ML20217A670

DPC Catawba,Unit 2,Reactor Vessel Radiation Surveillance Program

ML20217A670
Person / Time
Site:	Catawba
Issue date:	11/30/1985
From:	Meyer T, Singer L, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20210U635	List: ML20210U632 ML20210U640 ML20217A662 ML20217A670
References
WCAP-10868, NUDOCS 9709220008
Download: ML20217A670 (46)

Text

{{#Wiki_filter:Attachment 7 Westinghouse Electric Corporation Topical Report WCAP-10868, Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program. 9709220008 970915 PDR ADOCK 05000413L P PDR

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~ p,y DUKE POWER COMPANY CATAWBA UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Db ch I DOCUMENE CONTROL' DATE L. R. Singer F E B 2 4 9 8 6.' DUKE POWER COMRARE DESIGN ENGINEERIME: 7 4 hW APPROVED: T. A. Meyer, Mandger Structural Materials and Reliability Technology tI Work Perfouned Under DDPJ-106 E w E.d4 l MAR 1 81986 $$$' b'Y ' S WESTINGHOUSE ELECTRIC CORPORATION '~ a-~~~ ~ Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 CNM 1201 r0 0-0110 - 00tigis

PREFACE This report has been technically reviewed and checked by S. E. Yanichko of Structural Materials and Reliability Technology. Mk S. . Yanichko Date: November 5,1985. D 4 ili

ABSTRACT } A pressure vessel steel surveillance program per ASTM E 18582 has been developed for the Duke Power Company, Catawba Unit No.2 to obtain information on the effects of radia-tion on reactor pressure vessel material under operating conditions. The radiation surveillance program for the Catawba Unit No. 2 is designed to, and in compliance with, federal govem-ment regulations identified in appendix H to 10CFR, part 50 entitled " Reactor Vessel Material Surveillance Program Requirements." Following is a description of the prey.m, a description of the material involved, the specimen and capsule design and fabrication, and the preirradiation test results. / 9 V -n

TABLE OF CONTENTS Section - Title Page 1 PURPOSE AND SCOPE 11 2 CAPSULE PREPARATION 21 2 1. Pressure Vessel Material 2-1 2 2. Machining 21 2 3. Charpy V notch Impact Specimens 2-3 2 4. Tensile Specimens 23 2 5. 1/2T Compact Specimens 23 26. Dosimeters 23 2.7. Thermal Monitors 23 2.8. Capsule Loading 29 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 41. Capsule Removal 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 4-5. Postirradiation Test Equipment 4-3 Appendix A - DESCRIPTION AND CHARACTERIZATION,OF THE CATAWBA UNIT NO 2 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS A1 1 vil

LIST OF ILLUSTRATIONS Figure Title Page 11 Location of the Irradiation Test Capsules in the Catawba Unit No. 2 Reactor Vessel 1-4 2 Charpy V-notch Impact Specimen 22 2-2 Tensile Specimen 2-4 2-3 Compact Specimen 2-5 2-4 Irradiation Capsule Assembly 2 7/2-8 2-5 Daimeter Block Assembly 2 10 24 Specimen Locations in the Catawba Unit No. 2 Reactor Surveillance Test Capsules 2 13/2-14 31 Proirradiation Charpy V-notch Impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B8605-1 (Longitudinal Orientation) 3-9 3-2 Prairradiation Charpy V-notch impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel intermediate Shell Plate B86051 (Transverse Orientation) 3-9 3-3 Preirradiation Charpy V-notch Impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Metal 3-10 S4 Proirradiation Charpy V-notch Impact Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Heat-Affected Zone Material 3 10 35 Preitradiation Tensile Re;+I:: for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B86051 (Longitudinal Orientation) 3 11 e 36 Proirradiation Tensile M9;+I:: for the Catawba Unit No. 2 Reactor Pressure Vessel intermediate Shell Plate B8605-1 (Transverse Orientation) 3-12 S7 Preirradiation Tenelle Proporties for the Catawba Unit No. 2 Rehetor Pressure Vessel Core Region Wold Metal 3-13 38 Typical Stress-Strain Curve for Tensile Test 3-14 lx

- LIST OF TABLES Table Title Page 21 Type and Number of Specimens in the Catawba Unit No. 2 Surveillance Test Capsules 29 22 Quantity of isotopes Contained in the Dosimeter Blocks 2-11 31 Preirradiation Charpy V-notch impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B8605-1 (Longitudinal Orientation) 3-3 32 Preirradiation Charpy V-notch impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B8605-1 (Transverse Orientation) 34 33 Preirradiation Charpy V-notch impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel 35 Core Region Wold Metal. 3-4 Preirradiation Charpy V-notch impact Data for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Wold Heat-Affected-Zone Material 3-6 35 Summary of the Catawba Unit No. 2 Reactor Pressure Vessel Impact Test Results for intermediate Shell Plate B86051 and Core Region Weld and Heat Affected-Zone Material 3-7 3-6 Preirradiation Tensile Properties for the Catawba Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate B86051 and Core Region Weld Metal 3-8 4 Surveillance Capsule Removal Schedule 4-1 A1 Chemical Analysis of the Intermediate Shell Plates used in the Core Region of the Catawba Unit No. 2 Reactor Pressure Vessel A-2 A-2 Chemical Analysis of the Lower Shell Plates used in the - Core Region o; the Catawba Unit No. 2 Reactor Pressure Vessel A-3 F A-3 Chemical Analysis of the Wold Metal used in the Core Region Weld Seams of the Catawba . Unit No. 2 Reactor Pressure Vessel A-4 A4 TNDT,RTNDT and Upper Shelf Energy for the Catawba Unit No. 2 Reactor Pressure Vessel Core Region Shell Plates and Wold Metal A-5 A5 Heat Treatment History of the Catawba Unit No. 2 Reac-tor Pressure Vessel Core Region Shell Plates and Weld Seams A-6 XI e 4- ,,-,n.-.

SECTION 1 PURPOSE AND SCOPE The purpose of this program is to monitor radiation effects under actual operating con-ditk>ns of the core region reactor vessel materials in the Duke Power Company, Catawba Unit No. 2, a four loop, nuclear power plant with a thermal output rating of 3427-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 NDT s determined i utilize the reference nil-ductility temperature, RTNOT, as a basis. RT from the dropweight nil-ductility transition temperature (TNDT) pdr ASTM E208 and the weak 'l direction 50 ft Ib Charpy V-notch temperature (or the 35-mil lateral expan-l NDT s defined as the dropweight TNDT or the sion temperature if it is greater) RT i temperature 60*F less than the 50 ft Ib (or 35-mil) Charpy V-notch temperature, whichever is greater. Therefore RTNDT = TNDT, if TNDT 4T50(35) - 60*F and RTNDT = T50(35) - 60*F, if T50(35) - 60*F > TNDT RNDT = Reference nil-ductility temperature TNDT = Nil-ductility transition temperature per ASTM E208 T50(35) = 50 ft Ib temperature from Charpy V-notch specimens oriented in the weak direction (or the 35-mi temperature if it is greater) i

1. Longitudinal axis of the specimen oriented normel to the mapr working direction of the plate.

i 1-1

An emporical relationship between RTNDT and fracture toughness for reactor vessel i stools has been developed in Appendix G. " Protection Against NorHiuctile Failure," to Section 111 of the ASME Boiler and Pressure Vessel Code. This relationship can be I 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 Ii temperatures, IW and thus cause the RTNDT o increase with radiation exposure. The t extent of the shift in the impact energy curve, that is, radiation embrittlement, is enhanced i by certain chemical olomonts (such as copper) present in reactor vessel stools,13I The adjustment in RTNDT with service can be monitored by a surveillance program involving periodic checking of 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 Iri addition to Charpy V-notch specimens to evaluate j 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-I mine the shift of the refenmco temperature *I with radiation exposure at plant temperatures,

a. The reference temperature as defined by 10CFR Part 50, Appendix G, Section Il E i

is as follows: " Adjusted reference temperature" means the reference temperature as adjusted for irradiation effects by adding to RTNOT the temperature shift, measured at the 30 ft Ib (41 J) level, 1. Poner, L. F., "Radisson Effects in Steel," in Atefoneis in Nuedeer Am*=**es. ASTM-STP 276, pp.147196, American , Soc 6esy for Tesang eral Motortels, PNiedelpNa,1980. 2. Steele, L E. and Hawtheme, J. R., "New intermaton on Neutron EnNetlement o' d Emtwtutement Renet of Reactor Pressure Vesosi Steels," NRL 4100. August 1984. 3. Potapows, U. and Hoseiome, J. R., "The Elesce of Residual Elements on 580*F treadiation Respones of Selected Pressure Vessel Stoets and Wetements," NRI 0003, Septemmer 1988. 4. Steele, L E.," Structure and Compoemlon Eftects on irrediamon SenenMey of Pressure Vesest Stests," in irradission spects on Stuceuref AAlope ter Nucieer Reactor %. ASTM-STP-444, pp.164175, Amortcon Society for Testing and Motortels, a PI N 1970.- 3 12 --+2--- ,+vrm-w ww y ,,-cr-aw-,- .---e.----

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 temperaturs. The postirradiation test results of the compact specimens will provide actual fracture toughness properties of the vessel material. These propeities may be used to establish allowable stress intensity factors for subsequent analyses. Six material test capsules are fabric;ted containing specimens from the reactor vessel shell plate identified as being most likely to limit the operation of the reactor vessel. The specirnens contained in the Catawba Unit No. 2 test capsules are from the in-termediate 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 hebt treatment received by the specimens has been simulated (Appendix A). The six materlal 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 1-1. b e 4 1-3 4

L O' REACTOR VESSEL s v CORE BARREL NEUTRON PAD 4 (301.5') Z CAPSULE U (58.5') )" V (81') ^ M"58.5 ' L 58.5- ~ 81-g a{, 270' 90* l (241 *) Y l

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l SECTION 2 CAPSULE PREPARATION 2-1. PRESSURE VESSEL MATERIAL P* actor vessel material was supplied by Combustion Engineering, Inc. from Interme-3 shell plate B8605-1, Heat No. C05431. Combustion Engineering, Inc., also

led a weldment which joined sections of material of the intermediate shell sie B8605-2 (See Note) and the adjacent lower shell plate B88061, Heat No. C2288-1.

Data on the limiting core region plate (B86051), weld, and weld heat affected zone material are provided in Appendix A. Note: The limiting material for the Catawba Unit No. 2 reactor vessel beltline region is intermediate shell plato B8605-1. This is based on the highest ARTNDT shift (94*F) as calculated using the latest ASTM revisions. The original material selected in 1978 was intermediate shell plate B8605-2. This selection was based at the time on the highest initial RTNDT. Therefore weld test plate "D" furnished to Westinghouse at that time was made up of plates B8605 2 and B8806-1. 2-2. MACHINING Test material obtained 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 b the test material and also from weld and heat-affected-zone metal of a stress-relieved weldment joining intermediate shell plate B8605-2 and adjacent lower shell plate B88061. All heat-affected-zone specimens were obtained from the weld heat affected-zone of intermediate shell plate B8605-2. 2-1 4

f 44' 44' 3 ) [ '{ O.0llR 0.009 r f .,,, - 0.39s +-- 0.393 ggo ye. f y ~ ~ ~ - \\,,] 0.396 0.393 u d

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2.3 Charpy V-notch impact Specimens -Charpy V-notch impact specimens corresponding to ASTM A370 Type A (Figure 21) were machined from intermediate shell plate B86051 in both the longitudinal orientation (longitudinal axis of specirnen 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 dimon-sion 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. 2-4. Tensile Specimens Tensile specimens (Figure 2 2) from shell plate B8605-1 were machined in both the longitudinal and transverse orientation. Tensile specimens from the weld were oriented normal to the welding direction. 2-5. 1/2T Compact Specimens Compact test specimens (Figure 2-3) from shell plate B86051 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 procracked 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 7 23s 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 monitors are sealed in Pyrex tubes and then inserted in spacers located as shown in i 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 bint: 310*C (590*F) 2-3 2

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5 2-8. CAPSULE LOADING o 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 in each capsule is shown in Table 21. TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE CATAWBA UNIT NO. 2 SURVEILLANCE TEST CAPSULES Capsules U, V, W, X, Y, and Z Material Charpy Tensile Compact Plate B8605-1 15 Longitudinal ce r.ni ce 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, 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 loss and contamination by the U" and Np* and their activetion products.- Each dosimeter block contains approximately 12 milligrams of U" and 17 milligrams of L Np*(Table 2 2) held in a 8/rinch-long by % inch outside diameter sealed stainless steel l tube, respectively. Each tube was placed in a %-inch-diameter hole in the dosimeter l block (one U* and one Np" tube per block), and the space around the tube was 2-9 l. l. L ~

E { I i l ( I 3 I MATERIAL NO ITEM TITLE SPECFICATION REOV l t SLOCK CARBON STEEL t 2 2 COVER CARBON STEEL 2 3 SPACER ALUMINUM 4 A 4 DN ST m t 0*06 - pa 3 (0 250 00 a 0 375 LG) STEEL b 5 N238 SEALED CAPSULE STAINLESS 1 0 , r (0 250 00 m O 3T5 LG) STEEL 6 CApMIUM OX10E AS REOV l I r-- - -y V 6 i i Figure 2-5. Dosimeter Block Assembly i ~

filled with cadmium oxide. After placement of this material, each hole was blocked with two 1/is-inch-thick aluminum spacer discs and an outer %-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 st,ainless steel to prevent corrosion of specimen surfaces during irradiation. The capsules are hydro-statically compressed in domineralized 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 details 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) Np 17 1 NpO 20 i 1 237 2 238 M.3 U 12.0 U 0, 3 i s 2-11

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n. u a

au u a on R _am ~i ,e o.: mi. as em am n a cli ui mi. a n. m mn as m. _ = _ _== ma, am en am ma an m on m m m = mm m. o.: mi. n. n, an era an en;== n a. mi. au m. ',~ ~ o.: a ma mis au ma an on u m u.4 on ou m >0 o.3 m. tai ou ofie ai. ofH aii oil na m mA oTI ott ote oil off oli oft anl o.i o,n mi ora m. ai m a, m. u. mi ai mi Figure 2-6. Specimen Location in the ANSTEC Catawba Unit No. 2 APERTURE Reactor Surveillance Test Capsules CARD 213/214 Also Availablo on Aperture Card qqon 0006-0

i 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 B86051. Specimens of both longitudinal and transverse orientations were tested at various test temperatures in the range from - 62'C to 160*C (- 80'F to 320*F), yleiding a full Charpy V notch transition temperature curve in both orientations (Tables 31 and 3 2 and Figures 31 and 3 2). Tests were also performed on the weld metal and HAZ metal 3 various temperatures fron -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), I 41 joule (30 ft Ib), 68 Juule (50 ft Ib), and 35 mils (0.89mm) lateral expansion index temperatures are presented in Table 3 5. The specimens were tested on a Sontag Universel Model Number Sl 1 Impact machine with a hammer energy capacity of 240 foot pounds and a striking veloci.y of 17 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 ISO Temperature Oven, Model 350. For low temperature specimen conditioning either liquid nitrogen or dry ice in isopropanol is used. The specimen temperatures are monitored by use of a "J" type thermocouple or a thermometer. 3-2. 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 B8605-1 and from the weld metal. Specimens from plate B8605-1 and the weldment were tested at 24'C (75'F), 149'O (300'F) and 288'C (550'F) in both the longitudinal and transverse directions. 31

. ~.. t An Instron Universal tensile testing machine Model TTD (20K 50K) was used with an instron load cell (Serial number 0598N and 044SN) which is calibrated daily and verified annually to the National Bureau of Standards. The gripping mechanism utilizes thread. ed adapters to pull rods attached to the cross headnoad col 8 and frame. The recording device utilizes an instron Model 3124 strip chart in consolo calibrated to the Instron Class B 1 extensometer, Model 4929. 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 A37077 (Mechanical Testing of Steel Products). A typical strose strain curve is shown in Figure 38. 33. DROPWElGHT TESTS The nil ductility transition temperature (TNDT) was determined for plate B88051 and the oore region weld metal and heat effected zone by dropweight tests (ASTM E 208) performed at Combustion Engineering, Inc. From this test data the RTNDTwas calculated NDT or Intermediate f using the methods as described in Section 1. The TNDT and RT shell plate B88051, well metal and heat affected zone (HAZ) are as follows: NOT or oN the boothne shen plates le given in Appendia A. f Note: TNOT and RT i Material TNDT (*F) RTNDT ('F) l Plate B86051 - 1 0 *3 + 15 WeiM Metal oniormedme one to-or shen - 80 *l - 80 l L am s.ame one on.no owei scom> l HAZ - 80*3 - 80

s. Combustion Engineering Materials Certification Report,

b. Combustion Engineering Welding Material Qualification Test.

.c. Combustion Engineering Surveillance Wold Test Plate "C" Materials Test Report i 4 6 32 i w .-.-----e---m ,-+w.,,-v.-.r--- - +,..,, _e y,.g. ..wm..- ,ver,p.,m,m._m,.,r-._._ww,_,m.7--,,._,y i-,,-,--w-,-,- y v, v.-

I i i i I 1 I t t. I i J l TABLE 31 l PREIRMADIATION CHARPY V440TCH IMPACT DATA l POR THE CATAWBA UNIT NO. 2 REACTOR j l PRESSURE VESSEL INTERMEDIATE SHELL i i PLATE 988051 (LONGITUDINAL ORIENTATION) 4 1 1 ) Temperature impost Energy Lateral Expansion Sheer l t ('C) ('P) (J) (ft Ib) (mm) (mile) (%) j ) - 62 - 80 5.5 4.0 0.03 1.0 3 4 - 62 - 80 8.0 8.0 0.08 3.0 3 l - 40 - 40 18.0 13.0 0.20 8.0 9 - 40 - 40 26.0 19.0 0.36 14.0 9 40 - 40 37.0 27.0 0.51 20.0 14 l

q

- 18 0 43.0 32.0 0.46 ift.0 25 - 18 0 51.5 38.0 0.79 31.0 30 - 18 0 57.0 42.0 0.69 27.0 25 i 7 20 e9.0 51.0 1.12 44.0 34 l L 7 20 69.0 51.0 0.97 38.0 29 l - -7 20 95.0 70.0 1.22 48.0 23 4 40 91.0 67.0 1.09 43.0 40 1 4 40 111.0 82.0 1.45 57.0 48 i 4 40 111.0 82.0 1.52 60.0 45 l 27 80 104.0 77.0 1.35 53.0 56 l 27 80 117.0 86.0 1.45 57.0 50 l 3 27 80 123.0 91.0 1.52 60.0 54 i-49 120 115.0 85.0 2.01 79.0 81 49 120 118.0 87.0 1.80 71.0 75 49 120 142.0 112.0 1.83 72.0 77 66 150 159.0 117.0 2.01 79.0 94. 66 150 182.0. 134.0 2.13 84.0 -100 4 82-180__ _180.0 133.0 2.13 84.0 100 I 82 190 183.0 135.0 2.24 . 88.0 100 o i 82 180 192.5 142.0 2.24 88.0 100 i \\'

116 240 183.0 135.0 2.24 86.0 100 116-240 188.5 130.0 2.18 86.0 100 i

2 180. -320 188.5 139.0 2.16 85.0 100 160 320 197.0 145.0 2.18 86.0 100 f

TABLE 3 2 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 586051 (TRANSVERSE OR!ENTATION) Temperature impact Energy Lateral Expansion Shear ('C) ('F) (J) (ft Ib) (mm) (mils) (%) - 62 - 80 7.0 5.0 0.05 2.0 3 - 62 - 80 8.0 6.0 0.13 5.0 3 - 40 - 40 12.0 9.0 0.15 6.0 13 - 40 - 40 22.0 16.0 0.25 10.0 18 - 40 - 40 24.5 18.0 0.33 13.0 9 - 18 0 46.0 34.0 0.53 21.0 29 - 18 0 47.5 35.0 0.69 27.0 25 - 18 0 56.0 41.0 0.89 35.0 29 4 40 47.5 35.0 0.69 27.0 38 4 40 70.5 52.0 1.04 41.0 43 4 40 73.0 54.0 1.02 40.0 34 27 80 80.0 59.0 1.19 47.0 41 27 80 87.0 64.0 1.22 48.0 44 27 80 96.0 71.0 1.27 50.0 45 38 100 72.0 53.0 1.09 43.0 47 38 100 89.5 66.0 1.27 50.0 55 38 100 104.5 77.0 1.37 54.0

• 59 49 120 89.5 66.0 1.19 47.0 62 49 120 114.0 84.0 1.55 61.0 68 49 120 127.5 94.0 1.78 70,0 100 82 180 125.0 92.0 1.63 64.0 100 t

82 180 130.0 96.0 1.83 72.0 100 82 180 134.0 99.0 1.80 71.0 100 116 240 136.0 100.0 1.80 71.0 100 116 240 138,0 102.0 1.98 78.0 100 .160 320 114.0 84.0 1.65 65.0 100 160 320 136.0 100.0 1.90 75.0 100 i r

TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT No. 2 REACTOR ~ PRESSURE VESSEL CORE REGION WELD METAL I Temperature impact Energy Lateral Expansion Shear ('C) (*F) (J) (ft Ib) (mm) (mils) (%) - 96 - 140 5.5 4.0 0.03 1.0 9 - 96 - 140 7.0 5.0 0.05 2.0 13 - 62 - 80 8.0 6.0 0.03 1.0 13 - 62 - 80 11.0 8.0 0.08 3.0 18 - 62 - 80 35.0 26.0 0.28 11.0 18 - 51 - 60 15.0 11.0 0.15 6.0 28 ) - 51 - 60 20.0 15.0 0.25 10.0 33 - 51 - 60 20.0 15.0 0.15 6.0 28 - 40 - 40 62.0 46.0 0.79 31.0 47 - 40 - 40 79.0 58.0 1.04 41.0 40 - 40 - 40 99.0 73.0 1.30 51.0 52 - 18 0 79.0 58.0 1.14 45.0 65 - 18 0 130.0 96.0 1.52 60.0 73 - 18 0 137.0 101.0 1.83 72.0 71 4 40 164.0 121.0 2.01 79.0 96 4 40 169.5 125.0 1.98 78.0 93 4 40 183.0 135.0 2.03 80.0 84 27 80 178.0 131.0 2.01 79.0 93 27 80 187.0 138.0 2.24 P8.0 96 27 80 199.0 147.0 2.18 86.0 94 49 120 192.5 142.0 2.24 88.0 100 49 120 198.0 146.0 2.24 88.0 100 49 120 205.0 151.0 2.21 87.0 100 104 220 188.5 139.0 2.18 86,0 100 104 220 201.0 148.0 2.31 ,91.0 100 160 320 206.0 152.0 2.29 90.0 100 6 160 320 222.0 164.0 2.18 86.0 100 35 . ~.. n._-.

TABLE 3-4 PREIRRADIATION CHAF;PY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATERIAL Temperature impact Energy Lateral Expannion Shear ('C) (*F) (J) (ft Ib) (mm) (mils) (%) - 118 - 180 9.5 7.0 0.08 3.0 3 - 118 - 180 12.0 9.0 0.05 2.0 3 - 118 - 180 15.0 11.0 0.05 2.0 9 - 84 - 120 20.0 15.0 0.05 2.0 10 - 84 - 120 31.0 23.0 0.18 7.0 29 - 84 - 120 38.0 28.0 0.25 10.0 25 - 62 - 80 12.0 9.0 0.13 5.0 29 - 62 - 80 38.0 28.0 0.36 14.0 29 - 82 - 80 84.0 62.0 0.74 29.0 43 - 51 - CO 37.0 27.0 0.43 17.0 32 1 - 51 - 60 79.0 58.0 0.86 34.0 50 - 51 - 60 90.0 66.0 0.86 34.0 47 - 40 - 40 72.0 53.0 0.79 31.0 56 - 40 - 40 108.5 80.0 1.19 47.0 59 - 40 - 40 127.5 94.0 1.37 54.0 68 - 18 0 138.0 102.0 1,55 61.0 73 - 18 0 142.0 105.0 1.42 56.0 90 - 18 0 183.0 135.0 1.78 70.0 100 4 40 167.0 123.0 2.06 81.0 100 4 40 186.0 138.0 1.98 78.0 100 4 40 201.0 148.0 2.03 80.0 100 27 80 163.0 120.0 1.78 70,0 100 27 80 174.0 128.0 1.80 71.0 100 27 80 197.0 145.0 1.90 75.0 100 , 60 140 184.5 136.0 2.08 82.0 '100 80 140 207.5 153.0 1.98 78.0 100 93 200 176.0 130.0 2.03 80.0 100 93 200 199.0 147.0 2.01 79,0 100 _ - -,.. ~. _ _.. -. _ _., _.. _ _. _ r

TABLE 3 5

SUMMARY

OF CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL lMPACT TEST RESULTS FOR INTERMEDIATE SHELL PLATE 586051 AND CORE REGION WELD AND HEAT-AFFECTED ZONE MATERIAL Upper Shelf 41-J 66-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 B86051 (Longitudinal 187 138 - 26 - 15 -7 20 -9 15 Orientation) Plate B86051 (Transverse 130 96 - 21 -5 4 40 4 40 Orientation) Weld 202 149 - 46 - 50 - 40 - 40 - 34 - 30 Heat Affected 184.5 136 - 71 - 95 - 54 - 65 - 43 - 45 Zon9 37 4 ~.

---

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• • Rama **Raam enRRam y

? M3!!RR 33!!ER 33!!RR i^l I iP al 1 111 i i

TEMP:RATURE (CC) 100 50 0 50 100 150 200 l I I I I I I m 200 h 8 M ~ 130 3 120 l-o 160 O 2 (Specimens) / 140 120 $ { 30 g 2 O a O f M g 40 20 20 l l l l l 0 o 200 100 0 100 200 300 400 TEMPERATURE (*F) FIGURE 31. PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT No. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 586051 (LONGITODINAL ORIENTATION) TEMPERATURE (*C) 100 50 0 50 100 150 200 120 160 1# A O 100 Of 120 m O a m 1 OO g m a 4 20 20 1 I I I I 0 o 200 100 0 100 200 300 400 TEMPERATURE (*F) I FIGURE 3 2. PRElRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT No. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE 586051 (TRANSVERSE ORIENTATION) S9

_ _.. - _.. ~_.- _-.- 1 TEMPERATURE (CC) 100 50 0 50 100 150 200 I I I I I I I m 240 180 O 220 100 0 O 200 ( o/8 'a 9 1e0 g 120 E 1* g 3oo i 120 80 100 O 80 40 ~ O 2 20 20 l l l I l 0 0 200 100 0 100 200 300 400 TEMPERATURE (*P) MOURE 3 3 PRERRADIATKW CHARPY V NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REOlON WELD METAL I TEMPERATURE (*C) 150 100 50 0 50 100 150 l I I I I I i 12e M ~ 1M O Oo 9 g 1" O O O 1" O O8 120 i0 8 ia 100 ms g im l. O = o 09 20 /O ~ O I O l l l l o o 300 200 100 0 100 200 300 TEMPERATURE (*F) MOURE 3 4. PRWRRADIATION CHARPY V440TCH IMPACT ENERGY FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL WELD HEAT AFFECTED ZONE MATERIAL l l l 3 10

TEMPERATURE (CC) 0 50 100 150 200 250 300 l l l l l l l 00 100 2 2 (SPECIMENS) so ULTIMATE TENS!LE STRENGTH { 500 3 E g%f 70 O 60 400 0.2% YlELD STRENGTH O 50 l l l l l 300 40 0 100 200 300 400 500 600 TEMPERATURE (*F) TEMPERATURE (*C) 1 80 l l l l l l l \\ b 70 g 2 REDUCTION IN AREA [ 50 U 40 E 9 8 30 2 9 O 20 2 TOTAL ELONGATION 9 ~ UNIFORM ELONGATION I I I I I O O 100 200 300 400 500 600 TEMPERATURE (*F) FIGURE 3 5. PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL !NTERMEDIATE SHELL PLATE B8605-1 (LONGITUDINAL ORIENTATION) 3-11

TEMP RATURE (CC) 0 50 100 150 200 250 300 l l l l l l l 700 100 2} g* -9 -g-Sm ULTIMATE TENSILE STRENGTH 2 g 70 2 2 g y m 60 400 0.2% YlELD STRENGTH 50 i I I I I 300 4o 0 100 200 300 400 500 600 TEMPERATURE (*F) TEMPERATURE (*C) 0 50 100 150 200 250 300 l l l l l l l l 70 8 9 o 30 s REDUCTION IN AREA ~ 50

l 40 o=

x g_ -9 _g o g2 20 TOTAL ELONGATION o e 10 UNIFORM ELONGATION I I I I I O O 100 200 300 400 500 600 TEMPERATURE (*F) FIGURE 3-6. PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UN6T NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE B8605-1 (TRANSVERSE ORIENTATION) 3 12

TEMPCRATURE (CC) 0 50 100 150 200 250 300 110 l I I I I I I ~ 100 g 9N 2\\ \\- ^ 90 sm 80 ULTIMATE TENSILE STRENGTH g g 70 E - Q -- 3 -9

1. 60 0.2% YlELD STRENGTH 400 50 I

I I I I 30 40 0 100 200 300 400 500 600 TEMPERATURE (*F) TEMPERATURE ('C) 0 50 100 150 200 250 300 F l I I I I I I 70 g ~ REDUCTION IN AREA b 50 40 d t; 30 g-8 -9 8 2 20 TOTAL ELONGATION 10 UNIFORM ELONGATON I I I I I O O 100 200 300 400 500 600 TEMPERATURE (*F) 1 FIGURE 3 7. PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3-13

2-. -u.2~.-.w.~.,-..x.,..__s .,ana.,-u._. ._a- ..~_-.___ _ __. ....-.a.-n-__--.~.n -~ ~ _. - -. - I 6 r 8 z a (C E \\ H $? g = U) h 4 o* I 4 i I I I I I I I i SS3W1S t 3 14 v -~.-- -r .,y__ ~~ s -_3

.p SECTION 4 POSTlRRADIATION TESTING 41. CAPSULE REMOVAL The first capsule (Capsule U) should be removed at the end of the first core cycle (1st refueling) as shown in Table 41. Subsequent capsules should be removed at 6,9, and 15 EFPY (Effective Full Power Years) as indicated. Each specimen capsule, removed after exposure, will be transferree to a postirradiation test facility for disassembly and testing of all the specimens. 1 TABLE 4-1 SURVEILLANCE CAPSULE REMOVAL SCHEDULE Orientation Capsule of Lead Removal Expected Capsule identification Capsules *3 Factor >l Time Fluence (n/cm ) l 2 U 58.5' 4.00 1st Refueling 3.28 x 10" Y 241

• 3.69 6EFPY 1.45 x 10'N'I V

61 3.69 9 EFPY 2.18 x 10'Mdl X 238.5' 4.00 15 EFPY 3.94 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

\\ 2 [ j q 4 2. CHARPY V NOTCH IMPACT TESTS - I The testing of the Charpy impact specimens from the intermediate chell plate 888051 g weld metal, and HAZ metal in 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 crimplete Charpy impact energy transition curve. The initial Charpy specircon from the first capsule removed should be tested at room temperature. The test value uf this temperature should be compared with preirradiation test data. The toot temperature for the remaining specimens should then be adjusted higher or lower so as to develop a complete transition curve. For succeeding tests after longer irradiation 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 of the selected irradiated materials shall be tested at a temperature represonative of the upper end of the Charpy energy transition region. The remaining tensile specimens from each material shall be tested at the service temperature (560*F) and the midtransition temperature. 1 4.4 FRACTURE TOUGHNESS TESTS ON 1/2 COMPACT SPECIMENS 1 1 in light of current requirements of 10CFR, Part 50, *;4mndix G and applications of ASME Socition lil, Appendix G and Section XI, Appendix A, the %-inch thicit compact specimens should be tested in such a manner as to determine both static, crack initiation, and y 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 test.ed in accordance with ASTM E399 81 procedures modified to account for the size of the specimens available.I'I Specific test procedures should include unloading compliance and data i ~ interpretation should utilize the Equivalent Energy and J Integral concepts.I8'*l

1. Wut. F. J., "Freneure Toughness Parameters Obtained troro Single Smas Spootmen Teens". WCAP4397, ooseber 1973.

2. Suchelet, C. and Weger, T, M. "a ; __ Veresessen e' Lower sound N.Vetuse Umasing the Equivalent Energy Concept."

in Proyees h Plow Grenen and Pressure Taughnese Teodng. ASTM 4TP438, pp. Sci agt. Amerleen Seolsey for Teeung and Messetels. Phandeerda,197s.

3. Landes, AD. and Segley, J. A.. "Recent Deveispmente in J. Teenne". In Deveinamenes h Frecours Mecheruse Test Mechode Sienderadassen ASTM sTP438, pp. 5741. Amerioen Sooteey ter Teeung and Maserials, Phuedolphia,1977,

4. Mecebe, D. E., "oneeransmen of RCurves for swussural Maneriele using Nonenser Wochentos Methods," in Flow GrouW and Frecouro. A81M STP431. pp. 246 296 Ameneen Sootesy ter Teenne and Mesensis, Phaedelphia,1977, 42 1

- me n ,.em ee.ev ,e e-e or-r- -- -=- eem,,,-s,-w.~,-we w.., --w.w-,m,w..e .._-e,,-,a.v.w,.--om-m,.w-vwse-,.-n--.m-m-e,wron-v,-n--,mw,-g..mr.,v,.v -vmv~m,wn--,ne

Fracture toughness data no obtained will be Kg, Ja 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 procedures actually performed on the specimens will reflect state-of the art at the time of testing. 4.5 POSTlRRADIATION TEST EQUIPMENT Required minimum equipment for the postirradiation testing operations is as follows: E Milling machine or special cutoff wheel for opening capsules, dosimeter blocks and spacers. E Hot cell tensile testing ma: hine with pin type adapter for testing tensile specimens. E 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 Sodlum lodido scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters. e e 4 9 4 W 6 4-3 4 A

APPENDIX A DESCRIPTION AND CHARACTERIZATION i OF THE CATAWBA UNIT NO. 2 REACTOR VESSEL BELTLINE AND SURVEILLANCE MATERIALS Based on the initial RTNDT, chemical compostion (copper and phosphorus) and the end-of life neutron fluence, the reactor vessel Intermediate shell plate B86051 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 beltline 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 Catawba Unit No. 2 reactor pressure vessel, specifically, from the 9%-inch intermediate shell plate B86051. Also supplied was a submerged are weldment made from sections of intermediate shell f plate B8605 2I'l and adjacent lower shell plate B88061. This test weldment was fabricated using h inch Mil B-4 weld filler wire, heat number 83648 and Linde 0091 flux, lot number 3536 and is identical to that used by Combustion Engineering, Inc. In the Catawba Unit No. 2 reactor vessel fabrication process specifically the closing girth seam between the intermediate and lower shell plates, and all iongitudinal weld seams of both the Intermodlate and lower shell plates. The chemical analyses,TNDT, RTNDT, upper shelf energy and heat treatment history of all the core region prescure vessel shell plates used in the fabrication of the Catawba Unit No. 2 reactor pressure vessel are summarized in Tables A-1 thru A-5 respectively. This data is as reported in the vessel fabricators (Combustion Engineering, Inc.) certification reports or from subsequent Westinghouse analyses of similar materials used for the Catawba Unit No. 2 surveillance program. Wold material identical to that used in the fabrication of the core region beltline welds [b] have been correlated with the Westinghouse surveillance program test weldment and available Combustion Engineering, Inc. weld certification reposts and their surveillance program test weldment. This data is also reported in Tables A-3 thru A 5 of this Append.1x. ,I

a. The limiting plate material selected in 197s was intermediate shell plate B8606-2. This selectim was based at the time on the highest initial RT,. Therefore weld test plate "D" fumished to Westinghouse at that time was made up of plates B6605-2 and B88061.

b. The belti6ne welds are considered to include the intermediate and lower shell plate longitudinal seams and the closing intermediate to lower shell girth seam. A1 l

X' TABLE A 1 CHEMICAL ANALYSIS OF THE INTERMEDIATE SHELL PLATES USED IN THE CORE REGION OF THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL rt Chemical Compositon (weight %) Element Plate'I Plate >l Plate >l l B86051PI 58605-1 'l 58605 2 B86161 1 C .25 .22 .24 .24 Mn 1.40 1.37 1.35 1.39 P .011 .012 .009 .010 S .013 .013 .013 .021 SI .28 .29 .28 .27 .59 Ni .63 .59 .61 Mo .57 .57 .57 .54 1 Cr .085 .05 Cu .09 .071 .07 .05 Al .042 .043 Co .007 .006 Pb .001 <.001 W <.01 <.01 Ti .004 <.01 Zr <.002 .001 V = ome.o .002 .003 w o.t.ei.o Sn .007 .003 As .008 .005 Cb <.002 <.01 N .008 <.01 2 B <.001 <.001

a. Surveillance program test plate.

b. Chemical Analysis by Combustion Engineering, Inc.

c. Chemical Analysis by Westinghouse.

A2 a.--e- ~ ,,e -n

TABLE A 2 / CHEMICAL ANALYSIS OF THE LOWER SHELL PLATES USED IN THE CORE REGION OF THE CATABWA UNIT NO. 2 REACTOR PRESSURE VESSEL Chemical Compositon 'I l Element Plate Plate Plate B88061 B8806 2 l 88806 3 C .23 .19 .20 Mn 1.40 1.33 1.35 P .009 .007 .006 S .016 .013 .013 Si .22 .23 .23 Ni .56 .59 .59 Mo .57 .55 .55 j Cr .12 .03 .03 Cu .05 .05 .05 Al .021 .027 .027 Co .006 .006 .005 Pb <.001 <.001 <.001 W <.01 <.01 <.01 Tl. <.01 <.01 <.01 Zr .002 .001 .001 V .004 .003 .002 Sn .001 .002 .001 As .003 .004 .002 Cb '<.01 <.01 <.01 N .007 .006 .006 2 B '<.001 <.001 <.001 N. Chemical Analysis by Combustion Engineering, Inc. A3 .--.,.,+r.

4 TASLE A-3 CHEMICAL ANALYSIS OF THE WELD METAL USED IN THE CORE REGION WELD SEAMS OF THE V CATAWSA UNIT NO. 2 REACTOR PRESSURt! VESSEL wom n. e e reven meannoi wome no a ww.ed w inohans v ini mee = and io.= 5 shes piene lenpavenal seems and v. Joining ireenneeste w lower ehes gMN esem. As core repen Seminel weWe were febrtomied using Wold Wire Heel No. 836et. Unde 0001 Phot. LA No. 3834. Chem on Element Wire Flux Actual Production Wootinghouse Test Wold Wold (Lower Surveillance Sample *l Shell Longituding Program T l I Seem 101 142A)I Weldment C 13 .14 15 Mn 1.23 .88 1.20 P 005 .008 010 S 009 .012 010 Si 13 .12 15 Ni .14 14 ( Mo 59 .44 60 .03 052 Cr Cu t)4 .04 036 Al .001 002 I .017 009 Co Pb l Not detected 001 W .01 <.01 Ti <.01 .005 Zr .002 <.002 V .006 .004 .004 Sn .008 .005 As .013 .002 Cb .017 <.002 N .007 .004 2 B <.001 <.001

a. Chemical Analysis of Wire-Flux Wold Sample, Test Number D32255 and Chemistry Data Sheet 101-142A by Combustion Engineering, Inc.

b. Chemical Analysis by Westinghouse of Test Sample Supplied by Combustion Engineering, Inc. Representative of the Closing Girth Seam Weld, Wold Wire Heat No.83648,0091 Flux, Lot No. 3536.

i A4 -.-,,-,,,.-.,.,.y-,-re.-,ww-y ir -e <--rr w .~....-,------...r--._w_..

TABLE A 4 TND;,RTNDT AND UPPER SHELF ENERGY FOR THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES AND WELD METAL i Average I*I I TNDT RTNDT Upper sheroi Meterial Energy ('C) ('F) (*C) ('F) (J) (ft Ib) Intermediate Shell Plates: B86051 - 23 - 10 -9 15 121 89 88605 2 - 29 - 20 1 33 111 82 B86161 -18 0 - 11 12 125 92 Lower Shell Plates: 88806 1 - 51 - 60 - 14 6 113 83 B8806 2 - 40 - 40 - 23 - 10 138 102 88806 3 - 40 - 40 - 13 l 8 142 105

a. Data obtained from Combustion Engineering, Inc. Reactor Vessel Material Certification Reports.

b. Drop weight data obtained from the transverse material properties (normal to the major working direction),

c. From impact data obtained from the transverse material properties (normal to the major working direction).

I TNDT RTNDT Material Energy ('C) ('F) ('C) ('F) (J) (ft Ib) Intermediate and Lower Shell Longitudinal Weld Seams and Closing Girth - 82 - 80 - 62 - 80 176 130 Weld Seam (Weld Wire Ho t No. 83648, Undo 0091 Flux, Lot No. 3536) (

d. Data obtained from Combustion Engineering, Inc. Wire / Flux Wold Deposit Material Certification Test No.1332.

A-5

TABLE A 5 H2AT TREATMENT HfSTORY OF THE CATAWBA UNIT NO. 2 REACTOR PRESSURE VESSEL r CORE REOlON SHELL PLATES AND WELD SEAMS Temperature TimeM (hr) Cooling Material ('F) Water-quenched Austenitizing: 4 1600 i 25 Intermediate (871'C) Shell Plates Tempered: 4 Alr cooled B86051 1225 i 25 B8605 2 (663'C) B86161 Stress Rollef: 20lbl Fumace-cooled 1150 i 50 (621*C) Austenitizing: 4 Water-quenched 1600 i 25 Lower (871*C) Shell Plates Tempered: 4 Air cooled B8806-1 1225 i 25 B8806-2 (663'C) 88806-3 Stress Relief: 1pbl Furnace-cooled 1150 i 50 3 (821'C) ~ Intermediate Shell Longitudinal Stress Relief: 20lbl Furnace-cooled Seam Welds 1150 i 50 (621 *C) Lower Shell Longitudinal 1/bl FeacealM Seam Welds Local Intermediate to Stress Rollef: 11 Furnace-cooled Lower Shell Girth 1150 i 50 Seam Weld (621 *C) Surveillance Program Test Material Surveillance Program Weidm,eg Test Post Weld (gD Stress Relief: 11 'l Furnace-cooled P I u ek*as om s m> 1150 i 50 l (621*C)

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