ML20101M541

Gap Aw Vogtle Unit 2 Reactor Vessel Radiation Surveillance Program

ML20101M541
Person / Time
Site:	Vogtle
Issue date:	04/30/1992
From:	Meyers T GEORGIA POWER CO.
To:
Shared Package
ML20101H833	List: ML20101H830 ML20101M531 ML20101M541
References
WCAP-11381, NUDOCS 9207080276
Download: ML20101M541 (27)

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- WCAP-11011 WESTINGHOUSE CLASS 3 GEORGIA POWER COMPANY ALVIN W. VOGTLE UNIT NO.1 REACTOR VESDEL RADIATION SURVEILLANCE PROGRAM L. R. Singer February 1986 APPROVED: .

T. A. Meyer, Manager Structural Materials And Reliability Teclinology Work Performed Under GAEJ-100 c -

SNESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsourgh, Pennsylvania 15.^30 L

~ -: , _

PREFACE

+ This report has been technically reviewed and checked by S. E. Yanichko of Structural Materir.ls and Reliability Technology.

, b S. E. Yanichko Date: February 14,1986 ill

ABSTRACT A pressure vesset steel surveillance program per ASTM E-185-82 has been developed for the Georgia Power Company, Alvin W. Vogtle Unit No.1 to obtain information on the effects of radiatlon on reactor pressuro vessel material under operating conditions. The radiation survcillance program to the Alvin W. Vogtle Unit No.1 is designed to, and in compliance with, federal govemment regulations identified in appendix H to 10CFR, part 50 entitled "Reac-tor Vessel Material Surveillance Program Requirements."

Following is a description of the program, a description of the material involved, the specimen and capsule design and fabrication, and the preirradiation test results.

v 4

4 - v s

TABLE OF CONTENTS Section Title Page 1 PURPOS AND SCOPE 11 2 CAPSULE PREPARATION 2-1

21. Pressure Vessel Material 2-1 2-2. Machining 21 2-3. Charpy V-notch Impact Speimens 21 2-4. Tensile Specimens 23 2-5. 1/2T Compact Specimens 2-3 2-6. Dosimeters 2-3 2.7. Thermal Monitors 23 2.8. Capsule Loading 2-9 3 PREIRRADIATION TESTING 3-1

31. Charpy V-notch Tests 3-1 3-2. -Tensile Tests 3-1 3-3. Dropweight Tests 3-2 4- POSTIRRADIATION TESTING 41 4-1. Capsule Remcval 4-1 4-2. 'Charpy V-notch Impact Tests 4-2

'4-3. Tensile Tests 4-2 4-4. . Fracture Toughness Tests on 1/2T Compact Specimer.s .4-2

5. Postirradiation Test Equipment 4-3 Appendix A DESCRIPTION AND CHARACTERIZATION OF THE ALVIN W. VOGTLE UNIT NO.1 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS A-1 vii

LIST OF ILLUSTRATIONS Figure Title Page 1-1 Location of the Irradiation Test Capsules in the Alvin W. Vogtle Unit No.1 Reactor Vessel 1-4 21 Charpy V-notch Impact Specimen 2-2 2-2 Tensile Specimen 2-4 2-3 Compact Specimen 2-5 2-4 Irradiation Capsule Assembly 2-7/2-8 2-5 Dosimeter Block Assembly 2 10 2-6 Specimen Locations in the Alvin W. Vogtle Unit No.1 Reactor Surveillance Test Capsules 2-13/2 14 3-1 P6eirradiation Charpy V-notch Impact Energy for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientatior.) 3-9 3-2 Preirradiation Charoy V-notch Impact Energy for the Alvin W. Vogtle r' No.1 Reactor Pressure Vessel Intermediate Shen Plate B8805-3 (Transverse Orientation) 3-9 33 Preirradiation Charpy V-notch Impact Energy for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-10 3-4 Preirradistion Charpy V-notch Impact Energy for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Weld Heat-Affected-Zone Material 3-10 3-5 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) 3-11 3-6 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) 3-12 3-7 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No.1 Reacter Pressure Vessel Core Region Weld Metal 3 13 3-8 ' Typical Stress-Strain Curve for Tensile Test 3-14 ix

LIST OF TABLES Title Page Table 21 Type and Number of Specimens in the Alvin W. Vogtle Unit No.1 Surveillance Test Capsules 2-9 Quantity of isotopes Contained in the Dosimeter Blocks 2-11 2-2 3-1 Preitradiation Charpy V-notch Impact Data for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Intermediate Shell Plate B8805-3 3-3

( ngitudinal Orientation) 3-2 Preirradiation Charpy V-notch Impact Data for the Alvin W. Vogtle Unit No.1 Reactor Pressu e Vessei Intermediate Shell Plate B8805-3 3-4 (Transverse Orientation) 3-3 Preitradiation Charpy V notch Impact Data for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-5 3-4 Preirradiation Charpy V-notch Impact Data for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Weld Heat-Affected Zone Material 3-C 3-5 Summary of the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel impact Test Results for !ntermediate Shell Plate B8805-3 and Core Regicn Weld 3-7 and lleat Affected-Zone Material 3-6 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Intermediate 3-8 Shell Plate B8805-3 and Core Region Weld Metal Surveillance Capsule Removal Schedule 41 4-1 A-1 Chemical Analysis of the Intermediate Shell Plates used in the Core Region of the Alvin W. Vogtle Unit No.1 A-2 Reactor Pressure Vessel A-2 Chemical Analysis of the Lower Shell Plates used in the Core Region of the Alvin W. Vogtle Unit No.1 A-3 Reactor Pressure Vessel A-3 Chemical Analysis of the Weld Metal used in t'ae Core Region Weld Seams of the Alvin W. Vogdt Unit No.1 Reactor Pressure Vessel A4 A4 TNDT RTNDT and Upper Shelf Energy for the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Shell Plates and Weld Metal A-5 A-5 Heat Treatment History of the Alvin W. Vogtle Unit No.1 Reactor Pressure Vessel Core Region Shell Plates and Weld Seams A-6 xi

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SECTION 1 PURPOSE AND SCOPE The purpose of this program is to monitor radiation effects under actual operating con- i

'ditions of the core region reactor vessel materisis in the Georgia Power Company, ,

Alvin W. Vogtle Unit No.1, a four-loop, nuclear po'ver plant with a thermal output rating of 3425 megawatts. Evaluation of the radiation effects is based on preirradiation testing of Charpy V-notch, tensile, and dropweight specimens, and postirradiation testing of Charpy V notch, tensile, and compact specimens,

' Current reactor pressure vessel material test requirements and acceptance standards

. utilize the reference nil-ductility temperature, RTNDT, as a becis. RT NDT is determined _

from the dropweight nil-cuctility transition temperature (TNDT) per ASTM E208 and the weakl 'I direction 50 ft Ib Charpy V-notch temperature (or the 35-mil lateral expan-e sion temperature Q it is greater). RTNDT is defined as the dropweight TNDT or the

, temperature 60*F. less than the 50 ft Ib (or 35-mil) Charpy V-notch temperature, whichever is graater.

' Therefore RTNDT = TNDT, if TN DT >.,. T50(35) - 60*F and RTNDT = T50(35) - 60*F, if T50(35) - 60*F > TNDT RNDT = Reference nil-ductility temperature

.TNDT = Nil-ductility transition temperatura per ASTM E208

~

~

-T50(35): = 50 ft'Ib temperature from.Charpy _V-notch specimens oriented in the weak direction (or the 35-mil temperature if it is greater) s

1. longitudinal aus of the, specemen onented normal to the major workmg jwacton of the f. ate.

• 1 x -

Pi emperical relationship between RTNDT and fracture toughness for reactor ves.sel steels has been developed in Appendix G," Protection Against Non-ductile Failure," to Section lli of the ASME Boiler and Pressure Vessel Code. This relationship can be cmployed to set allowable pressure-temperature limitations for normal operation of reactors which are based on f.acture mechanics concepts. Appendix G defines an iacceptable rethod for calculating these limitations.

It is known that radiation can shift the Charpy V notch impact energy curve to higher temperatures, [1,21 and thus cause the RT NDT to increase with radiation exposure. The extent of the shift in the impact energy curve, that is, radiation embrittlement, is enhanced by certain chemical elements (such as copper) present in reactor vessel steels.13'dl The adjustment in RTNDT with service can be monitored by a surveillance program involving periodic checking cf irradiated reactor vessel surveillance specimens. The sur-veillance program is based on ASTM E185-82 (Standard Practice for Conducting Sur-veillance Tests for Light Water _ Cooled Nuclear Power Reactor Vessels). Compact fracture mechanics specimens will be used in addition to Charpy V-notch specimens to evaluate

-the effects of radiation on the fracture toughness of reactor vessel materials.

Postirradiation testing of the Charpy V-notch impact specimens will provide a guide for determining pressure-temperature limits on the plant. Charpy impact test data will deter-mine the shift of the rciarence temperature l *I with radiation exposure at plant temperatures, a; The reference temperature as defined by 10CFR Part 50, Appendix G, Section Il-E is as follows:

" Adjusted reference temperature" means the reference temperature as adjusted for irradiation effects by adding to RTNDT the temperature shift, measured at the 30 ft Ib (41 J) level.

1. Porter. L F., "RadiatH>n Ettocts in Steet," in Marenats in Nuclear Apphcations. ASTM-STP 276, pp 147195, Amencan Society for Testing and "r'enais, Pndadelphia,1960.

2. stee!9, L E. and Hawthorne, J 9.. "New Information on Neutron Embntt:ement and Embnttlement Rehef of Reactor Pressure Vesse! Steels," NRL 6G d.ugust 1964.

3 Potapovs, U. and Hawtho ne, J. R... "The Effe = 1 Residual Elements on 550*F Irradiation Response of Selected Pressure Vessel Steels and Widments," NRL 6803, September 1968.

4. staele, L E. "strccture and Compositen Effects on irradiation Sensitrvtty of Pressure Vesset Steels,"it' irradsation Effects on StuctureMitcys for Nuclear 8icacy Appheations, ASTM sTP 484, pp.164-175, Amencan Society for ieating anti Matenals, Phdad lphia,1970.

i 1-2

i These data can then be reviewed to verify or revise pressure temperature limits of the vessel during heatup and cooldown and will allow a check of the predicted shift in the reference temperature. The postirradiation test results of the compact specimens will

. provide actual fracture toughness properties of the vessel material. These properties may  ;

be used to establish allowable stress intensity factors for subsequent analyses.

Six material test capsules are fabricated containing specimens from the reactor vessel shell plate identified as being most likely to limit the operation of the reactor vessel.

The specimens contained in the Alvin W. Vogtle Unit No.1 test capsules are from the intermediate shell plate of the reactor vessel and representative weld metal and heat-affected-zone (HAZ) metal.

The thermal history or heat treatment given these specimens is similar to the thermal

- history of the reactor vessel material with the exception that the postweld heat treatment -

received by the specimens has been simulated (Appendix A).

. The six material test capsules are then installed in the reactor in guide tubes attached to the neutron shield pads which are located in the reactor between the core barrel and the reactor vessel wall opposite the center of the core as shown in Figure 11.

1-3 4

o' REACTOR VESSEL CORE BARREL NEUTRON PAD (301,5 *) 2 g / ,, ,

CAPSULE U (58.5 *)

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b_Q_. CORE TEST CAPSULES IN THE  !  !

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ALVIN W. VOGTLE UNIT NO.1 (k REACTOR VESSEL

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9 SECTION 2 CAPSULE PREPARATION 2-1, PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by Ccmbustion Engineering, Inc. from interme-diate shell plate B8805-3, Heat No C0623-1. Combustion Engineering, Inc., also supplied a weldment which joined sections of material of the intermediate shell plate B8805-1 (see note) and the adjacent lower shell plate B8606-3, Heat No. C2085-2.

Data on the limiting core region plate (B8805-3), weld, and weld heat-affected zone material are provided in Appendix A.

Note: The limiting matertal for the AMn W. Vogtle Unit No.1 reactor vessel bettiine regon is imtermediate shell plate B8805-3 This is based on the Nghest ARTNOT Shift (96*F) as calculated using the latest ASTM revisens.

The original matenal selected in 19h was intermediate shell plate 08805-1. TNs selection was based at the time on the Nghest initial RTNDT. Therefore the Westinghouse survemance weld test plate "O" furnished by Combustion Engineer-ing at that Gme was made up of plates B88051 and D8606-3.

2 2. MACHINING Test material Jbtained from the intermediate shell plate (after the thermal heat treat-ment and forming of the plate) was taken at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the % and %-thickness location of the plate after performing a simulated postweld, stress-relieving treatment-on the test material and also from weld and heat-affected zone metal of a stress-relieved ,

weldment joining intermediate shell plate B8805-1 and adjacent lower shell plate 88606-3. All heat affected-zone specimens were obtained from the weld heat-affectede.one of intermediate shell plate B8805-1.

2.3 Charpy V notch impact Specimens-Charpy V-notch impact specimens corresponding to ASTM A370 Type A (Figure 2-1)_

were machined from intermediate shell plate B8805-3 in both the longitudinal orientation (longitudinal axis of specimen parallel to major rolling direction) and transverse orientation (longitudinal axis of specimen normal to major rolling direction). The core region weld Charpy impact specimens were machined from the weldment such that

/the long dimension of the Charpy specimen was normal to the weld direction. The-notch was machined such that the direction of crack propagation in the specimen was

. In the welding direction.-

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2 5.'. 1/2T Compact Speamens

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the longitudinal and transverse orientations. Compact test specimens frorn the weld n stal were machined with the notch oriented in the direction of weloing. All specimens were fatigue precracked.according to ASTM E399.

2-6. DOSIMETERS Each of the six test' capsules.of the type shown in Figure 2-4 contain dosimeters of copper, iron, nickel and aluminum 0.15 weight percent cobalt wire (cadmium shielded and unshielded) and cadmium-shielded Np 237 and U 238 which will measure the integrated flux at specific neutron energy levels.

u 2 7. THERMAL MONITORS The capsules contain two low-melting-point eutectic alloys to more accurately define the maximum temperature attained by test specimens during irradiation. The thermal monitors are sealed in ~ Pyrex tubes and then inserted in ~ spacers located as shown in Figure 2-4. - The ' two eutectic alloys and their melting points are the following:

2.5 percent Ag,97.5 percent Pb - Melting point: 304'C (579'F) 1.5 percent Ag,1.0 percent Sn,97.5 percent Pb Melting point: 310*C (590'F) d:- -,

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2 8. CAPSULE LOADING The six test capsules coded U, V, W, X, Y, and Z are positioned in the reactor between the neutron shielding pads and vessel wall at the locations shown in Figu e 2-4. Each capsule contains 60 Charpy V notch specimens,9 tensile specimens and 12 compact specimens. The relationship of the test material to the type and number of specimens in each capsule is shown in Table 2-1.

TABLE 2-1 TYPE AND NUMBER OF SPECIMENS IN THE ALVIN W. VOGTLE UNIT NO.1 SURVEILLANCE TEST CAPSULES Capsules U, V, W, X, Y, and Z Material Charpy Tensile Compact Plate B8805-3 15 Longitudinal (smr.,. c., cm., 3 4 Transverse 15 3 4 Weld Metal 15 3 4 HAZ 15 - -

Dosimeters of copper, iron, nickel, aluminum 0.15 weight percent cobalt, and cadmium-shielded aluminum-cobalt wires are secured in holes drilled in spacers located at capsule positions shown in Figure 2-4. Each capsule also contains a dosimeter block (Figure 2-5) located at the center of the capsule Two cadmium-oxide-shielded tubes, 238 237 one containing an isotope of U and the other an isotope of Np , are located in the

, dosimeter block.The double containment afforded by the dosimeter assembly prevents 237 loss and contamination by the U 238 and Np and their activation products. Each 23a dosimeter block contains approximately 12 milligrams of U and 17 milligrams of Np 237(Table 2-2) held in a 3/ 8 -inch-long by %-inch outside diameter sealed stainless steel tube respectively. Each tube was placed in a %-inch-diameter hole in the dosimeter block (one U 23a and one Np 237 tube per block), and the space around the tube was 2-9

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The numbering system for the capsule specimens and their locations is shown in Figure 2-6. The specimens are seal welded into a square capsule of austenitic stainless steel to prevent corrosion of specimen surf aces during irradiation. The capsules are hydro-statically compressed in demineralized water to collapse the capsule on the specimens so that optimum thermal conductivity between the specimens and the reactor coolant is obtained. The capsules are then leak-tested with helium after pressurization and then dye penetrant tested as a final inspection procedure. Fabrication cetails and testing procedures are listed in Figure 2-4.

TABLE 2 2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS isotope Weight (mg) Compound Weight (mg) 237 Np 17 1 NpO 2

20 11 U 23a 12.0 U0 3 8 14.25 2-11

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' I ews3 se sus sin et's s'?: etn s'e l s,es rw esse ru em sta Sws2 9,452 553 A s4 p?'a A ta ET71 k 71 9'as h6e 38791 8.45 6 's.1 M2 t'E 9' ' t 404 $?" # 154 twet SH61 kt3 8"3 BLt3 8"O Bt 'C ST67 l' 6a B.64 6't ' Siti g tit)

. 4. m2 v. w rs, ou, os. m. ni u, 14 ko

• T;s, l

tweF 8847 552 Slit 9?'A $L50 BTidi BtS6 ETM PLit STE 8tSO 874' P 4' Ghi fTO stia g f 7' 3T ,1 t*4 Steet 800 E?w 8650 $?S$ BL55 STM ILM Stat 8649 lie 6 B;46 t?tt Sw33 8m3 BLD 8f45 BA ret ke2 tus 6de s!)s e. 36 B?u ku *

• 37 9W12 6*t32 541 8t4 ST44 SL44 B741 BL41 9 ?10 Si2 tim SL35 802 St 32 5 '26 ST21 STi4 l'8 9't SW31 SH31 SL7 ST *3 864J 8?40 SL40 BT17 Rt37 9T34 tt34 t?31 6t3+ ST7

~1 Swie Sm3 6L4 ST1D BLJD B??? kJ' B?24 Bt24 8T2t Sat S?'8 But t?$

SW17 8Ht7 55C BL 6T29 809 6!3 9L2s t'z3 BL23 8 720 800 B? u SW Sit l'T BTS OTS OTS SW14 Mit SL4 ST78 BL28 8??S BL25 STU 8L22 87's k ti 6'18 kie #T4 9m DM3 90 8T15 f.L1$ S*12 8L12 BTs Bog sts ett gT3 ges et3

= u. ua in. 304 r,, e, c. .a ei .a n .a r. n n in n l SW1 BH1 Bu 8777 BL t3 8710 k te 977 SL7 8?4 Sta tit Sti ST1 ens BT-11, BT-12 and BT-13 were yielded Figure 2-6. Speciman Location in the

i. Extra specimens BT-25, BT-26 and BT-27 Alvin W. Vogtle Unit No. 2 1 their place. Reactor Surveillance Test Capsules 2-13/2-14 N.O ']h fhf Y ' b _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ . _

~

l SECTION 3 PREIRRADIATION TESTING

31. CHARPY V-NOTCH TESTS Charpy V-notch impact tests were performed according to ASTM E23 with specimens from the vessel intermediate shell plate 88805-3. Specimens of both longitudinal and transverse orientations were tested at various test temperatures in the range from - 40*C to 160'C (-40*F to 320*F), yeilding a full Charpy V notch transition temperature curve in both orientations (Tables 31 and 3-2 and Figures 3-1 and 3-2). Tests were also performed on the weld metal and HAZ metal at various temperatures from - 118'C to 160'C (- 180*F to 320'F) and are shown in Tables 3-3 and 3-4 and Figures 3-3 and 3-4.

A summary of the Charpy V-notch impact tests results including upper shelf energy (USE),

41-joule (30 ft Ib), 68-joule (50 ft lb), and 35 mils (0.89mm) lateral expansion index temperatures are presented in Table 3 5.

The specimens were tested on a SONNTAG UNIVERSAL impact machine, Model Number SI 1 with a hammer energy capacity of 240 foot pounds and a striking velocity of 17.02 feet per second. The machine is calibrated every 6 months using Charpy V notch impact specimens of known energy values supplied by Watertown Arsenal. Specimen condi-tioning for high temperature testing is maintained using a Fisher chest type ceramic fur-nace with a Newport temperature controller with direct digital temperature readout. For all low temperature specimen conditioning liquid Nitrogen is used. The specimen temperatures are monitored by the use of Chromel Aluminal thermal couples at high temperatures and by the use of Copper Constantan thermal couples at low temperature testing.

32. TENSILE TESTS Table 3-6 and Figures 3-5,3-6, and 3-7 show the results of tensile tests (per ASTM E8 and E-21 test criteria) from vessel intermediate shell plate B8805-3 and from the weld metal. Specimens from plate B8805-3 and the weldment were tested at 24*C (75'F),'

149'C (300*F) and 288'C (550*F) in both the longitudinal and transverse directions.

3-1

A SATEC UNIVERSAL tensile testing machine Model 120CS, serial number 1012, was used with a SATEC 120,000 lb. load cell as an integral part of the testing machine. The testing machine is calibrated daily and verified annually to the National Bureau of Stan-dards. The gripping mechanism utilizes threaded adapters to pull rods attached to the cross head / load cell and frame. The recording device utilizes a Hewlett Packard X-Y recorder and strip chart in console, serial number 7045A calibrated to a dual range high temperature extensometer, serial number BDRE 1. The extensometer is calibrated by test equipment which has been certified by the National Bureau of Standards. The measurement and control of speeds in the tests conform to ASTM A370-77 (Mechanical Testing of Steel Products). A typical stress strain curve is shown in Figure 3-8. For high temperature tests an Appliec Systems 3 zone type furnace was used with independent zone control. Temperatures were controlled by a Watlow, Model Number 3246-11 temperature controller utilizing type "K" thermal couples with direct digital temperature readout.

3-3. DROPWEIGHT TESTS The nil ductility transition temperature (TNDT) was determined for plate B8805-3 and the core region weld metal and heat affected-zone by dropweight tests (ASTM E-208) performed at Combustion Engineering, Inc. From tNs test data the RTNDT was calculated using the methods as described in Section 1. The TNDT and RT NDT for intermediate shell plate B8805-3, weld metal and heat-affected-zor,e (HAZ) are as follows:

Ncte TNDT and RT NDT for al the befthne sheu piates is gn,en n Append a A Material TNDT ( F) RTNDT ( F)

Plate B8805-3 - 20 30 Weld Metal (Longitudinal Seams - 80 I"I - 80 and closing Girth Seam)

HAZ - 90 # - 90

a. Combustion Engineering Welding Material Qualification Test D-32255

b. Combustion Engineering Surveillance Weld Test Plate "C" Materials Test Report 708194 3-2

I TABLE 31 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8805 3 (LONGITUDINAL ORIENTATION)

Temperature impact Energy Lateral Expansion Shear

('C) (*F) (J) (ft Ib) (mm) (mils) (%)

- 40 12.0 9.0 0.15 6.0 9

- 40

- 40 - 40 15.0 11.0 0.18 7.0 5

- 40 - 40 26.0 19.0 0.30 12.0 14

- 29 - 20 16.0 12.0 0.20 8.0 5

- 20 38.0 28.0 0.56 22.0 14

- 29

- 29 - 20 73.0 54.0 0.97 38.0 27

- 18 0 57.0 42.0 0.84 33.0 25

- 18 0 64.0 47.0 0.66 34.0 30

- 18 0 70.5 52.0 0.99 39.0 30 4 40 65.0 48.0 0.91 36.0 30 4 40 81.0 60.0 1.12 44.0 36 4 40 84.0 62.0 1.09 43.0 36 27 80 87.0 64.0 1.14 45.0 45 27 80 95.0 70.0 1.40 55.0 40 27 80 126.0 93.0 1.68 66.0 60 38 100 114.0 84.0 1.57 62.0 55 38 100 145.0 107.0 1.75 69.0 75 38 100 149.0 110.0 1.96 77.0 75 49 120 136.0 100.0 1.83 72.0o 80 49 120 148.0 109.0 1.88 74.0 85 49 120 157.0 116.0 2.06 81.0 85 82 180 156.0 115.0 2.08 82.0 100 82 180 157.0 116.0 2.03 80.0 100 82 180 171.0 126.0 2.13 84.0 100 127 260 164.0 121.0 2.11 83 0 100 127 260 175.0 129.0 2.34 92.0 100 160 320 161.0 119.0 2.16 85.0 100 160 320 178.0 131.0 2.16 85.0 100 3-3

TABLE 3 2 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)

Temperature Impact Energy Lateral Expansion Shear

('C) (* F) (J) (ft Ib) (mm) (mils) (%)

40 - 40 12.0 9.0 0.13 5.0 9 E

- 40 - 40 22.0 16.0 0.28 11.0 14

- 18 0 32.5 24.0 0.43 17.0 14

- 18 0 32.5 24.0 0.48 19.0 25

- 18 0 35.0 26.0 0.48 19.0 20 4 40 47.5 35.0 0.69 27.0 18 4 40 62.0 46.0 0.94 37.0 36 4 40 72.0 53.0 1.04 41.0 40 27 80 61.0 45.0 0.94 37.0 41 27 80 75.0 55.0 1.02 40.0 30 27 80 76.0 56.0 1.12 44.0 48 49 120 93.5 69.0 1.35 53.0 65 49 120 96.0 71.0 1.47 58.0 70 49 120 99.0 73.0 1.42 56.0 70 60 'i40 107.0 79.0 1.57 62.0 80 60 140 114.0 84.0 1.60 63.0 90 82 180 122.0 90.0 2.03 80.0 100 82 180 127.0 94.0 1.85 73 0 100 82 180 133.0 98.0 1.80 71.0 100 116 240 131.5 97.0 1.96 77.0 100 116 240 138.0 102.0 1.88 74.0 100 160 320 130.0 96.0 1.88 74.0 100 160 320 131.5 97.0 2.03 80.0 100 3-4 s

TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD ME*1M.

Temperature impact Energy Lateral Expansion Shear

('C) ('F) (J) (f t Ib) (mm) (mils) (%)

- 84 - 120 7.0 5.0 0.05 2.0 0

- 84 - 120 95 7.0 0.08 3.0 0

- 62 - 80 17.0 12.0 0.18 7.0 5

- 62 - 80 20.0 15.0 0.25 10.0 5

- 62 - 80 22.0 16.0 0.25 10.0 5

- 51 - 60 13.5 10.0 0.18 7.0 5

- 51 - 60 15.0 11.0 0.18 70 5

- 51 - 60 32.5 24.0 0.46 18.0 25

- 40 - 40 30.0 22 0 0.41 16.0 15

- 40 - 40 47.5 35.0 0.66 26.0 33

- 40 - 40 79.0 58.0 1.02 40.0 43

- 29 - 20 103.0 76.0 1.42 56.0 56

- 29 - 20 117.0 86.0 1.32 52.0 50

- 29 - 20 127.5 94.0 1.70 67.0 65

- 18 0 81.0 60.0 1.14 45.0 35

- 18 0 92.0 68.0 1.27 50.0 48

- 18 0 129.0 95.0 1.73 68.0 65 4 40 136.0 100.0 1.83 72.0 75 4 40 137.0 101.0 1.83 72.0 75 4 40 137.0 101.0 1.88 74.0 70 16 60 160.0 118.0 1.96 77.0 85 16 60 167.0 123.0 2.18 86.0 85 16 60 176.0 130.0 2.21 87.0 85 27 80 171.0 126.0 2.13 84.0 80 27 80 173.5 128.0 2.21 87.0 80 27 80 207.5 153.0 2.72 107.0 100 49 120 183.0 135.0 2.26 89.0 100 49 120 190.0 140.0 2.24 88.0 100 49 120 191.0 141.0 2.29 90.0 100 82 180 195.0 144.0 2.46 97.0 100 82 180 209.0 154.0 2.31 91.0 100 82 180 214.0 158.0 2.08 82.0 100 116 240 183.0 135.0 2.24 88.0 100 116 240 195.0 144,0 2.29 90.0 100 116 240 197.0 145.0 2.24 88.0 100 160 320 194.0 143.0 2.34 92.0 100 160- 320 194.0 143.0 1.96 77.0 100 160 320 195.0 144.0 2.18 86.0 100 3-5

TABLE 3-4 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT-AFFECTED ZONE MATERIAL Temperature Impact Energy Lateral Expansion Rhear

('C) (* F) (J) (ft Ib) (mm) (mils) (%)

- 118 - 180 5.5 4.0 0.03 1.0 0

- 118 - 180 8.0 6.0 0.03 1.0 0

- 96 - 140 5.5 4.0 0.05 2.0 0

- 96 7.0 5.0 0.08 3.0 0

- GO

- 79 -110 13.5 10.0 0.13 5.0 5

- 79 - 110 22.0 16.0 0.23 9.0 5

- 79 - 110 23.0 17.0 0.23 9.0 5

- 62 - 80 34.0 25.0 0.38 15.0 10

- 62 - 80 39.0 29.0 0.46 18.0 25

- 62 - 80 37.0 27.0 0.48 19.0 18

- 51 - 60 57.0 42.0 0.74 29.0 20

- 51 - 60 61.0 45.0 0.71 28.0 20

- 51 - 60 81.0 60.0 1.02 40.0 34

- 40 - 40 39.0 29.0 0.51 20.0 25

- 40 - 40 79.0 58.0 0.91 36.0 30

- 40 - 40 161.0 119.0 1.68 66.0 80

- 29 - 20 ts1.0 60.0 1.12 44.0 35

- 29 - 20 107.0 79.0 1.30 51.0 65

- 29 - 20 152.0 112.0 1.70 67.0 80

- 18 0 103.0 76.0 1.24 49.0 65

- 18 0 127.5 94.0 1.55 61.0 60

- 18 0 146.0 108.0 1.78 70.0 56 4 40 129.0 95.0 1.63 64.0 65 4 40 140.0 103.0 1.63 64.0 75 4 40 169.5 125.0 1.88 74.0 100 27 80 126.0 93.0 1.78 70.0 90 27 f.,9 153.0 113.0 1.88 74.0 90 27 80 199.0 147.0 2.06 81.0 100 49 120 169.5 125.0 1.96 77.0 100 49 120 184.0 136.0 2.01 79.0 100 49 120 184.0 136.0 1.98 78.0 100 104 220 171.0 126.0 2.06 81.0 100 104 220 190.0 140.0 2.13 84.0 100 104 220 190.0 140.0 2.21 87.0 100 3-6

. - _ _ _ _ - - - - _ _ _ _ - - _ _ - _ = _ _ - _ __ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l I

TABLE 3 5

SUMMARY

OF ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL IMPACT TEST RESULTS FOR INTERMEDIATE SHELL PLATE 88805-3 AND CORE REGION WELD AND HEAT AFFECTED-ZONE MATERIAL Upper Shelf 41 J 68-J 0.89 mm Energy (30-ft Ib) (50-ft Ib) (35 mils)

Material (USE) Index Temp Index Temp Index Temp (J) (ft Ib) (*C) (* F) ('C) (* F) ('C) ('F)

Plate B8805 3 (Longitudinal 165 122 - 26 - 15 -7 20 - 12 to Orientation)

Plate B8805-3 (Transverse 130 96 -9 15 18 65 13 55 Orientation) __

Weld 197 145 - 40 - 40 - 32 - 25 - 36 - 35 Heat Affected 184 136 - 59 - 75 - 46 - 50 - 43 - 45 Zone 3-7

1

+r' i

t i

TABLE 3-6 PREIRRA01ATtON TENSfLE PROPERTIES FOR THE BLVIN W. VOGTLE UNIT NO 1 SEACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 80805-3 AND CORE REGION WELD METAL j Reduction .

0.2% Uttimate Tensile Fracture Fracture Fracture Urdtosm ' Total in l Test Yleid Ares  !

Strength Load Strees Strorgth Erdongation Entongation Material Temperature Stren2th (ktp) (N) (ks4 (MPs) (kaJ) ( (MPa) (%) _ (%) (%)

• C '

'F (ksi) (MPa) (ksQ (MPa) 641 0 29 12.900 1950 1.344 0 59 0 4070 11 0 3( 0 70 0 Plate B8805-3 24 75 70 0 483 0 93 0 1850 1.276 0 57 0 393 0 11 0 23 0 69 0 -

(Longtudinal 24 75 68 0-, 469 0 90 0 621.0 29 12.454 500 0 28 12.454 1770 1.220.0 57.0 393 0 90 24 0 M0 Orientaton) .149 300 64 0 j 4410 85 0 182 0 1.256 0 57 0 393 0 70 23 0 99 0 149 300 77.0- 531.0 85 0 5860 26 12.454 62 0 4280 11 0 24 0 63 0 Y 288 550 63 0' 4340 90 0 621 0 30 30 13.344 1J.344 168 0 173 0 1.158 0 1.193 0 62 0 4280 10 0 25 0 64(s 288 55'; 62 0 4280 90 0 621.0 181 0 1.248 0 65 0 446 0 11 0 27 0 64 G P' ate B8805 3 24 75 71 0 490 0 94.0 6480 32 14.234 4r,0.0 94 0 6480 31 13.800 163.0 1.124 0 63 0 434 0 11 0 270 61 0 (Transverse 24 75 71 0 600 0 31 13.8CC 161 0 1.110 0 63 0 434 0 90 21 0 6f 0 Orientation) 149 300 65 0 448 0 870 30 13,344 1650 1.138 0 61.0 4210 90 20 0 63 0 149 300 64 0 44?.0 86 0 593 0 142.0 9790 64 0 4410 to 0 23 0 55 0 288 550 64 0 441.0 96 0 6270 31 13.800 152 0 1.048 0 71 0 490 0 11 0 22 0 53 0 268 550.- 64 0 441.0 92.0 6340 35 15.570 25 190 0 1,310 0 h00 345 0 90 25 0 73 0 24 75 72 0 496 0 64 0 579 0 11.170 190 0 1.310 0 51 0 3520 to 0 25 0 73 0 24 75 73 0 503 0 84 0 579 0 25 11.120 181.0 1.248 0 5* 9 3520 70 21.0 72 0 WeL1 M#31 149 300 67 0 462.0 78 0 5380 25 11.120 164 0 1.269 0 50 0 345 0 80 23 0 73 0 149 .300 67 0 4620 80 0 552 0 24 10.675 179f> 1,234 0 54 0 372 0 10 0 24 0 70 0 288 550 64 0 441.0 34 0 5790 26 11.564 174 0 1.200 0 52 0 3590 to 0 24 0 70 0 288 550 M0 448 0 84 0 5790 26 11.564

TEMPERATURE ('C) 100 50 0 50 100 150 200 I i l I I I I -

2w 140 -

180 O

o 8 120 -

fg O -

160 140

$ 100 -

O

/0 120 _

O 80 --

100 ~3-f 60 -

O /0

~

- 60 80 5 _

40 0

/

20 -

- 20

~

I I I I I I 0 O

-200 100 0 100 200 300 400 TEMPERATURE (*F)

FIGURE 31. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE D8805-3 (LONGITUDINAL ORIENTATION)

TEMPERATURE ('C) 100 -50 0 50 100 150 200 l l I I I I I -

180 120 - -

160

~

'" 9-

@8

-~

h 3 _

120

,[

h 30 -

100

> BD

- /" - 80

~

40 -

g 2 70 _ 4o 20 -

_ 20 0 l' I  ! .

I I I O

200 100 0 100 200 300 400 TEMPERATURE (*F)

FIGURE 3 2. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION) 3-9

1 TEMPERATURE ('C) 100 50 0 50 100 150 200

'80 l I I I I I I 160 -

- 220 0 0 - 200 140 -

/f-' 0 O 2 180  ;

g 120 - 160

} _

.. 2 _ 140

[ go - 120 3 80 "-

. @ O - 100 w 0 z 60 -

oO - 80

~

40 -

l O' _ 4o 20 - O _ 20 0

- I I  !  ! ' 0

-200 100 0 100 200 300 400 4

TEMPERATURE (*O FIGURE 3-3. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL TEMPERATURE ('C) 150 100 - 5(, 0 50 100 150 160 g g g g g 2 2 t O -

200 140 -

s 180 120 -

O -

160 O O k 100 - O -

140

f OOo --

120

{ 80 60 00 00 -

100 80 h 0 -

0 40 -

0 -.

40

' 20 -

/

20 0 I ' I I  ! I O

-300 -200 -100 0 100 200 300 TEMPCFIATURE (*F)

. FIGURE 3-4. PRE 18 RADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE Al. VIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL WELD HEAT-AFFECTED-ZONE MATERIAL 3-10

TEMPERATURE ('C) 0 50 100 150 200 250 300

" I I I I I I I 100 -

2 (EPECIMENS) ,

.y 90 -

8 'N '-

x 600 gg -

ULTIMATE TENSILE STRENGTH [

$ o a W 70 -

500 E

M O sg 60 -

0 2% YlELD STRENGTH -

400 50 -

40 1 l l l l -

300 0 100 200 300 400 500 600 TEMPERATURE ('F) k.

TEMPERATURE (*C) 0 50 100 150 200 250 300 gg l I I I I I i 70 -

@ - Q --

60 -

REDUCTION IN AREA 9 -

5 E 50 -

b 40 -

O 30 - 9 TOTAL ELONGATION U

g N g -a~

20 - 2 UNIFORM ELONGATION 10 -

w- g -@

I I ' I I 0

O 100 200 300 400 500 600 TEMPERATURE ( F)

FIGURE 3-5. PREIRRADIATION TENSILE PROPERTIES FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 88805-3 (LONGITUDINAL ORIENTATION) 3-11

TEMPERATURE ('C) 0 50 100 150 200 250 300 l l l l l l l g _2 -

700 90 -

Q -

600 80 -2 ULTIMATE TENSILE STRENGTH c

$ 7 -

2 500 g

60 N.__ g 1-0.2% YlELD STRENGTH -

400 50 -

40 1 I I I I -

300 0 100 200 300 400 500 600 TEMPERATURE (*F)

TEMPERATURE ('C)

O 50 100 150 200 250 300 80  ;  ;  ;  ;  ; j 70 -

60 -

O EF 50 -

REDUCTION IN AREA

.; 40 -

p 2

@ 30 -

w TOTAL ELONGATION n

20 2 b 9

\ 2N UNIFORM ELONGATION 10 -

6 b 9 o I I I I I O 100 200 300 400 500 600 TEMPERATURE (*F)

FIGURE 3-6. PREIRRADIATION TENSILE PROPERTIES FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 88805 3 (TRANSVERSE ORIENT ATION) 3-12

TEMPERATURE (*C) 0 50 100 150 200 250 300

0 l I I I l I I 399 _

700

~

e -

600

~

.E

~ 80 - -

0 --

v e Q ULTIMATE TENSILE STRENGTH - 500 h

e

~

2 / ~9 0 2% YlELD STRENGTH 400 50 -

40 l l l l l -

300 0 100 200 300 400 500 600 TEMPERATURE (*F)

TEMPERATURE (*C) 0 50 100 150 200 250 300 80  ; i  ;  ;  ;

c 8 ~o 2)- REDUCTION IN AREA 2 60 -

3 al 50 -

t- 40 -

J 2 h 30 -

TOTAL ELONGATION D }

O 20 -

g O UNIFORM ELONGATION 10 -

@ g 3 I I ' I I O 100 200 300 400 500 600 TEMPERATURE (*F)

FIGURE 3-7. PREIRRADIATION TENSILE PROPERTIES FOR THE ALl'N W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3-13

~

l l  :

l m l

<n  !

w -_

T- :

a W '

m ,

u \

I  ! l  ! I I I i l

STR AIN Figure 3 8. Typical Stress-Strain Curve for Tensile Test F 1MmM WmM

\

SECTION 4 POSTIRRADIATION TESTING 4-1. CAPSULE REMOVAL The first capsule (Capsule U) should be removed at the end of the first core cycle (1st refueling) as shown in Table 4-1. Subsequent capsules should be removed at 5,9 and 15 EFPY (Effective Full Power Years) as indicated. Each specimen capsule, removed after exposure, will be transferred to a postirradiation test f acility for disassembly and testing of all the specimens. \

- TABLE 4-1 SURVEILLANCE CAPSULE REMOVAL SCHEDULE Orientation Capsule of Lead Removal Expected Capsule 2

identification CapsulesI 'l Factortbl Time Fluence (n/cm )

U 58.5' 4 00 ':efueling 4.69 x 10*

Y 241

• 3.09 >EFPY 1.73 x 10*kl V 61

• 3.69 9 EFPY 3.11 x 10"Idl X 238.5* 4.00 15 EFPY 5.63 x 10" W 121.5* 4.00 Stand-By --

Z 301.5* 4.00 Stand-By

a. Reference Irradiation Capsule Assembly Drawing, Figure 2-4.

b. The factor by which the capsule fluence leads the vessels maximum inner wall fluence.

c. Approximate Fluence at %-wall thickness at End of-Life.

d. Approximate Fluence at vessel inner wall at End-of-Life.

4-1

4 2. CHARPY V NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the intermediate shell plate G8805-3 weld metal, and HAZ metalin each capsule can be done singly at approximately ten different temperatures. The extra specimens should be used to run duplicate tests at temperatures of interest to develop the complete Charpy impact energy transition curve.

The initial Charpy specimen from the first capsule removed should be tested at room temoerature. The test value of this temperature should be compared with preirradiation test data. The test temperature for the remainina specimer.3 should then be adjusted higher or lower so as to develop a complete transition curse. For succeeding tests after longer irradiation periods, the test temperature in each chse should be chosen in the

!!ght of results from the previous capsule.

4.3 TENSILE TESTS A tensile test specimen from each of the selected irradiated materials shall be tested at a temperature representative of the upper end of the Charpy energy transition region.

The remaining tensile specimens from each material shall be tested at the service temperature (550*F) and the midtransition temperature.

4.4 FRACTURE TOUGHNESS TESTS ON 1/2 COMPACT SPECIMENS in light of current requirements of 10CFR, Part 50, Appendix G and applications of ASME Secition 111, Apendix G and Section XI, Appendix A, the %-inch thick compact specimens should be tested in such a manner as to determine both static, crack initiation, and propagation parameters throughout the temperature range of interest with emphasis on the sharp fracture toughness transition and upper shelf regions consistent with specimen availability The specimens should thus be statically tested in accordance with ASTM E399-81 procedures modified to account for the size of the specimens available.I4 Specific test procedures should include unloading compliance and data interpretation should utilize the Equivalent Energy and J Integral concepts.M'I 1 Witt. F J , "Fiacture Toughness Parameters Otatained from Singie Smail Speomen Tesss". WCAP.9397. October 1978

2. Buchatet C. and Mager. T R. "Expenmental Venf<ation of Lower Bound gK Vaues Utung the Equrvaet Energy Concept, in Prog ess in Flaw Growth and Fracture Toughness Testing. ASTM-STP-536, pp 281-296 Amencan Somety for Test ng and Matenals. Philadelphia.1973 3 Landes, J D and Begley, J. A . "Recent Devet ..aents in J Testing 1 in Deveepments in Fracture Mechan,cs 7csf M,nhods Standsdaaten. AdTM STP 632. po 57-81, Amu can Gooety for Test 4ng and Matonals. Philadeiphia,1977.

4 McCabe. D E . " Determination of R-Curves for Shoctural Matenals Us.ng Nonlinear Mechantcs Methods," in FOw Growth and Fracture. ASTM.STP-631. pp 2450.26. Amencan Gooety for Testing and Matenats. Philadelphia,1977, 4-2

i Fracture toughness data so obtained will be Kic, Jic and dJ/da or engineering estimates thereof. Advantages should be taken of the Charpy impact and tensile data in the selec-tion of initial test temperatur1s. Test procedures actually performed on the specimens will reflect state-of-the-art at the time of testing.

4,5 POSTIRRADIATION TEST EQUIPMENT Required minimum equipment for the postirradiation testing operations is as follows-W Milling machine or special cutoff wheel for opening capsules, dosimeter blocks and spacers.

M Hot cell tensile testing machine with pin-type adapter for testing tensde specimens.

M Hot cell static CT testing machine with clevis and appropriate measuring equipment modified to account for the size of the specimens.

E Hot cell Charpy impact testing machine.

E Sodium iodide scintillation detector ard pulse height analyzer for gamma counting of the specific activities of the dosimeters.

4-3

APPENDIX A DESCRIPTION AND CHARACTERIZATION OF THE ALVIN W. VOGTLE UNIT NO.1 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS Based on the initial RTNCT, chemical compostion (copper and phosphorus) and the end-of-life neutron fluence, the reactor vessel intermediate shell plate 88805-3 is expected to have the highest u,,d of-life a RTNDT using the prediction methods of Regulatory Guide 1.99 Revision 1. and latest ASTM revisions. This material is therefore considered to be the limiting vessel bettine region material and has been used in the reactor vessel surveillance program.

For the surveillance program Combustion Engineering, Inc., supplied Westinghouse with sections of the A533 Grade B C; ass 1 Steel plate produced by Lukens Steel Company. This steel was used in the f abrication of the Alvin W. Vogtle Unit No.1 ieactor pressure vessel, specifically, from the 9& inch intermediate shell plate B8805-3. Also supplied was a submerged arc weldment made from sections of intermediate shell plate B8805-1 {al and adjacent lower shell plate B8606-3, This test weldment was fabricated using % inch Mit B 4 weld filler wire, heat number 83653 and Linde 0091 flux, tot number 3536 and is identictl to that used by Combustion Engineering. Inc.

in the Alvin W. Vogtle Unit No.1 reactor vessel fabrication process specifically the closing girth seam between the intermediate and lower shell plates, and all longitudinal weld seams of both the intermediate and lower shell olates.

~

The chemical analyses, TNDT, RTNDT, upper shelf energy and heat treatment history of all the core region pressure vessel shell plates used in the fabrication of the Alvin W. Vogtle Unit No.1 reactor pressure vessel are summarized in Tables A 1 thru A-5 respectively. This data is as reported in the vessel fabricators (Combustion Engineer-ing, Inc.) certification reports or from subsequent Westinghouse analyses of similar materials used for the Alvin W. Vogtle Unit No.1 surveillance program. Weld material identical to that used in the fabrication of the core region beltline weldsib] have been correlated with the Westinghouse surveillance program test weldment "D" and available Combustion Engineering, Inc. weld certificMion reports arid their surveillance program test weldment "C", This data is also rv 'd in Tables A-3 thru A-5 of this Appendix,

a. The hmiting plate matenal selecied in 1974 w( . ,rmediate shell plate B8805-1 Th4s setection was based at the timo on the highest initial RTr4DT. Therefore weld test plate "D" furnished to Westinghouse at inat time was made up of plates B8805-1 and BS60C-3

b. The beitrme welds are considered to include the intermediate and lower shell plate longitudinal seams and the Ciosing intermediate to lower shell girth seam A-1 l

TABLE A 1 CHEMICAL ANALYSIS OF THE INTERMEDIATE SHELL PLATES USED IN THE CORE REGION OF THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL Chemical Compositon (weight %)

Element Plate *I Plate B88051 88805 2 I 08805 3 *IIDI B8805 31*II*I C .24 .24 .25 .22 Mn 1.37 1.37 1.32 1.32 P .004 .004 .003 .017 S .011 .011 .010 .011 Si .23 .23 .26 .28 Ni .59 .59 .60 .61 7

Mo .56 .56 .53 .57 Cr .08 .07 .04 .057 Cu .08 .08 .06 .058 Al .025 .024 .029 .030 Co .012 .012 .009 .006 Pb <.001 <.001 < 001 <.001 W <.01 <.01 <.01 <.01 Ti <.01 <.01 <.01 C04 Z' < 001 <.001 <.001 <.002 V .003 .003 .003 <.002 Sn .004 .004 .017 .019 As .002 .002 .001 .003 Cb <.01 <.01 <.01 <.002 N2 .007 .008 .008 .006 B < 001

. <.001 < 001

. < 001

a. Surveillance program test plats,

b. Chemical Analysis by Combustion Enginee.ing, Inc.

c. Chemical Ana . ='s by Westinghouse.

A2

i TABLE A 2 CHEMICAL ANALYSIS OF THE LOWER SHELL PLATES USED IN THE CORE REGION OF THE ALVIN W. VOGTLE UNIT NO,1 REACTOR PRESSURE VESSEL Chemical Compositon l'1 Element Plate Plate Plate 08606 2 88606 3 B86061 -_

.23 .22 C .20 1.35 1.38 Mn 1.32

.009 .007 P .005

.016 .012 S .010

.21 .25 Si .22

.58 .64 Ni .59

.56 .56 Mo .55

.02 .03 Cr .04

.05 .06 Cu .05

.03 .033 Al .028

.010 .007 Co .005

< 001 nm nww <.001 Pb

<.01 <.01 <.01 W <.01 Ti <.01 <.01

.002 .001 Zr .001

.003 .003 V .003

.003 .004 Sn .001

.006 .00S As .003

<.01 <.01 <.01 Cb

.009 .008 N2 .007

<.001 <.001 <.001 B

a. Chemical Analysis by Combustion Engineering, Inc.

A-3

TABLE A 3 CHEMICAL ANALYSIS OF THE WELD METAL USED IN THE CORE REGION WELD SEAMS OF THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL Not t in. m. ,,gon inom n,i .co. .,. ton..e...a innavo. e , nie,%<, emo r, theil [ dale If)ngitudinAI $@ nml and (5g yo,hing lYprmeia'e 10 k)*,, bhed Q 15 Seam u ec... egen it+innei ..e .. v.o,u.o vs.ng w. o w.r. s..t 90 036n t.no, 009t rius, Lot No 3s.T6 Chemical (weight Comp)ositon 9o Element Wire Flux Actual Production Westinghouse Test Weld Weld (Intermediate Surveillance Sample l *l To Lower Shell Gir1h Program Test Seam,101 171)lbl Weldment "D"I'l C .14 .09 .13 Mn 1.06 1.17 1.15 P .007 .008 .017 S .009 .009 .010 Si .16 .17 .19 Ni ---

.10 .10 Mo .52 .63 .61 Cr ---

.05 .052 Cu .03 .04 .037 Al =

.009 .002 Co .01 .005 Pb ---

<.001 < 001 W ---

.02 <.01 Ti -

<.01 .006 Zr --

.001 < 002 V .005 ,007 .003 Sn .003 <.002 As -

.006 .004 Cb ---

.01 <.002 N2 -

.021 ,003 8 =

<.001 <.001

a. Chemical Analysis of Wire-Flux Weld Sample. Test Number D32255 by Combustion Engineenng. Inc.

b Chemistry Data Sheet for Final Vesse' Assembly. Contract 8971. Job 708171 Control Number 009, Weld 101171 by Combust.on Engineenng. Inc

c. Chemical Analysis by Westinghouse of Weld Test Plate "0" Supphed by Combustion Engineenng. Inc Representative of the Closing Girth Seam Weid A-4

i I

TABLE A 4 TNDT,RTNDT AND UPPER SHELF ENERGY FOR THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES AND WELD METAL p Upper Shelfl 'Il'I T NDT )gbj RTNDT Energy Material

('C) (*F) ('C) (*F) (J) (f t Ib)

Intermediate Shell Plates:

-18 0 - 18 0 122 90 B8805-1

- 23 - 10 -7 20 136 100 B8805-2

- 29 - 20 -1 30 145 107 B8805 3 Lower Shell Plates:

-46 - 50 -7 20 157 116 B8606-1

- 23 - 10 -7 20 153 113 B8606 2

- 29 - 20 - 12 10 160 118 B8606 3 a Data obtained from Combustion Engineenng. Inc. Reactor Vessel Material Certstication Repons b 3 rop weight data OLiained from the transverse material propert es (normal to the maior wor 6ing dn choni

c. from impact data obtained from the transverse mater <at properties (normal to tne major work.ng d,re: tron) to Upper Shelfl 'I TNDT) RTNDT Energy Materia!

('C) ('F) ('C) ('F) (J) (f t Ib)

Intermediate and Lower She!I longitudinal Weld Seams and Closing Gir'.h - 62 - 80 - 62 - 80 182 134 Weld Seam (Weld Wire Heat No. B3653. Linds 0091 Flux, Lot No. 3536)

d. Data obtained from Con bustion Engineenng, Inc. Welding Matenal Quahfication Job Numbet 032255.

Project No 960009

e. Data obtained from Combust:on Engineennc inc. Matenais Com0 cation Report for their Weid Test Plate ' C" A5

l TADLE A 5 HEAT TREATMENT HISTORY OF THE ALVIN W. VOGTLE UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES AND WELD SEAMS Temperature Timel'1 Material (* F) (br) Cooling Austenitizing: 4 Water-quenched 1600 1 25 Intermediate (871'C)

Shell Plates Tempered: 4 Air cooled B8805-1 1225 25 -

B8805 2 (663'C)

B8805-3 Stress Relief: 17.5*l Furnace cooled 1150 i 50 (621'C)

Austenitizing: 4 Water-quenched 1600 i 25 Lower (871'C)

Shell Plates Tempered: 4 Air cooled B86061 1225 1 25 B8606 2 (663*C)

B8606-3 Stress Relief: 15.5bl Furnace-cooled 1150 i 50 (621 *C)

Intermediate Shell Longitudinal Stress Relief: 17.5bl Furnace-cooled Seam Welds 1150 1 50 (621 *C)

Lower Shell Longitudinal 15.5bl Furnace-cooled Seam Welds Local Intermediate to Stress Relief: 12.75 Furnace-cooled Lower Shell Girth 1150 1 50 Seam Weld (621'C)

Surveillance Program Test Material Surveillance Program Weldment Test Post Weld Piate ,D;',

n, , ,,n , Stress Relief 12.75 tcl Furnace cooled cios<ng oinn seam) 1150 1 50 (621 *C) i Lukens Steel Company, Comoustion Engineering, Inc. Certtication Reports.

b Stress Relief includes the intermediate to Lower Sh.il Closing Girth Seam Post Wald Heat Treatment.

c. The Stress Rehet Heat Treatment received by the burveillance Test Wetdment has been simu!ated A-6

__ ___ - _ _ _ _ _ _ _ _