ML20101M531

Gap Aw Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program

ML20101M531
Person / Time
Site:	Vogtle
Issue date:	02/28/1992
From:	Meyer T GEORGIA POWER CO.
To:
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ML20101H833	List: ML20101H830 ML20101M531 ML20101M541
References
WCAP-11011, NUDOCS 9207080273
Download: ML20101M531 (27)

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! WCAP 11381 WESTINGHOUSE CLASS 3 GEORGIA POWER COMPANY ALVIN W. VOGTLE UNIT NO. 2 '-

REACTOR VESSEL RADIATION

' SURVEILLANCE PROGRAM L. R. Singer April 1986 APPROVED: IA ASW !Me d -

T. A. Meyer,EAanader Structural Materials And Reliabili.ty Technology Work Performed Under GBEJ-106 WESTINGHOUSE ELECTRIC CORPORATION Generation Technology Systems Division P. O. Box 2728 Dittsburgh, Pennsylvania 15230

i_

I PREFACE This report has been technically reviewed and checked by S. E. Yanichko of Structural Materials and Reliability Technology.

/f 0ys c, bl&

S. E.sYanichko Date: April 14,1987 iii

l ABSTRACT A pressure vessel steel surveillance program per ASTM E-185-82 has been developed for the Georgia Power Company, Alvin W. Vogtle Unit No.2 to obtain information on it.e effects of radiation on reactor pressure vessel material under operating conditions. The radiation surveillance program for the Alvin W, Vogtle Unit No. 2 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 materialinvolved, the specimen and capsule design and fabrication, and the preirradiation test results, v

l TABLE OF CONTENTS Title Page Section 1-1 1 PURPOSE AND SCOPE 2-1 2 CAPSULE PREPARATION 2-1 2-1. Pressure Vessel Material 2-2. Machining 21 2 3. Charpy V-notch Impact Specimens 21 2-4. Tensile Specimens 2-3 2 5. 1/2T Compact Specimens 2-3 Dosimeters 2-3 2-6.

Thermal Mcnitors 2-3 2.7.

2.8. Capsule Loaoing 29 3-1 3 PREIRRADIATION TESTING 3 Charpy V-notch Tests 31 Tensile Tests 3-1 3-2.

Dropweight-Tests 3-2 3-3.

4-1 4 POSTIRRADIATION TESTING Capsule Removal 4-1 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 Specimens 4-2 Postirradiation Test Equipment 4-3 4-5.

Appendix A DESCRIPTION AND CHARACTERIZATION OF THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR VESSEL A-1 BELTLINE AND SURVEILLANCE MATERIALS vii

l LIST OF ILLUSTRATIONS Title Page Figure 1-1 Location of the Irradiation Test Capsules in the Alvin W. Vogtle Ur'it No. 2 Reactor Vessel 1-4 Charpy V notch Impact Specimen 2-2 2-1 Tensile Specimen 2-4 2-2 2-3 Compact Specimen 25 2-4 Irradiation Capsule Assembly 2-7/2-8 1 2-5 Dosimeter Block Assembly 2 10 2-6 Specimen Locations in the Alvin W. Vogtle Unit No. 2 Reactor Surveillance Test Capsules 2-13/2-14 3-1 Preirradiation Charpy V-not i impact Energy for the Alvin W. Vogtle Unit No. 2 Heactor Pressure Vessel Lower Shell Plate B8628-1 (Longitudinal Orientation) 3-9 3-2 Preirradiation Charpy V notch Impact Energy for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 3-9 (Transverse Orientation) 3-3 Preirradiation Charpy V-notch Impact Energy for the Alvin W. Vogtle Unit No. 2 Reactor Pressore Vessel Core Region Weld Metal 3-10 3-4 Preirradiation Charpy V-notch Impact Energy for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat-Affected-Zone Material 3-10 Preirradiation Tensile Properties for the 3-5 Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 (Longitudinal Orientation) 3-11 3-6 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 3-12 (Transverse Orientation) 37 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-13 3-8 Typical Stress-Strain Curve for Tensile Test 3-14 iX

l l LIST OF TABLES Title Page Table 2-1 Type and Number of Specimens in the Alvin W. Vogtle Unit No. 2 Surveillance Test Capsules 2-9 Quantity of Isotopes Contained in the Dosimeter Blocks 2-11 2-2 3-1 Preirradiation Charpy V notch Impact Data for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 (Longitudinal Orientation) 3-3 3-2 Preirradiation Charpy V notch Impact Data for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 3-4 (Transverse Crientation) 3-3 Preirradiation Charpy V-notch Impact Cata for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-5 .

3-4 Preirradiation Charpy V-notch impact Data for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat iffected-Zone Material 3-6 3-5 Summary of the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Impact Test Results for Lower Shell Plate B8628-1 and Core Region Weld and Heat-Affected-Zone Material 3-7 3-6 Preirradiation Tensile Properties for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Lower Shell Plate B8628-1 and Core Region Weld Metal 38 Surveillance Capsule Removal Schedule 4-1 4-1 A-1 Chemical Analysis of the Intermediate Shell Plates used in the Core Region of the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel A-2 A-2 Chemical Analysis of the Lower Shell Plates used in the Core Region of the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel A-3 A-3 Chemical Analysis of the Weld Metal used for the intermediate and Lower Shell Plates Longitudinal Seams for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel A-4 A-4 Chemical Analysis of the Weld Metal used for the Intermediate to Lower Shell Closing Girth Seam of the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel A-5 A-5 TNDT,RTNDT and Upper Shelf Energy for the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Shell Plates and Weld Metal A-6 A-6 Heat Treatment History of the Alvin W. Vogtle Unit No. 2 Reactor Pressure Vessel Core Region Shell Plates ,

and Weld Seams A-7 xi

SECTION 1 PURPOSE AND SCOPE The purpose of this program is to monitor radiation effects under actual operating con-ditions of the core region reactor vessel materials in the Georgia Power Company, ,a Alvin W. Vogtle Unit No. 2, a four-loop, nuclear power plant with a thermal output rating ,

of 3425-megawatts. Evaluation of the radiation effects is based on preitradiation 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 stendards utilize the reference nil-ductility temperature, RTNDT, as a basis. RT NDT is determined from the dropweight nil-ductility transition temperature (TNDT) per ASTM E208 and the weakI4 direction 50 ft Ib Charpy V-notch temperature (or the 35-mil lateral expan-sion temperature if it is greater). RT NDT is defined as the dropweight TNDT or the ternperature 60*F less than the 50 ft Ib (or 35-mil) Charpy V noten temperature, whichever is greater Therefore RTNOT = TNDT if TNDT 3 T50(35) - 60*F and RTNDT = T50(35) - 60*F, if T50(35) - 60*F > TNDT where RNDT = Reference nil-ductility temperature TNDT = Nil-ductility transition temperature per ASTM E208 T = 50 ft Ib temperature from Charpy V-notch specimens oriented 50(35) in the weak direction (or the 35 mil temperature if it is greater)

1. Longitudinal avis of the specimen ottanted norrnal to the rnaqor working direchon of the plate 1-1

An emperical :alationship between RTNDT and fracture toughness for reactor vessel steels has been developed in Appendix G," Protection Against Non-ductile Failure," to Section til of the ASME Boiler and Pressure Vessel Code. This relationship can be employed to set allowable pressure temperature limitations for normal operation of reactors which are based on fracture mechanics concepts. Appendix G defines an acceptable method for calculating these limitations.

It is known that radiation can shift the Charpy V notch impact energy curve to higher temperatures, I"I 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.l"I _

The adjustm3nt in RTNDT with service can be monitored by a surveillance program involving periodic checking of irradiated reactor vessel surveillance specimens. The sur-veillance prcyram is based on ASTM E185 82 (Standard Practics for Conducting Sur-veillance Tests for Light Water Cooled Nuclear Power Reacar Vesse's). Compact fracture mechanics specimens will be used in addition to Chnny " intch specimens to evaluate the effects of radiation on the fracture toughness of reactor vessel materials.

Postirradiation testing of the Charpy V-notch irtpact specimens will provide a guide for determining pressure-temperature limits on the plant. Charpy impact test data will deter-mine the shift of the reference temperaturefalw '.h iadiation 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 irracation effects by adding to RTNDT the temperature sh;ft, measured at the 30 ft Ib (41 J) level.

I Porter, L F , ' Radiation Effects in Steel." m Marenals en Nuclear Apppcafjons. ASTM STP 276 pp 147 191 Amencan Sonety tor Testing and Matenats. Phdadelphia,1960

2. SteMe. L E and Hawthorne. J R . "New information c6 Nec ron Embottlement and Emonttlement Refe' of Reactor Pressure Vee. 91 Steels," NRL 6160, August 1964 3 Potapov% U and Haw'..orne. J R . "The Elk it of Resulual Dements on 550*F irradiation Response of Seiected Pesu re Vessel S . .s and Weidments." NRL 6803. Setemt..r 1968 4 Steein. L G , " Structure and Composition Effects on liradiatic,n Sensdivity of Pressure Vessel Steels? in lerasabpn EfWts on StoSturaI Atleys for Nacicar Rextor Apphcations ASTM ST P-484. pp 164-175

• mencan Saoety for Testmg and Mateoais F uladelphia.1970 12

t These data can then be reviewed to .erify or revise pressure temperature limits of the l i

vessel during heatup and cooldown and will allow a check of the predicted shif t in the reference temperature. The postirradiation test results of the compact specimens will provide Jctual fracture toughness properties of the vessel material. These properties may be used to establish allowable stress intensity factors for subsequent analyses.

I Six material test capsules ara fabricated containing specimens from the reactor vessel

[

i shell plate ideritifie6 as being most likely to limit the operation of the reacter vessel. ,

The specimens contained in the Alvin W. Vogtle Unit No. 2 test capsules are from the lower shell plate of the reactor vessel and representative weld metal and heat affected-zone (HAZ) metal.

- The thermal history or neat troaiment 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 rnatorial test capsules are aen installed in the reactor in guide tubes attached to the neutron shield pads which are located in the reactor between the core barrel and ,

)he reactor vessel wall opposite the center of the core as shown in Figure 11. ,

b d

v

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PLAN VIEW b 9 ~ VESSEL Q,/ WALL f  ! ;. [ CAPSULE RE j s' / ASSEMBLY FIGURE 1 1, LOCATION OF THE IRRADIATION \ 11111lllljllJ i

,_ . . .1,, CORE TEST CAPSULES IN THE -

N MIDPLANE

' l ALVIN W. VOGTLE UNIT NO. 2 [

, 2

. REACTOR VESSEL \ L}h NEUTRON PAD kk 1  :

/ CORE BARREL ELEVATION VIEW 1-4

SECTION 2 CAPSULE PREPARATION

21. PRESSUDE VESSEL MATERIAL Reactor vessel material was supplied by Combustion Engineering, Inc. from lower shell plate B86281, Heat No. C3500 2 Combustion Engineering, Inc., also supplied a weldment which joined sections of material of the lower shell plate B86281 and the adjacent lower shell plate B88251, Heat No. C35001. Data on the limiting core region plate (B86281), weld, and weld heat affected zone material are provided in Appendix A.

2 2. MACHINING Test material obtained f rom the lower shell plate (after the thermal heat treatment 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 % 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 lower shell plate B86281 and adjacent lower shell plate B8825-1. All heat-affected zone specimens were obtained from the weld heat affected zone of lower shell plate B8628 '

2.3 Charpy V notch Impact Specimens Charpy V-notch impact specimens corresponding to ASTM A370 Type A (Figure 21) were machined from lower shell plate B86281 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.

a. The compact test specimens wete obtained from the %T thickness Location of the plate.

21

460 44 0

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2.105 ALL OVER UNLESS OTHERwl5E $P(Ctf IED Figure 21, Charpy V notch Impact Specimen 2*2

2-4. Tetisile Specimens Tensile specimens (Figure 2 2) from shell plate B86281 were machined in both the longitudinal and transverse orientation. Tensile specimens from the weld were oriented norrnal to the welding direction.

25. 1/2T Compact Specimens Compact test specimens (Figure 2 3) from shell plate B86281 were machined in both the longitudinal and transverse orientations. Compact test specimens from the weld metal were machined with the notch oriented in the direction of welding. 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 237 238 and unshielded) and cadmium-shielded Np and U which will measure the integrated flux at specific neutron energy levels.

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 moi;itors are sealed in Pyrex tubes and then inserted in spacers located as shown in Figure 2-4. The two outectic 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) 2-3

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. 9 eoa Figure 2 4. Irradiation Capsule Assembly m ru .n -,mema From Westinghouse Dwg 1453E10 2 7/2 8 A

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 Figure 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 I in each capsule is shown in Table 21.

l TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE ALVIN W. VOGTLE UNIT NO. 2 SURVEILLANCE TEST CAPSULES Capsules U, V, W, X, Y, and Z Material Charpy Tensile Compact Plate B8628-1 15 3 4 Longitudinal . ee-, c w, c.,ve, 15 3 4 Transverse 15 3 4 Weld Metal 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 and the other an isotope of Np 237 , are located in the one containing an isotope of U dosimeter block The double containment afforded by the dosimeter assembly prevents loss and contamination by the U 238 and Np 237 and their acH Mion products. Each 23 dosimeter block containa approximately 12 milligrams of U ' and 17 milligrams of Np237(Table 2 2) neld in a 3/a 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 238 and one Np 237 tube per block), and the space around the tube was 29

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MATERAL t+O TITLt 5PEC:FICATG4 HEO O ITEPt -

1 B.OCK CA990*4 STEEL 1 2 CAABON STEEL 2 2 COVER ALUP.*m;UM 4 3 SPACER 1

4 NEPTUNIUM 23I SEALEO CAP 3ULE ST APA E SS A

~ (O 250 OO u O 3 75 LG, STEEL 3

to 9

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filled with cadmium oxide. Af ter placement of this material, each hole was blocked with two 1/3s inch-thick aluminum spacer discs and an outer '/8-inch thick steel cover disc welded in place.

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 stael to prevent corrosion of specimen surfaces 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 tt un dye pene* rant tested as a final inspection procedere. Fabrication details and testing procedures are listed in Figure 2 4.

TABLE 2-2 OUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS lsotope Weight (mg) Compound Weight (mg)

Np 237 17 1 NpO 2

20 t 1 U 238 12.0 UO 3 g 14.25 2-11

4 v ,,

r L

GAE CAPSAJS SPACERS 7tteSE28 C0edP ACTS C04sPACTS CMAAPTS CMAAPTS 00449 4C78 60eAPACft CMAFTl CMARPT9_

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U ~ - - ~ ~ ~ A" u AW1 AW9) AM t3 AW10 AM10 AW7 AH7 A*4 AH4 LEGEND: AL - INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDtNAL)

AT -

INTERMEDIATE SHELL PLATE B8805 3 (TRANSVERSE)

AW WELD METAL AH - HEAT-AFFECTED-ZONE MATERIAL

't w  %,

SI APl?RTURE c ':n SI Aho Asallable ()n k A l>erture Card Cdkb CMAMT9 DetaatttR$ ftha1JS CMAMTS CNAMYS CNAWYS CHAMTS CMAWY3 C04AP ACTl CoefpACT1 ftutaal 4'8 # t(19 Alit Atto AL90 AT57 AL87 AfD4 Alle Aitt AL81 ATM AL78 Af'O WTF AM77 648 AL17 ATW ALat AT06 US A*t3 AL 83 Aft 0 AL80 Af7 AL77 Atle AT23 AT22 Af21 Af t7 W76 AM4 AL'8 AtN AL40 Afe4 ALS$ Aft! AL82 AT7 y i ** ATM AL76 AT14 Wt3 AH63 Attl A,75 AL75 AT72 ALM ATIB 4.1 AfgG A te Aff s %B3 AT14 WEJ Mt $47 Aita Af7f AL7r AT71 A671 AT' 4 Alte AT06 AL 65 At 5 AL&2 ATE Afi$ Afil Af t? AT14 W61 AHll AL13 AT73 AL73 AT70 4 70 Aft? g 4 87 Alto AL64 Afli g AL81 Aft) was AH45 ' ALi2 Af 00 lAL90 Af61 ALSF ATM AL&4 AT%t AL61 Af40 Alet Af t2 W47 AH47 bet Alit ATLC AL$t ATM ALSe ATL3 ALA3 AT50 ALno Af47 8147 Afil Atil AT14 Afi3 Af t1

. Wit AH44 ALIG A150 ALM af55 AL55 Afh! ALU A149 AL 44 Af48 AL 44 Af10 W33 AH11 ALS Af45 AL46 Af42 AL42 AT30 AL33 ATM AL38 AT33 W3 Afg if(12 AH12 646 ALS A144 AL44 AT41 AL41 Afl0 AL M AT36 AL,s6 A T32 AL32 Af t) AT11 AT10 Att Att m31 AH35 AL7 Af43 ALA3 Af40 AL40 AT37 AL37 Afle AL;d AT31 AL31 AT7 Mfte AH16 ALS AT30 AL30 Aft? AL27 A724 Al24 AT21 AL21 Alig Ag *g Afg

.W17 AHt7 $44 AL$ Af29 A&l9 A*26 AL44 Af23 AL23 AT20 AL20 Aft? AL17 AT4 ATF Att AT$ ATS Mi$ AH18 AL4 Af26 AL20 AT2$ AL25 AT22 AL22 Atil AL19 AT14 AL10 Af4 M3 AH3 AL3 AT'% AL1$ A112 AL12 ATS Att Att 464 AT3 AL 3 AT3

'W3 AH2 543 AL2 AT14 Alte Af t1 At ti AT8 Aos ATS ALS AT) 4,2 414 ATS aft 471 AT2 int AH1 AL S Afi3 Ai13 Afi0 AL10 Af? AL7 AT4 AL4 At t 4t t att Figure 2-6. Specimen Location in the Alvin W. Vogtit. Unit No.1 Reactoi Surveillance Test Capsules 2-13/2 14 l

\ . - . . .. - - - - - - - - -

ryg g0 2 73 - 6 7 -

~

a

SECTION 3 PREIRRADIATION TESTING 3 1. CHARPY V NOTCH TESTS Charpy V-notch impact tests were performed according to ASTM E23 with specimens from ,he vessellower shell plate B86281. Specimens of both longitudinal and transverse orientations were tested at various test temperatures in the range from - 62*C to 149*C

(-80*F to 300*F). yeilding a full Charpy V notch transition temperature curve in both orientatior.s (Tabies 31 and 3 2 and Figures 31 and 3 2). Tests were also performed on the weld metal and HAZ metal at various temperatures from -118'C to 149'C

(- 180*F to 300*F) and are shown in Tab les 3-3 and 3 4 and Figures 3 3 and 3-4. .

Charpy V notch impact tests by Westinghouse on the surveillance plate B86281 in the transverse direction resulted in an upper shelf energy of 70 ft..lbs. as shown in Figure 3-2. Material qualifacation tests performed by Ocmbustion Engineering Inc., per Section 3 of the ASME Boiler and Pressure Vessel Code resulted in a somewhat higher shelf energy (85 ft. Ib.) and therefore the plate is in compliance with the 10 CFR 50 Appendix G requirement "that reactor vessel bel!!ine tr.aterials have a Charpy upper shell energy of no less than 75 ft. Ib. initially." The difference in upper shelf energy between the two tests is attributed to 1.) variation in sulfur content due 'o segregation during ingot solidifaca-tion which then can result in significant sulfur variation in the final rolled plate,2.) dif-ference in post weld stress relief time which can produce significant difference in toughness A summary of the Charpy V-notch impact tests results including upper sheh energy (bSE),

41 joule (30 ft Ib), 68 joule (50 ft Ib), and 35 mils (0.89mm) lateral expansion index_

temperatures are presented in Table 3-5.

The cpecimens were tested on a SONNTAG UNIVERSAL impact machine, Model Number SI 1 with a hammer energy capacity of 240 toot pounds and a striking velocity of 17.02 feet per second. The machine is calibrated every 6 months using Charpy V netch impact sper: mens of known energy values supplied by Watertown Arsenal. Specimen condi-tioning for high temperature testing is maintained using 6 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 Chrome! Aluminal thermal couples at high temperatures and by the use of Copper Constantan thermal couple,., at low temperature testing.

3-2. TENSILE TESTS Table 3-6 and Fiuures 3-5,3 6, and 3-7 show the results of tensile testa (per ASTM E8 and E-21 est criteria) from vessel lower sholl plate 88628-1 and from the weld metal.

Specimenu from plate B8628-1 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.

31

i r

A SATEC UNIVERSAL tensile testing machire Model 30WBN, was used with a SATEC 30,000 lb. load cell as an integral part of the testing macnine. The testing machine is calibrated daily and verified annually to the National Bureau of Standards. The gripping

• mechanism utilizes threaded adapters to pull rods attached to the cross headlload cell and frame. The recording devico utilizes a Hewlett Packard X-Y recorder and s Chart in console, serial number 7047A calibrated to a dual range high temperaturu exten- i someter, serial number DDRE 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 tosts conform to ASTM A370 77 (Mechanical Testing of Steel Products).

A typical stress strain curve is shown in Figure 3 8. For high temperati;re tests an Ap-plied Systems 3 zone type furnace was used with independent zone control. Temperatures were controlled by a Athena Controls, Type K T/,. Model Number 4000 T 302F 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 B86281 and the core region weld metal by dropweight tests (ASTM E 208) performed at Combustion Engineering, Inc. From this test data the RTNDT was calculated using the methods as described in Secti: r 1. The TNDT and RT NDT for lower shell plate B8628-t , weld metal are as follows:

Note TNDT and RTfJDT IO' an the Detthne sheil plaws is gwen in Appenda A Material TNDT ( F) RTNDT ( F)

Plate C8628-1 - 20'I l 50 Weld Metal (Longitudinal Seams - 10#I - 10 Weld Metal (Girth Seam) - 50kl - 30

a. Combustion Engineering Inc. Materials Certification Report

b. Combustion Eagineering Welding Material Qualification Test H 32255

c. Combustion Engineering Welding Material Qualification Test P 32255 32

j __ __

TABLE 3-1 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE 086281 (LONGITUDINAL ORIENTATION)

Temperature Impact Energy Lateral Expansion Shear

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

62 - 80 5.0 4.0 0.03 1.0 0 62 - 80 9.5 7.0 0.08 3.0 0

- 34 - 30 16.0 12.0 0.25 10.0 5

- 34 - 30 27.0 20.0 0.36 14.0 5

- 34 - 30 35.0 26.0 0.46 18.0 5

- 18 -

0 32.5 24.0 0.53 21.0 10

- 18 0 37.0 27.0 0.53 21.0 10

- 18 0 46.0 34.0 0.63 25.0 10 1 30 41.0 30.0 0.61 24.0 20 1 30 43.0 32.0 0.66 26.0 20 1 30 69.0 5 * .0 1.02 40.0 15 16 60 64.0 4 /.0 0.91 36.0 30 i 16 60 73.0 54.0 1.09 43.0 35 -

16 60 70.0 58.0 1.14 45.0 45 27 80 92.0 68.0 1.42 56.0 75 27 80 96.0 71.0 1.37 54.0 75 27 80 100.0 70.0 1.40 55.0 65 49 120 110.0 81.0 1.80 71.0 100 49 120 118.0 87.0 1.78 70.0 100 49 120 119.0 88.0 1.80 71.0 100 71 160 118.0 87.0 1.78 70.0 100 71 160 125.0 92.0 1.93 76.0 100 116 240 122.0 90.0 1.96 77.0 100 116 240 122.0 90.0 1.88 74.0 100 116 240 130.0 96.0 1.75 69.0 100 149 300 119.0 88.0 1.88 74.0 100 149 300 122.0 90.0 1.P8 74.0 100

~

3-3

TABLE 3 2 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE 888281 (TRANSVERSE ORIENTATION) lateral Expansion Shear Temperature impact Energy (f t Ib) (mm) (mils) (%)

('C) (* F) (J) 0.05 2.0 0 62 - 80 7.0 5.0 5.0 0.08 3.0 0 62 - 80 7.0 0.36 14.0 5

- 30 22.0 16.0

- 34 14.0 5

- 30 24.5 18.0 0.36

- 34 18.0 10 0 26.0 19.0 0.46

- 18 20.0 10 0 30.0 22.0 0.51

- 18 20.0 10 0 32.5 24.0 0.51

- 18 21.0 25 30 28.5 21.0 0.53

- 1 10 37.0 27.0 0.66 26.0

- 1 30 10 45.0 33.0 .79 31.0

- 1 30 35 51.5 38.0 .89 35.0 16 60 40 51.5 38.0 .97 38.0 16 60 35 56.0 41.0 6 .97 38.0 16 60 80 76.0 56.0 1.32 52.0 38 100 95 87.0 64.0 1.42 56.0 38 100 95 99.0 73.0 1.63 64.0 38 100 95 89.5 66.0 1.52 60.0 49 120 100 100.0 74.0 1.73 68.0 49 120 100 107.0 79.0 1.80 71.0 49 120 100 91.0 67.0 1.47 58.0 71 160 100 92.0 68.0 1.63 64.0 71 160 100 92.0 68.0 1.63 64.0 71 160 100 85.5 63.0 1.52 60.0 116 240 100 93.5 69.0 1.70 67.0 116 240 100 89.5 66.0 1.73 68.0 149 300 100 95.0 70.0 1.78 70.0 149 300 100 84.0 62.0 1.50 59.0 177 350 100 88.0 65.0 1,57 62.0 177 350 100 98.0 72.0 1.65 65.0 204 400 100 102.0 ~75.0 1.52 60.0 204 400 100 96.0 71.0 1.60 63.0 232 450 100 103.0 76.0 1.68 66.0 232 450 3-4

1 TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION l WELD METAL Temperature impact Energy Lateral Expansion Shear

('C) (* F) (J) (f t Ib) (mm) (mlis) (%)

- 73 - 100 7.0 5.0 0.03 1.0 0

- 73 - 100 9.5 7.0 0.08 3.0 0

- 73 - 100 9.5 7.0 0.08 3.0 0

- 51 - 60 15.0 11.0 0.23 9.0 5

- 51 - 60 15.0 11.0 0.18 7.0 5

- 51 - 60 37.0 27.0 0.56 22.0 10 l - 34 - 30 11.0 8.0 0.23 9.0 10 l - 34 - 30 12,0 9.0 0.25 10.0 15

- 30 27.0 20.0 0.41 16.0 10 1 - 29 - 20 20.0 15.0 0.28 11.0 15 l - 29 - 20 51.5 38.0 0.86 34.0 20

- 29 - 20 66.0 -49.0 0.99 39.0 25

- 18 0 47.5 35.0 0.76 30.0 25.

l - 18 0 60.0 44.0 0.97 38.0 20 l

- 18 0 77.0 57.0 1.14 45.0 35

- 1 30 73.0 54.0 1.17 46.0 35

- 1 30 88.0 65.0 1.45 57.0 55 1 30 102.0 75.0 1.57 62.0 65

27. 80 102.0. 75.0 1.63 64.0 80 27- 80 107.0 79.) 1.73 68.0 85 27 80 114.0 84.0 1.83 72.0 90 49 120 117.0 86.0 1.88 74.0 95 49 120 119.0 68.0 1.90 75.0 95 49 120 121.0 89.0 1.98 78.0 95 71- 160 122.0 90.0 2.06- 81.0 100 71 160 122.0 90.0 1,98 78,0 100 71 160 127.5 94.0 2.08 82.0 100 116 240 125.0 92.0 2.06 81.0 100 116 240 125.0 92.0 2.11 83.0 100 149 300 125.0 92.0 2.08 82.0 100 149 300 125.0 92.0 2.11 83.0 100 3-5 L ..

TABLE 3 4 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATEAIAL Temperature impact Energy Lateral Expansion Shear .

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

- 118 - 180 9.5 7.0 0.10 4.0 0

- 118 - 180 12.0 9.0 0.08 3.0 0

- 84 - 120 18.0 13.0 0.13 5.0 5

- 84 - 120 22.0 16.0 0.23 9.0 5

- 84 - 120- 26.0 19.0 0.28 11.0 5

- 73 - 100 19.0 14.0 0.13 5.0 5

- 73 - 100 41.0 30.0 0.46 18.0 5

- 73 - 100 47.5 35.0 0.56 22.0 10

- 62 - 80 38.0 28.0 0.46 18.0 10

- 62 - 80 58.0 43.0 .0.63 23.0 20

- 62 - 80 70.5 ,

C2.0 0.81 32.0 25

- 51 - 60 35.0 26.0 0.48 19.0 20

- 60 46,0 34.0 0.53 21.0 25

- 51 - 60 54.0 40.0 0.74 29.0 35

- 34 - 30 70.5 52.0 0.94 37.0 60

- 34 - 30 81.0 60.0 1.09 43.0 45

- 34 - 30 91.0 07.0 1.12 44.0 60

- 18 0 108.5 80.0 1.32 52.0 70

- 18 0 115.0 85.0 1.27 50.0 60

- 18 0 131.5 97.0 1,68 66.0 90

- 1 30 130.0 96.0 1.74 68.0 100 1 .30 134.0 99.0 1.68 66,0 100

- 1 30 148.0 109.0 1.85 73.0 100 27 80 130.0 96.0 1.83 72.0 100 27 80 138.0 102.0 1.65 65.0 100 27 80 154.5 114.0 1.83 72.0 100 49 120 136.0 100.0 1.63 64.0 100 49 120 138.0 102.0 1.63 64.0 100 49 165.0 122.0 1.98 78.0 100 120~ .'

71 160 130.0 96.0 1.85 73.0 100

-71 160 149.0 110.0 1.80 71.0 100 71 160 161.0 119.0 1.83 72.0 100 99 210 130.0 96.0 1.83 72.0 100 99 210 168.0 124.0 1.85 73.0 100 3-6

TABLE 3 5

SUMMARY

OF ALVIN W. VOGTLE UNIT NO. 2 I REACTOR PRESSURE VESSEL IMPACT TEST RESULTS FOR LOWER SHELL PLATE B86281 AND CORE REGION WELD AND HEAT AFFECTED ZONE MATERIAL Upper Shelf 41 J 60 J 0.89 mm Energy (30 ft Ib) (50 ft Ib) (35 mils)

Material (USE) Index Temp Index Temp Index Temp (J) (f t Ib) (* C) (*F) ('C) ('F) ('C) (*F)

Plate B86281 (Longitudinal 121. 89 - 12 - 10 -7 45 -2 35 Orientation)

Plate B86281 95.0 70 -1 30 24 75 4 40 (Transverse Orientation) .

Weld -5 125 92 - 26 - 15 -15 5 - 21 Heat Affected 144 106 - 62 - 80 - 43 - 45 - 47 - 50 e Zone ,

3-7 c __

. l .

s L

L i

TABLE 3-6 PREIRRADIATKH4 TENSILE PROPERTIES FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE '

VESSEL LOWER SHELL PLATE 80826-1 i ANO CORE REGON WELD METAL-t Reduction  !

0.2% Ultimate I Fracture Fracture Uneterm Totat in Test . Yseld Tennile Fracture Strength Er.4ongetson Entongation Aree Temperature Strength Strength ' Load Strees Material (%)' (%)

(kip) (N) (kai) (MPs) (ks') (MPa) 1%)  ;

'C - - *f (ksi) (MPs) (ksa J (MPa) g '

61.0 42 0 13 0 27 0 64 0 24 75 69 0 476.0' 90 0 621.0 30 13.344 0 170 0 1.172 0 Plate BB628-1 26 0 62 0 l 476.0 90.0 621 0 30. 13.344 0 165 0 1.138 0 62.0 4280 13 0 (Longitudinal 24 75 69 0 11 0 23 0 64 0  !

434.0 83 0 572 0 24 12.454 0 154 0 1.062 0 56 0- 3860 Onentation) 149 300. 63 0 23 0 60 0 l 428 0 83 0 572.0 28 12.454 0 1440 9930 $7.0 3930 12.0 149 300 62 0 22 0 60 0 l 87.0 600 0 30 t3.344 0 t49 0 1.027 0 63 0 414 0 12 0 288. 550 61 0 421 0 60 0 i 13.344.0 1550 1.069 0 62 0 4280 13 0 24 0

'288 550 61.0 421 0 870 600 0 30 i 938O 64 0 441 0 12 0 24 0 53 0 t Plate B8628.t ' 24 75 ' 69 0 4760 89 0 614 0 32 14.234 0 1360 60 0 4140 t2 0 25 0 63 0 15 69 0 4760' 90 0 621 0 30 13.344 0 160 0 ' t.103 0-

.(Transverse 24 22 0 55 0 428 0 83 0 572 0 30 13.344 0 129 0 8890 58 0 4000 tt 0 [

Orwraat on) 149 300 ts2 0 21 0 49 0 428 0 83 0 572.0 3.1 13.789 0 125 0 8620 63 0 4340 t t .0 149 300 62.0 22 0 51 0 l

607.0 .34 15.123 0 140.0 965 0 69 0 476 0 12 0 ,

! 288 '550 61.0 '421.0 .88 0 " 24 0 57 0 607.0 33 14.676 0 1540 1.062 0 67 0 462 0 13 0 l 288 550 -61.0 4210 88.0 l 56 0 386 0 13 0 25 0 64 0 t 67 0 4620 63 0 572 0 28 12.454 0 158 0 1.089 0 l '24 75 4070 13 0 25 0 66 0 428 0 84 0 5790 30 13.344 0 173 0 1.193 0 59 0 24 ' 75 62 0 10 0 21 0 64 0

{ .. 2.8 . 12.454 0 ' 155 0 1.0ti9 0 57 0 3930 [

77.0 531 0

. Weld Metal 149'. 300 61.0 42: O. 67 0 il 63 0 4340 79 0 545 0 26 11.565 0 1640 1.131 0 54 0 372 0 11 0 22 0 .;

149 300 22 0 63 0 4340 83 0 5720 30 13.344 0 158 0 1.089 0 59 0 407.0 tt 0 288 550 63 0

j. i.:10.0 59 0 l 407 0 t: 0 2i 0 s30 28s 550 63 0 434 0 83 0 572 0 36 53.344 0 i6i 0 l

1. ..

I '

l

!; i t-4 - - , , ; -- . J,.., .a . . . , ._ .-m-. ._ . _ , , _ . - , . .. .,.----m. , , - . ,- - . - .

TEMPERATURE (*C)

-100 -50 0 50 100 150 200 l l l l l l I --

180 120 - -

160 E

140 100 _.

0

-h -@ -

120 cr 80 ~

p' g

(2 -

100 -9 80 y 60 - -

W -

40 -

O/ - 40 20 -

b'b __ 20 0

b I I I I I O 200 100 0 100 200 300 400 TEMPERATURE (*F) ,

FIGURE 31. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE ALVIN VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE B86281 (LONG!TUDINAL ORIE.JTATION)

TEMPERATURE (*C) 100 50 0 50 100 150 200 250 l I l I I l l -

180 120 - -

160

~

'$ 100 -

g --

120

~

60 9 0 w

[ ag _

/0 2

_ 60 2

- 40 20 -

Q, _

20 o t / i j l 1 I I O

-200 -100 0 100 200 300 400 500 l eMPERATURE (*F)

FIGURE 3 2. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE B8628-1 (TRANSVERSE ORIENTATION) 3-9

f.

TEMPERATURE (*C) 100 50 0 50 100 150 200 l l l l l l l -

180 120 -

160 E 100 - -

140

} 80 - , g ~ ~

120

[ Of 2 _

300 g

$ 60 -

80

$ O m 40 - _ 60 g

2 p- - 40 20 -

, - 20 0 -

l I O 200 100 0 100 200 300 400 l EMPERATURE (*F)

FIGURE 3-3. PREIRRADIATION CHARPY V NOTCH IMPACT Eh ERGY FOR THE A8. VIN W. VOGTLE UNIT NO. 2 REACOR PRESSURE VESSEL CORE REGION WELD ME'i At TEMPERATURE (*C)

-150 it;D 50 0 50 100

150 I I I I i l I 140 -

200 180 120 -

O O

^ O O 160 O

fO 100 - C oQ d 8 O O 120 g 80 -

100 $

-b . 60 -

z O O 80 W

40 -

Oo/ -

60 20 - 'dg' -

40 hC -

20 I o --

I I I I I o

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

FIGURE 3-4. PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VEGSEL WELD HEAT-AFFECTED ZONE MATERIAL 3 10

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

" I I I I I I I

700 100 -

2 (SPECIMENS)

~

N 2

%- 600 2* 80 -

ULTIMATE TENSILE STRENGTH h,

$ 70 -

500 E y ,, _

' ~-- g  %

0 20e YlELD STRENGTH 00 50 -

l l l l l - 300 40 ___

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

TEMPERATUR8E (*C) 0 50 100 150 200 250 300 80 I I I I I I l 70 - 2-60 - -

o REDUCTION IN AREA g 40 -

0 2 30 1

8 9 g 20 -

,. 2 TOTAL ELONGATION 10 - b UNIFu . ELONG. . TION 0 I I I I I 0 100 200 300 400 500 600 TEMPERATURE (*F)

FIGURE 3 5. PREIRRADIATION TENSILE PROPERTIES FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE B8628-1 (LONGITUDINAL ORIENTATION)

E 3-11

l TEMPERATURE ('C) 0 5 100 150 200 250 300 n0 l I I I I I I 100 -

700 e 90 -

Q 2s NO- 600 80 -

$ g ULTIMATE TENSILE STRENGTH $

70 60

[N 2 V  %

~

400 2

0 2% YlELD STRENGTH 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 I I I I I I 70 -

60 -

y O O O 50 -

o O

{

p REDUCTION IN AREA J 40 -

E 30 -

20 -

C- b 2 2 TOTAL ELONGATION 10 -

O UNIFORM ELONGATION O I I I I I 0 100 200 300 400 500 600 TEMPERATURE (*F)

FIGURE 3 6. PREIRRADIATION TENSILE PROPERTIES FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SHELL PLATE B86281 (TRANSVERSE ORIENTATION) 3-12

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

00 90 -

600 5 80 -

N Q_

n M V LL ULTIMATE TENSILE STRENGTH 500 2

$ 70 -

2 60 -

400 0 20o YlELD STRENGTH 40  ! I I I I -

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

TEMPERATURE ('C) 0 50 100 150 200 250 300 80 i  ;  ;  ; i 70 -

2 b- (

^

60 -

REDUCTION IN AREA 50 -

40 -

0 2 30 -

/

$ d %-

b, 20 -

,2 TOTAL ELONGATION

]b-10 -

b UNIFORM ELONGATION

(

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

FIGURE 3 7. PREIRRADIATION TENSILE PROPF#ES FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOh TdESSURE VESSEL CORE REGION WELD METAL 3-13

g I U p# _

'i '

~e6

_ j I

l l

,y 1-in in W -

T a &

L in l l 1 l l l  !  !  ! _ ..

STR AIN Figure 3-8. Typical Stress-Strain Curve fcr Tensile Test

_. , ~ <

x

.i SECTION 4 POSTlRRADIATION TESTING

41. CAPSULE REMOVAL The fi st capsule (Capsule U) should be removed at the end of the first core cycle (1st refuelibj) as shown in Table 41. 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 pcstirradiation test facility for disassembly and testing of all the specimens.

- TABLE 4-1 SURVEILLANCE CAPSULE REMOVAL SCHEDULE

~ _.

Orientation Capsule of Lead Removal Expected Capsule identification: Capsules *II FactorIDI Time Fluence (n/cm2)

U 58.5* 4.00 1st Refueling 4.84 x 10 18

-Y 241

• 3.69- 5EFPY 1.64 - x 10"lCI V- 61- 3.69 9 EFPY 3.21 x 10"Idl-X 238.5* 4.00 15 EFPY 5.80 x 10" W 121.5* 4.00 Stand-By -

Z 301.5* 4.00 Stand By

a. Reference Irra 'iation Capsule Assembly Drawing, Figure 2-4.

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

Lc. Approximate Fluence at %iwall thickness at End-of-Life.

- dD Approximate Fluence at vessel > .ner-wall at End-of4ife.

4-1

a

4-2. CHARPY V-NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the lower shell plate B8628-1 weld metal, and HAZ metal in each capsule can be done singly at approximately ten diff erent 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 temperature. The test value of this tempet.ture should be r'1 pared with preitradiation test data. The test temperature for the rema.ning specimen :hould then be adjusted higher or lower so as to develop a complete transition curve. For succeeding tests after longei Ivadiation periods, the test temperature in each case should be chosen in the light of results from the previous capsule.

4.3 TENSILE TESTS A tensile test specimen from each cf 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 COhwACT SPECIMENS in light of current requirements of 10CFR, Part 50, Appendix G and applications of ASME Secition 111, Apendix G e.1d Section XI, Appendix A, the %-inch thick compact specimens should be tested in sus.1 a manner as to determine both static, crack initiation, and pro-pagation 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.M Specific test procedures should include unloading compliance and data interpietation shoulci utilize the Equivalent Energy and J-integral concepts 12ul 1 Witt. F J . 7Fra re Toughness Parameters Obtained from single smati specimen Tests" WCAP-9397. Octete 1978 2 Buchalet. C and Mager, T. R "Espenmental Ventication of Lower Bound K gValues Utah:tng the Equivalent Energy Concept sn Progress on Flaw Growth and Fracture Toughness Testing, ASTM-sTP-536. pp 281.;96 Amencar soaety for Testung ana Materials. Philadelphia,1973 3 Landes. J D and Begier. J. A, "Recent Developments in Jg Testing". in Developments in Fracturc Afechanics Test Afethods Standardstation, ASTM-sTP 632. pp 57-81. Amencan scomy Wr Testing and Materials. Philadelphia,1977-4 McCabe, D E ," Determination of R Curves for structurai Matenals Using Nonhnear Mechanics Methods," in Flaw Gmwth and Fracture,. ASTM-sTP 631, op 245226. Amencan Sooety for Teshng a-1 Matenals Phe!adett;h+a,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 temperatures. Test proceduros 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 postirradiatian testing operations is as follows:

E Milling machine or special cutoff wheel for opening capsules, dosimeter --

blocks and spacers.

E Hot cell tensile testing machine with pin-type adapter for testing tensiie specimens.

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

5 Hot cell Charpy impact testing machine.

5 Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeiers.

4-3

m APPENDIX A DESCRIPTION AND CHARACTERIZATION OF THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS i

Based on the initial RTNDT, chemical compostion (copper and phosphorus) and the end-of-life noutron fluence, the reactor vessellower shell plate B8628-1 is expected to have the highest end-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 beltine 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 Class 1 Steel plate produced by Lukens Steel Company.

-This steel _was used in the fabrication of the Alvin W. Vogtle Unit Nr 2 reactor pressure vessel,'specifically, from the 9%-inch lower shell plate B8628-1. Also supplied was a submerged arc weldment made from sections of lower shell plate B8628-1 and adjacent lower shell plate B8825-1. This test we!dment was fabricated using % inch Mi! B-4 weld filler wire, heat number 870's5 and Linde 124 flux, tot number 1061 and is identical to that used by Combustion Engineering, Inc. in the Alvin W. Vogtle Unit No. 2 reactor vessel f abrication process specifically the closing girth seam between the intermediate and lowu sh_ ell plates.

The chemical analyses. TNDT. RTNDT, upper shelf energy and heat treatment history

-of all the core region pressure vessel shell plates med in the fabrication of.the Alvin LW. Vogtle Unit NoJ2 reactor pressure vessel are summarized in Tables A 1 thru A-6 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. 2 surveillance program. Weld material identical to that used in the fabrice: ion of all the core region beltline weldsl I has been

. correlated with the Westinghouse surveillance program test weldment "D"_ test results -

'and available Combustion Engineering, incaveld certification reports and their surveillance program test weldment "C" This data is also reported in Table: A-3 thru A-6 of this Appondix.

a .- The beltline' welds are considered to include the intermediate and lower shell plate-longitudinal seams and the closing intermediate to lower shell girth seam.

A-1

TABLE A 1 CHEMICAL ANALYSIS OF THE INTERMEDIATE SHELL PLATES USED IN THE CORE REGION OF THE ALVIN W. VOGTLE UNIT- NO. 2 REACTOR PRESSURE VESSEL Chemical Compositon l *l Element Plate Plate Plate R 4-1 R-4 2 R 4-3 _

=C .21 .20 .25 Mn . 1.28 1.25 1.37 P .009 .009 .009

-S .010 .009 .012 Si- .23 .22 .20 Ni ,64 - .62 .59

Mo

.57 .55 .56 Cr .03 - .03 .03 C u '- .06 .05 .05

'Al .026 .024 .017 Co- .012 .011 .008 Pb - notoeicerco not meiected note-e-teo W .. <.01 <. 01 <.01 Ti <.01- <.01 <.01 Zr <.001 <.001 <.001 '

V -- '005 .004 .005

Sn .005 .005 004

.As .007 .007 .008 Cb < 01-. <.01 <.01 C Ng .009 .008 .008 B -- < 001 . < .001 < '.001

- a. Chemical Analysis by Combustion Engineering, Inc.

3 A-2

TABLE A-2 CHEMICAL ANALYSIS OF THE LOWER SHELL PLATES USED IN THE CORE REGION OF THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL Chemical Compositon (weight %)

Element '

Plate lbl Plate I I 88825-1 R-8-1 08628-1 'l IDI B8628-1 "I ICI

.23 .21 .24 .23 C

Mn 1.31 1.34 1.34 1.30 P .006 .007 .007 .007 S .014 .012 .016 .014

.25 .25 .25 .23 Si Ni 59 .62 .% .59

.62 .59 ,50 Mo .59 Cr .02 .02 .02 .07 Cu .05 .06 .05 .05 Al .031 .019 .029 .034 Co .004 .011 .004 .008 Po mi eiected not oeiectro mai det cteo < 07 _

W <.01 < 01 .

< 01 .

<.05 Ti <.01 <.01 <.01 .005 s Zr < 001 <.001 < 001 .

<.03 V 004 .004 .004 <.005 an .004 .004 017 007 As .006 008 .007 .008 Cb <.01 < 01 .

.01 <.05 N, .009 .011 .008 .007 8 <.001 <.001 < 001 . .008

a. Surveillance program test plate.

b. Chemical Analysis by Combustion Engineering. Inc,

c. Chemical Analysis by Westinghouse.

A-3

TABLE A-3 CHEMICAL ANALYSIS OF THE WELD METAL USED FOR THE INTERMEDIATE AND LOWER SHELL PLATES LONGITUDINAL SEAMS FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL Chemical (weight Comp)ositon 0<o Element Wire Flux l Test Weld Sampleta)

(Weld Wife Heat Na 67005 Linde 0091 Flun. Lot No 0145, C .15 Mn 1.34 P .007 S .011 Si .13 Ni .13 Mo .55 Cr - - - - - -

l Cu .07 l Al -

Co --

Pb W - - -

Ti -- -

Zr --

V .005 Sn As Cb -

N2 -

B -

a. Chemical Analysis by Combustion Engineering. Inc.

A-4

i

TABLE A;4

-CHEMICAL ANALYSIS OF THE WELD METAL USED - . ,

" -- FOR THE INTERMEDIATE TO LOWER SHELL CLOSING GIRTH SEAM OF T

- ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL Chemical (weightComp)ositon 0o / r Element Wire Flux b- Surveillance -

' Test Weld Sample tal Weldment tw< i w.,. m.a no are Test Plate D tbl 1

Liende 124 Ibe lot No 196 )

C .075- .099 Mn- 1.27 1.25 P. .007 .008 S .010 .013 Si .50 43 Ni . .12 .17

.Mo .52 .47 -

Cr .07 .061 L 040 Cu- .06 Al .015 b

Co- - .002

-- F b

- - - <.01

- W: <.01 L

Ti -- <.001=

. Zr-

-- c 01 ,

V' .004c < 004 Sn - - .

< 001

As --- - - 003-Cb - - - - < 002 N7 002-8 ---: .009- .

' a. Chemical Analysis.by Combustion Engineering. Inc.

b Chemical' Analysis by Westinghouse of the Surveillance Program test plate "D" representative of the closing girth.- ~

- seam Weld wire Heat No. 87005. Linde 124 Flux'

.L'ot No 1061.

A ,

5

.wy gr % e w - r --F*- ee

LTABLE A 5 TNDT,RTNDT AND UPPER SHELF ENERGY FOR THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR-PRESSURE VESSEL- CORE REGION SHELL PLATES AND WELD. METAL -

Upper bhelf talIcl INDT(a) [bj RTNDT Material Energy

(*C) (*F) (*C) (*F) (J) (ft Ib)

-Intermediate

-Shell Plates:

R-4 1 - 29 - 20 - 23 10 129 95

-R42 - 23 - 10 - 23 10 141 104 _

R-4 3 -18 0 -1 30 114 84 Lower

- Shell Plates:

B8825-1 - 29 - 20 4 40 113 83 R-8-1 - 29 - 20 4 -40 118 87

~ B0628-1, - 29 - 20 10 50 115 ,

85

a. Data obtained from Combustion Engineenng Inc. Reactor Vessel Material Certification Deports b Orop weigitt data obtained from the transverse matenal properties (normal to the i J. lor working 6tectior > -

c Fmm impact data obtained from the transverse material properties (normat to the major working directiom I"I' Upper Shelf ldI Material TNDT RTNDT Energy

(*C) (* F) - - (*C) (* F) (J) - (ft Ib)

~

Intermediate and Lower Shell_ Longitudinal Weld - 23 - 10 - 23 - 10 206 152 Seams-(Weld Wire Heat .-

No. 87005, Linde 0091:

- Lot No. 0145)

. Closing Girth Weld. Seam-

Joining the Intermediate. - 46 - 50 - 34 - 30 122 90

-to Lower Shell (Weld Wire Heat No. 87005, Lindc la1 '

Flux. Lot No.1061L ,

[ . .

d. Data obtamed from Combustion Engineenng. Inc Wire /Flun Weld Deposit Matenal Certification Tests.

A-6 l

'l TABLE A 6 HEAT TREATMENT- HISTORY OF THE ALVIN W. VOGTLE UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES AND WELD SEAMS Temperature Time l Cooling Material (* F) (br) 4 idi Water-quenched Austenitizing:

1600 i 25 Intermediate (871 *C)

!- Shell Plates Tempered: 4 II Air-cooled R-4 1 1225 25 R-4 2 (663*C)

R-4 3 Stress Relief: 16.53I Furnace-cooled 1150 50 (621 *C)

~

Austenitizing: 4 I' Water-quenched l 16c0 t 25 l Lower (871 *C) I^I Shell Plates Tempered: 4 Air-cooled 88825 1 1225 25 R-8-1 (663*C) 88628-1 Stress Relief: 12.0W Furnace-cooled l

1 ; 50 1 50 (621 *C)

Intermediate Shell Longitudinal Stress Relief: 16.5M Furnace-cooled Seam Welds 1150 50 (621 C)

Lower Shell Longitudinal .

12.0lbl Furnace-cooled Seam Welds ,

Local

-Intermediate to Strass rem 5.0 Furnace-cooled Lower Shell Girth 1153 1 50 Seam Weld (621 *C)

Surveillance Program Test Material Surveillance Program Weldment Test Post We'd

",)r e y, Stress Relief: 6.0 I 'I Furnace-cooled cm.ny com seana -1150- 50 (621 *C) a Laens Steel Company. Combustior Engineenng Inc. Certificanon Reports b Stress Rehef includes the intermediate to Lower Shell Clusing Girth Seam Post Weld Heat Treatment.

c The Stress Rehet Heat Treatment received by the Surveillance Test Weldrrent has oeen simulated.

A-7