WCAP-9195, November j l 1977. l 23, " Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor ~i Vessel Radiation Surveillance Program, Westinghouse Electric Corporation, WCAP-10786, February 1985. l 24. " Analysis of. Capsule X from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program. Westinghouse Electric Corporation, WCAP-12354, August 1989. 25. "PWR Pressure Vessel Neutron Spectra at McGuire-1," EPRI NP-5622, February 1988. 26. Harbison, L.S., " Master Integrated Reactor Vessel Surveillance Program," BAW-1543, Rev.4, February 1993. l i 27. Status of US A Nuclear Reactor Pressure Vessel Surveillance for Radiation Effects. ASTM l e STP 784, January 1983. 28. Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review IIhird Volumet ASTM STP 1011, February 1989. l t 29. Analvsis of Reactor Vessel Radiation Effects Surveillance Procrams. ASTM STP 481, i December 1970. ll 30. Calvert Cliffs Updated Final Safety Analysis Report. j 31. Letter from G.E.Gryczkowski (BG&E) to G.P.Cavanaugh (C-E), dated August 1993, "Calvert Cliffs Unit 1 Cycle 1-10 Neutron Energy Spectra at the Vessel-Clad Interface and at the Vessel l 1/4 T Location," NEU-93-234. I Ref-3 l

32. Letter from G.E.Gryczkowski_ (BG&E) to G.P..Cavanaugh (C-E), dated August 1993, "Calvert Cliffs Unit 1 Cycle 11 Neutron Energy Spectra at the Vessel-Clad Interface and at the Vessel 1/4 T Location," NEU-93-235. l 33. McGuire Unit 1 Updated Final Safety Analysis Report. .i 34. Lucas, G.E. and G.R. Odette, *Recent Advances in Predicting Embrittlement of Reactor. 'j Pressure Vessel Steels, " Proceedines of the Second International Symposium on Environmental-Deeradation of Materials in Nuclear Power Systems - Water Reactors," ANS, TMS-AIME-NACE, Monterey CA, Sept. 9-12, 1985. l 3S. " Standard Practice for Analysis and Interpretation of LWR Surveillance Results, ASTM E l 853-87. 36. " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom j ,} (dpa)," ASTM E 693-79. I 37. " Compilation of Contract Research for the Materials Engineering Branch, Division of l Engineering," NUREG-0975 Vol.8. March 1991. i 0 38. " Damage Function Analysis of Neutron Energy and Spectrum Effects Upon the Radiation l Embrittlement of Steels," NRL Report 6925, July 25,1969. l 39. " Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance," ASTM E 944-89. .i 4 40. " LWR Pressure Vessel Surveillance Dosimetry improvement Program: PCA Experiments and Blind Test," NUREG/CR-1861, July 1981. 41. " Compilation of Contract Research for the Materials Engineering Branch, Division of f Engineering," NUREG-0973 Vol.7. May 1989. 42. Effects of Radiation on Materials. ASTM STP 725, June 1980. 43. Results of the IAEA Coordinated Research Procrams on irradiation Effects on Advanced Pressure Vessels. ASTM STP 870,1985. I 44. " Experimental Assessments of Gundremmingen Reactor Pressure Vessel. Archive Material for i Fluence Rate' Effects Studies," NUREG/CR-5201, MEA-2286 October 1988. j i 45. " Influence of Fluence Rate on Radiation Induced Mechanical Property Changes in Reactor Pressure Vessel Steels," NUREG/CR-5493, MEA-2376, March 1990. i 46. " Duke Power Company McGuire Nuclear Station Units 1 and 2 Precautions, Limitations and l Setpoints for Nuclear Steam Supply Systems," Westinghouse Electric Corporation, Pittsburgh, April 1987. 47. Letter from G.C. Creel (BG&E) to NRC Document Control Desk, dated May 22,1992, " Response to NRC's Request for Additional Information Regarding Baltimore Gas & Electric i Company's Response to the 1991 Pressurized Thermal Shock Rule, dated March 31, 1992." l 48. Boggs. R.S., et al., " Analysis of Capsule U from the Alabama Power Company Joseph M. l Farley Unit 1 Reactor Vessel Radiation Surveillance Program." WCAP-10474, February 3 1984. -l ? Ref-4

i l l 49. Yanichko, S.E., et al., " Analysis of Capsule Y from the Alabama Power Company Joseph M. l Farley Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-9717, June 1980. 50. Shogan, R.P., et al., " Analysis of Capsule X from the Alabama Power Company Joseph M. l Farley Unit i Reactor Vessel Radiation Surveillance Program," WCAP-Il563, Revision 1, + September 1987 51. Farley Unit 1 Updated Final Safety Analysis Report. i 52. " Reactor Vessel Weld Materials for Calvert Cliffs Unit 1 Supplemental Surveillance l Program," Combustion Engineering, CEN-612-P, November 1993. I 53. " Summary Report on Manufacture of in-Vessel and Ex-Vessel Surveillance Capsule Assemblies for Calvert Cliffs Unit 1 Reactor Vessel," Combustion Engineering, TR-l 2987-MCC-001. January 1989. j ?i h I f i i. t 4 6 i f i ~ i Ref !

4 ATTACHMENT (2) f i L i PROPRIETARY AFFIDAVIT TO NITACHMENT (3) j i r s i i i a Baltimore Gas and Electric Company Docket No. 50-317 i November 29,1993 1

t t y i i AFFIDAVIT PURSUANT j TO 10 CFR 2.790 Combustion Engineering, Inc. ) State of Connecticut ) County of Hartford ) SS.: I, S. A. Toelle, depose and say that I am the Manager, Nuclear l Licensing, of Combustion Engineering, Inc., duly authorized to make this affidasit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conjunction with the application of Baltimore Gas & Electric Company for withholding this information in conformance with the provisions l of 10 CFR 2.790 of the Commission's regulations. The information for which proprietary treatment is sought is. j contained in the following document: CEN-612-P, " Reactor Vessel Weld Materials for Calvert Cliffs Unit 1 Supplemental Surveillance Program," November 1993. l 1 i This document has been appropriately designated as proprietary. I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial { information. l Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in-determining whether the i ? I l o

} i i i information sought to be withheld from public disclosure, included in the above referenced document, should be withhold. 1. The information sought to be withheld from public disclosure, which is owned and has been held in confidence by Combustion i Engineering, is the specific material data for surveillance welds for reactor vessels fabricated by Combustion Engineering, i i ) 2. The information consists of test data or other similar data t concerning a process, method or component, the application of l l which results in substantial competitive advantage to combustion j l Engineering. 3. The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the i public. Combustion Engineering has a rational basis for determining the types of information customarily held in { confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The details of the aforementioned system were l provided to the Nuclear Regulatory Commission via letter DP.537 from F. M. Stern to Frank Schroeder dated December 2, 1974. This system was applied in determining that the subject document I herein is proprietary. l 4. The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the

1 'n I': i understanding that it is to be received in confidence by the Commission. i 5. The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties j has been made pursuant to regulatory provisions or proprietary l agreements which provide for maintenance of the information in _j confidence. l i 6. Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because: i a. A similar product.is manufactured and sold by major pressurized water reactor competitors of Combustion l Engineering. b. Development of this information by C-E required thousands { of manhours and hundreds of thousands of dollars. To the i best of my knowledge and belief, a competitor would have to undergo similar expense in generating equivalent information. i c. In order to acquire such information, a competitor would-also require considerable time and inconvenience to I determine the specific material data for surveillance welds ) for reactor vessels fabricated by Combustion Engineering. d. The information required significant effort and expense to obtain the licensing approvals necessary for application of

i ~. l ! '[ l t the information. Avoidance of this expense would decrease j l a competitor's cost in applying the information and l i marketing the product to which the information is l 1 applicable. e. The information consists of the specific material data for surveillance welds for reactor vessels fabricated by j i Combustion Engineering, tha application of which provides l a competitive econom.' c af. vantage. The availability of such { 3 information to competitors would enable them to modify their product to better compete with Combustion j Engineering, take marketing or other actions to improve l l their product's position or impair the position of Combustion Engineering's product, and avoid developing l \\ similar data and analyses in support of their processes, j methods or apparatus. f. In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, l manufacturing, licensing, quality assurance and other costs and expenses must be included. The ability of Combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to i sell at prices reflecting significantly lower costs. g. Use of the information by competitors in the international marketplace would increase their ability to market nuclear j steam supply systems by reducing the costs associated with their technology development. In addition, disclosure

i i t would have an adverse-economic impact on Combustion l Engineer.ing's potential for obtaining or maintaining foreign licensees. i Further the deponent sayeth not. I S. A. I. S. A. Toelle l Manager l Nuclear Licensing Sworn to before me this MM M day of MA604 'A > 1993 (- !)) LUC n( /.. A -). -j. c m ) N j'. s " Notary Public My commission expires: 3 '3 I -7 f i l 4 y

f ENCLOSURE TO ATTACHMENT (I) i REACTOR VESSEL WELD MATERIALS FOR CALVERT CLIFFS 1 SUPPLEMENTAL SURVEILLANCE PROGRAM NON-PROPRIETARY VERSION i Baltimore Gas and Electric Company Docket No. 50-317 Novemher 29,1993}}

ML20058G175

Text

2.0 BACKGROUND

3.0 INTRODUCTION

SUMMARY

2.0 BACKGROUND

3.0 INTRODUCTION

Navigation menu

Search