ML20217A662
| ML20217A662 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/31/1990 |
| From: | Mager T, Singer L, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20210U635 | List: |
| References | |
| WCAP-9734, NUDOCS 9709220006 | |
| Download: ML20217A662 (43) | |
Text
{{#Wiki_filter:, Attachment 6 Westinghouse Electric Corporation Topical Report WCAP-9734, Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program.
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WESTINGHOUSE CLASS 3 5 E E 4 DUKE POWER COMPANY CATAWBA UNIT NO.1 REACTOR VESSEL RADIATION E SURVEILLANCE PROGRAM I S. E. Yanichko g July 1980 1 APPROVED: T. R. Mager, Manager l Metallurgicaland NDE Analysis l Work Performed Under DCP 106 I WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems l P. O. Box 355 l . Pittsburgh, Pennsylvania 15230 o
k P t d PREFACE i This report has been technically reviewed and checked by L. R. Singer of Metallurgical g and NDE Analysis. 3 g S$AL. R. Singer Date: June 10,1980 N E lii i
7-J --a i l-] W ABSTRACT
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A pressure vessel steel surveillance program per ASTM E 185 73 has been developed for the Duke Power Company Catawba Unit No.1 to obtain information on the effects of w radiation on reactor pressure vessel material under operating conditions. The radiation surveillance program for the Catawba Unit No.1 is designed to, and in compliance with,
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l federal government regulations identified in appendix H to 10CFR, part 50, entitled
" Reactor Vessel Material Surveillance Program Requirements."
Following is a description of the program, a description of the material involved, the specimen and capsule design, and the preirradiation test results. V
d ]
]
TABLE OF CONTENTS . Section Title Page ] 1 PURPOSE AND SCOPE 11 2 CAPSULE PREPARATION 21 6- 21. Pressure Vessel Material 21 2 2. Machining 21 2 3. Charpy V notch Impact Specimens 21 2 4. Tensile Specimens 23
, 2 5. 1/2T Compact Tension Specimens 23 rd 2 6. Dosimeters 23 2 7. Thermal Monitors 23 2 8. Capsule Loading 29 3 PREIRRADIATION TESTING 31 3 1. Charpy V notch Tests 31 3 2. Tensile Tests 31 3 3. Dropweight Tests 32
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4 POSTIRRADIATION TESTING 41 4 1. Capsule Removal 41 4 2. Charpy V notch Impact Tests 42 4 3. Tensile Tests 42
- 44. Fracture Toughness Tests on 1/2T Compact Tension Specimens 42 4 5. Postirradiation Test Equipment 43 Appendix A CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A-1 vii
_ _ _ _ = _ . _ _ . _ _ _ _
4 LIST OF ILLUSTRATIONS i 4 Figure Title Page l l 21 22 Charpy V notch impact Specimens Tensile Specimen 22 24 23 Compact Tension Specimen 25 24 Irradiation Capsule Assembly 2 7/2 8 25 Dosimeter Block Assembly 2 10 l 26 Specimen Locations in the Catawba Unit No.1 Reactor Surveillance Test Capsules 213/214 31 Preirradiation Charpy V notch Impact Energy for the l Catawba Unit No.1 Reactor Pressure Vessel Intermediate Shell Forging 05 (TangenthI Orientation) 38 32 Preirradiation Charpy V notch Impact Energy for the 5 Catawba Unit No.1 Reactor Pressure Vessel Intermediate Shell Forging 05 (Axial Orientation) 39 33 Preirradiation Charpy V notch impact Energy for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3 10 34 Preirradiation Charpy V notch Impact Energy for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Affected Zone Material 3 11 35 Preirradiation Tensile Properties for the Catawba Unit No.1 Reactor Pressure Vessellntermediate Shell Forging 05 (Tangential Orientation) 3 12 36 Preirradiation Tensile Properties for the Catawba ' Unit No.1 Reactor Pressure Vesselintermediate Shell Forging 05 ( Axial Orientation) 3 13 37 Preirradiation Tensile Properties for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3 14 38 Typical Stress Strain Curve for Tensile Test 3 15 ix
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E E 1 M LIST OF TABLES E Table Title Page 21 Type and Number of Specimens in the Catawba E Unit No.1 Surveillance Test Capsules 29 22 Quantity of Isotopes Contained in the Dosimeter Blocks 2 11 Preirradiation Charpy V notch Impact Data for the l 31 Catawba Unit No.1 Reactor Pressure Vessel ;, Intermediate Shell Forging 05 (Tangential Orientation) 33 l 32 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Intermediate Shell Forging 05 ( Axial Orientation) 34 33 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 35 34 Preirradiation Charpy V notch Impact Data for the Catawba Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Aff,ected Zone Material 36 35 Preirradiation Tensile Properties for the Catawba - Unit No.1 Reactor Pressure Vessellntermediate Shell Forging 05 and Core Region Weld Metal 37 41 Capsule Lead Factors 41 xi
m 4 SECTION 1 l PURPOSE AND SCOPE The purpose of this program is to monitor radiation effects on the reactor vessel materials of the Duke Power Company Catawba Unit No.1, a four loop,3565 megawatt plant, under actual operating conditions. 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 tension specimens. Current reactor pressure vessel material test requirements and acceptance standards utilize the reference nil ductility temperature, RTNDT, as a basis. RT NDT is determined from the dropweight nil ductility transition temperature (NDTT) per ASTM E208 and the weakl11 direction 50 ft Ib Charpy V notch temperature (or the 35 millateral expansion temperature if it is greater). RT NDT is defined as the dropweight NDTT or the temper-ature 60*F less than the 50 ft Ib (or 35 mil) Charpy V notch temperature, whichever is greater. Therefore RTNDT = NDTT, if NDTT >, T50(35) 60* F and RTNDT = T50(35) 60*F,if T50(35) 60 F > NDTT where RTNDT = Reference nil ductility temperature NDTT = Nil ductility transition temperature per ASTM E208 T 50(35) = 50 ft Ib teurperature from Charpy V notch specimens oriented in the weak direction (or the 35 mil temperature if it is greater) 1 Longitud,nal aus of the specirnen oriented normal to !!.4 maior working direct.on of the forging 1-1
gi An empirical relationship between RTNDT and fracture toughness for reactor vessel steels has been developed in appendix G, " Protection Against Non ductile Failure," to [ Section ill of the ASME Boller and Pressure Vessel Code. This relationship can be employed to set allowable pressure temperature limitations for normal operation of l reactors which are based on fracture mechanics concepts. Appendix G defines an t acceptable method for calculating these limitations.
!I lt is known that radiation can shift the Charpy V notch impact energy curve to higher temperatures,li,21 Thus, the 50 ft Ib temperature and RT NDT ncrease i with radiatbn ;
exposure. The extent of the shift in the impact energy curve, that is, radiation embntile-ment, is enhanced by certain chemical elements. (such as copper) present in reactor r vesselsteels.lMI . I1 The 50 f t Ib temperature or RT NDT increase with service can be monitored by a surveil- t lance program involving periodic checking of irradiated reactor vessel surveillance e Il specimens. The surve"hince program is based on ASTM E185 73 (Standard Recom-mended Practice for Surveillance Tests for Nuclear Reactor Vessels). Compact tension i fracture mechanics specimens will be used in addition to Charpy V notch specimens to evaluate the effects of radiation on the fracture toughness of reactor vessel mater-lals. l5,6,7.8,9,10,11 ) i 1 Porter, L< F., Radiaton Iffects in Steel." in Matenals a Nuclear Apphearsons. ASTM STP 2F6. pp 141 195. Amercan ' 's. Philadelphia 1960 Socoty for Testin0 and M { 2 Steele, L. E and H...ino R , "New informaton on Neutron Embrittlement and Emt tittiment Relet of Reactor Pressure i vestel Steels " NRL-6100 att1964 3 Potapovs. u and Hawthorns 4 A , "*he Effect of Residual Elements on 550' F irradiation Response of Selected Pressure Vessel Steels and Weidments." NRL 6803 September 1968 4 Steele L E , " Structure and Compositen Ettects on arradiation Sens*vity of Pressure Vessel Steets." anseradetson Effects oss Structural A#oys for Nucieni Reactor Apphearsons. ASTM STP 484. pp 164175, Amercan Society for Testing and Materals. Ph6tadelphu.19 TO 5 Landerman, E, Yanschko, S. E. and Hazelton W S ,"An Evaluation of Radiation Damage to Reactor Vessel Steels Using Both Transiten Temperature and Fracture Mechanics Approaches." in The fftects of Rad, arson on Structural Metals. ASTM-STP. 426, pp 260 277, Amorcan Society for Testing and M6tenais. Philadelpha.1967. 6 Manpne, M. J ,"Benel Bott e Fracture Tests." Trans Am Soc. Mech. Engrs. 8 7, Ser es D. 293 298 (1965)
- 7. Porse. L., " Reactor Vessel Design Considenng Radeten Ettects." Trans Am Soc Afech Enges 86. Ssties D, 743 749 (1964L 8 Johnson. A E ," Fracture Mechan *cs A Base for Bnttle Fracture Preventon." WAPD TM 505. November 1965 ,
- 9. Wessel. E. T. and Pryte, W H . "investigaton of the Apphcability of the Basel Bnitte Fracture Test for Determining Fracture l Toughness."WERL 884411. August 1965 .
- 10. Wilson. W K., " Analytic Determinaten of Stress intensity Factcs for the Mangne Bnttle Fracture Test Specimen." WERL.
0029 3, s u9ust 1965 l 11 Johnson, R [. and Panerb. E J " Fracture Toughness of trradiated A302-8 Steel as influenced by Merostructure," Trans. i Amer. Nuct Soc. 9. 39D 392 (1966L i 12
4 l Postirradiation testing of the Charpy impact specimena will provide a guide for deter-mining pressure temperature limits on the plant. Charpy impact test data will determine the shift of the reference temperature with radiation exposure at plant temperatures. These data can then be reviewed to verify or revise pressure temperature limits of the vessel during startup and cooldown (the Charpy) specimens are most nearly indicative of the radiation exposure experienced by the vessel:. This will allow a check of the predicted shift in the reference temperature. The postirradiation test results of the compact tension specimens will provide actual fracture toughness properties of the vessel material. These properties may be used to establish allowable stress Intensity factors for subsequent analyses. Six material test capsules, located in the reactor be' ween the neutron st.leiding pads and vessel wall, are positioned opposite the center .:f the core. The test capsules are located in guide tubes attached to the neutron shielding pads. The capsuU contain test specimens from a forging from the reactor vessel Intermdiate shall course adjacent to the core region, representative weld metal, and heat affected zone (H A2, metal. The thermal history or heat treatment given these specimens is similar to the thermal history of the reactor vessel material with the exception that the postweld heat treatment received by the specimens has been simulated (appendix A), 13
h l N SECTION 2 5 CAPSULE PREPARATION 4 2 1. PRESSURE VESSEL MATERIAL Reactor vossol material was supplied by Rotterdam Dockyard Company, from intermo-diato shell forging 05 (Heat No. 411343). Rotterdam Dockyard Company also supplied a woldment which joined sections of material from intermediato shell forging 05 and the adjoining lower shell courso forging 04. Data on the limiting coro region forging, wold, and wold heat affected zone materialare provided in Appendix A. 2 2. MACHINING Test material obtained from the intermediato shell course forging 05 (after the thermal heat treatment) was taken at least one forging thickness from the quench'od ends of the forging. All test specimens were machined from the 1/4 thickness location of the forging af ter performing a simulated postwold stress relieving treatment on the test material and also from wold and heat affected rono metal of a stress relieved woldmont joining inter-mediato shell forging 05 (Hoat No. 411343) and lower shell forging 04 (Heat No. 527708) All heat affected zone specimens were obtained from the weld heat affected zone of forging 05. 2 3. Charpy V notch impact Specimens Charpy V notch impact specimens (figures 21) from intermediate shell forging 05 woro machined in both the tangential orientation (longitudinal axis of specimen parallel to major working direction) and axial orientation (longitudinal axis of specimen perpendicular to major working direction). The core region wold Charpy impact specimens wore machined from the woldmont such that the long dimension of the Charpy specimen was normal to the wold direction. The notch was machined such that the direction of crack propagation in the specimen was in the wold direction. 21
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E E 2 4. Tensile Specimens Tensile specimens (figure 2 2) from forging 05 were machined so as to produce some l with the longitudinal axis of the specimen normal to and some perpendicular to the major working direction of the forging. Specimens from the weld were oriented normal to the E weld direction. l 2 5. Il2T Compact Tension Specimens Compact tension test specimens (figure 2 3) from forging 05 were machined in both the g axial and tangential orientations. Compact tension test specimens from the weld metal were machined normal to the wold direction with the notch oriented in the direction of the wold. All specimens were fatigue procracked according to ASTM E399. B . I 2 6. DOSIMETERS l Each of the six test capsules of the type shown in figure 2 4 contain dosimeters of pure copper, iron, nickel, and aluminum 0.15 weight percent cobalt wire (cadmium shielded g and unshielded) and cadmium shielded NP237 and Uras which will measure the integrated flux at specific neutron energy levels. l 2 7. THERMAL MONITORS The capsules contain two low melting point outectic alloys to more accurately define the maximum temperature attained by test specimens during irradiation. The thermal monitors are sealed in Pyrox tubes and then inserted in spacers located as shown in figure 2 4. The two outoctic alloys and their melting points are the following: I
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28, CAPSULE LOADING The six test capsules coded U, V, W, X, Y, and Z are positioned in the reactor between i the neutron shielding pads and ve:sel wall at the locations shown in figure 2 4. Each I capsule contains 60 Charpy V notch specimens,9 tensile specimens, and 12 compact tension specimens. The relationship of the test material to the type and number of speci-mens in each cap 3ule is shown in table 2 1. TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE CATAWBA UNIT NO.1 SURVEILLANCE TEST CAPSULES Capsules U, V, W, X, Y, and Z Material Charpy Tensile CT Forging 05 (Tangential) 15 3 4 (Axial) 15 3 4 Wold Metal 15 3 4 HAZ 15 - - Dosimeters of pure copper, iron, nickel, aluminum 0.15 weight percent cobalt, and cad-mium shielded aluminum cobalt wires are secured in holes drilled in spacers located at capsulo 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 capsules, each containing isotopes of either of U U or Np237, are located in the dosimeter block. The double containment afforded by the dosimeter assembly prevents loss and contamination by the U238 and Np237 and their activation products. Each dosimeter block contains approximately 12 milligrams of U238 and 17 milligrams of Np237 (table 2-
- 2) held in a 3/8 inch long by 1/4 inch cuiside diameter sealed stainless steel tube, respectively. Each tube was placed in a 1/2-inch diameter hole in the dosimeter block 29
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SECTION 3 N \ PREIRRADIATION TESTING
- 31. CHARPY V NOTCH TESTS Charpy V notch impact tests were performed per ASTM E 23 with specimens from the
' vesselintermediate shell forging 05. Specimens of both axial and tangential orientations were tested at various test temperatures in the range -100 to 210'F yielding a full Charpy V notch transition curve in both orientations (tables 31 and 3 2 and figures 31 and 3 2). Tests were also performed on weld metal and HAZ metal at various temper-atures from -200 to 210'F. The results are reported in tables 3 3 and 3 4 and figures 3 3 and 3 4.
The specimens were tested on a Sontag SI 1 impact machine which is inspected and calibrated every 12 months Charpy V notch impact specimens of known energy values, supplied by the Watertown Arsenal, are used for the calibration. 3 2. TENSILE TESTS Table 3 5 and figures 3 5,3 6, and 3 7 give results of tensile tests (per ASTM E 8 and E 21 test criteria) from vessel intermediate shell forging C5 and from the weld metal. Specimens from the shell forging were tested at room temperature, 300* F, and 550' F in both the axial and tangential directions. An Instron TT C tensile testing machine was used with the standard Instron gripping devices. A Baldwin Lima Hamilton Class B 1 extensometer and chart recorder provided a full stress strain curve for each specimen. The chart recorder was calibrated to the Class B 1 extensometer. The measurement and control of speeds in the tests conformed tc %STM A370 68 (MechanicalTesting of Steel Products). The Instron TT C and the Baldwin Lima Hamilton extensometer are calibrated by test equipment which has been certified by the National Bureau of Standards. A typical stress strain curve is
, shown in figure 3 8.
31
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I l 3 3. DROPWEIGHT TESTS The nil ductility transition temperature (TNDT) was determined for forging 05 and the coro region weld metal and heat affected zone by dropweight tests (ASTM E 208) performed at Rotterdarn Dockyard Co. The following results were obtained: l Material T NDT (* F) Forging 05 - 40 .
- 76 ,
Weld Metal
- 67 HAZ __
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E 4 TABLE 31 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05(TANGENTIAL ORIENTATlON) Test Temperature impact Energy Shear Lateral Expansion ('F) (ft lb) (%) (mils)
- 100 12 0 '
1
- 40 15 0 10.5
- 40 11 0 8 0 15 23 12 0 50 27 37 0 17 33 15 10 17 42 14 10 24 33 19.5 20 85 45 56 20 85 48 62 20 75 52 51 75 126 73 82 75 116 66 80 120 139 81 85 120 130 81 86 210 158 100 94 210 177.5 100 83 210 168 100 88 33
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- F) (ft Ib) (%) (mils)
- 100 2 0 0
- 40 8 0 5.5
- 40 16 0 11
- 15 48 25 34 0 27 20 19 6l 0 35 30 25 0 51 30 38 20 60 34 44 20 56 40 44 20 69 45 52 75 81 52 60 75 89 61 67 120 111 82 77 120 119 81 79.5 150 135 -
100 90 . 210 137 100 88 : 210 132 100 91 9 210 133 100 87 .
~
l I t 4 4 e P 34 l l
., m .. o ..-.r
TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA g FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Test Temperature impact Energy Shear Lateral Expansion ('F) (ft Ib) (%) (mils)
- 100 12 13 6
- 60 13 28 9.5
- 60 15 33 11.5
- 40 37 33 26
- 40 26 28 18
- 16 40 42 31
- 16 60 37 45.5
- 16 54 50 39 5 44 54 38 25 91 72 66 '
25 98.5 87 74 75 119 87 86 75 110 95 77 120 132 100 90.5 120 119 100 87 210 133 100 90
- 210 130 100 90 210 124 100 89 I
35 m
." ' . _ _lil. l_'~"^ " " '
TABLE 3 4 , PREIRRADIATION CHARPY V NOTCH IMPACT DATA : FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATERIAL Test Temperature impact Energy Shear Lateral Expansion (' F) (ft ib) (%) (mils)
- 200 13 0 4
- 170- 35 33 22
- 150 96 80 54
- 100 82 38 43
- 40 74 54 48
- 40 108 50 58
-7 135 81 74
-7 110 77 60
-7 127 81 73 '
25 159 100 85 25 147.5 89 S1 75 150 100 83 75 138.5 93 80.5 120 154 100 84 120 162 100 83 210 180 100 81.5 210 159 100 85 , 210 166 100 83 4 sm l l 36 1 4 - - .. .. , ...,_ .....
h; TABLE 3-5 I PREIRRADIATION TENSILE PROPERT!ES FOR THE CATAWBA UNIT NO.1
; REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 } AND CORE REGION WELD METAL L
I h O.2% Ultimate [ Test Yield Tensile Fracture Fracture Uniform Total Reduction l Temp. Strength Strength Load Stress Elongation Elongation in Area j Material *F (ksi) (ksi) (Ib) (ksi) % % %
! Forging 05 (Ht. 411343) 70 69.0 89.1 2580 191.4 16.2 29.6 72.6 (Tangential Orientation) 70 68.7 89.0 2600 198.9 15.6 28.6 73.4 300 61.9 81.2 2500 177.3 12.4 24.4 71.4 300 60.9 80.3 2700 186.0 13.0 24.2 70.3 550 61.2 85.6 2650 167.4 13.5 25.3 67.9 g 550 60.3 85.1 2625 154.7 13.4 24.8 65.6 O Forging 05 (Ht. 411343) 70 66.5 87.1 2600 198.9 17.0 29.0 73.4 (AxialOrientation) 70 66.5 86.9 2700 191.5 16.4 28.2 71.4 300 62.0 80.8 2425 157.6 8.3 17.2 68.7 300 60.4 80.2 2700 150.9 11.2 23.0 63.6 550 58.8 84.6 2825 162.0 14.7 25.5 64.7 550 59.7 85.0 2725 160.6 15.0 25.3 65.6 Weld Metal 70 75.3 88.1 2575 194.0 15.0 27.6 73.2 70 75.0 88.0 2600 195.9 14.7 27.4 73.2 300 68.6 79.3 2375 171.0 11.3 24.3 71.8 300 67.0 79.2 2450 171.2 11.1 24.0 71.0 550 64.8 80.2 2550 156.6 11.7 22.4 67.0 550 65.2 80.3 2500 164.8 11.6 22.8 69.3 ,
200 i O 160 - O
~
O O
- 120 -
6 I E y 80 -
-0
= !
o ! 40 - O 4 EP o l I I I 100 0 100 200 300 TEMPERATURE (
- F)
I FIGURE 31 PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO.1 , REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 (TANGENTIAL ORIENTATION) 38 l
f 160 O 120 - I
> 80 -
0 0 E O w Z w C O' 40 - 0 0 ! ! I
-1'00 0 100 200 300 TEMPERATU RE ( ' F)
FIGURE 3 2 PREIRRADIATION CHARPY V NOTCH IMPA"T ENERGY FOR THE CATAWBA UNil NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 l (AXIAL ORIENTATION) 39
~
? 160 1 O 120 - O E (
>' Bo -
W Z ! W 40 - O I I I I o 100 o 100 200 soo TEMPERATURE (
- F)
I FIGURE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE CATAWBA UNIT NO,1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3 10
E E 200 "- O . N 100 - o g o o O O
- O 3 120 -
0 Z < E s o O 80 - O i. N ' 40 - I I I I I I o ,
-200 -100 0 100 200 300 TEMPERATURE (* F)
FIGURE 3-4 PREIRRADIATION CHARPY V NOTCH IMPACT l ENERGY FOR THE CATAWBA UNIT NO 1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED-ZONE MATERIAL 3 11
100 l I I I I S 4 - x Bo - ULTIMATE TENSILE STRENGTH So - O E 0.2% YlELD STRENGTH do , Bo Q et v o V REDUCTION IN AREA
, So -
3 40 - F b %_ TOTAL ELONGATION D 20 - S
& D D UNIFORM ELONGATION o_ l i I I I o 100 200 300 400 soo Soo TEMPERATURE (
- F)
FIGURE 3 5 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT N,0,1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FOHGING 05 (TANGENTIAL ORIENTATION) 3 12
I I
'U l l l l l g ULTIMATE TENSILE STRENGTH 5
g 80 - 5 W
-g y 2- k 5 g 60 -
O- g 0.0% YlELD STRENGTH E 40 ! l1 1 80 I R E 8 o 60 - REDUCTION IN AREA m a e 1 g tg 40 - F TOTAL ELONGAT W A C UNIFORM ELONGATION O l l l l l 0 100 200 300 400 500 600 TEMPERATU RE (' F) FIGURE 3-6 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 05 l (AXIAL ORIENTATION) W l g 3 13
+ - --
'UU l l l l l ,
C ULTIMATE TENSILE STRENGTH k 80 - 1 m m E O F 60 - 0.2% YlELD STRENGTH m 40 80 2-c g- n REDUCTION IN AREA
- 60 -
F 2 40 - F o 3 Q_- TOTAL ELONGATION O 20 - O g (1_
~
UNIFORM ELONGATION 0 t l I I l l 0 0 100 200 300 400 500 600 TEMPERATURE (* F) FIGURE 3 7 PREIRRADIATION TENSILE PROPERTIES FOR THE CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3 14 ' m ww wm m
p,
~
m G H NM C Q H 8 d o O x .E 5 $ 0; o?
! e i M g
g CO b a A m cn C I i 1 SS3EIS 4 I l 3 15
- = - _ = _ ____ .__ _ _rr__ _ _ _ _ _ _-
SECTION 4 j POSTlRRADIATION TESTING 1
- I 8 4-1. CAPSULE REMOVAL Il Specimen capsules should be removed ftom the reactor only during normal refueling periods. Table 4 1 lists the capsule identification and their respectiva lead factors.
Each specimen capsule, removed af ter radiation exposure, will be transferred to a post-Irradiation test facility for disassembly and testing of all the specimens. Although the lead factors are slightly greater than the limits specified in Appendix H to 10CFR Part 50, the capsule results will not be affected. The first capsule (Capsule U) should be removed at the end of the first core cycle. Subsequent capsules should be removed per the new schedules identified in ASTM E185 and Appendix H to 10CFR Part 50 which are
. currently being revised.
TABLE 41 CAPSULE LEAD FACTORS Multiplying Factor By Which the Capsule Leads Capsule identification Vessel Maximum Exposure U 4.05 X 4.05 V u.37 Y 3.37 l W 4.05 Z 4.05
f 4 2. CHARPY V 1DTCH IMPACT TESTS The testing of the Charpy impact specimens from the intermediate shell forging, weld metal, and HAZ metalin each capsule can be done singly at approximately ten different temperatures. The extra specimens should be used to run duplicate tests at tempera-tures 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 for this temperature should be compared with preirradiation test data. The test ten.perature for the remaining specimens should then be adjusted higher or lower so as to develop a complete transition curve. For succeeding tests af ter 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 The tensile tests of specimens for each of the irradiated materials should be performed at room temperature,300' F, and 550* F, and in accordance with ASTM E 8 and E 21 testing criteria. 4 4. FRACTURE iOUGHNESS TESTS ON 1/2T COMPACT TENSION SPECIMENS In light of current requirements of 10 CRF, Part 50, appendix G,1/2T compact tension (CT) specimens should be tested dyramically to adequately characterize the fracture toughness properties of the reactor vessel up to the initiation of the fracture toughness upper shelf. The CT specimens for each of the irradiated materials should be tested in accordance with ASTM E399 74 with appropriate modifications necessary for dynamic tests. Testing dynamically in the fracture toughness ductile to brittle transition region and at upper shelf initiation temperatures results in not only lower bound data but also provides an opportunity for obtaining validlil fracture te*Jghness data up to the onset of upper shelf. This results from nonlinear cleavage behart which occurs only in dynamic testing at these temperatures. The load displacement curve exhibits a clear drop in load at the onset of crack initiation, thereby eliminating any possible doubt as to the start of crack initiation, as is the case in static loading conditions at these temperatures. Recom-i mended test temperatures are equal to or lower than those charactaristic of the upper fracture toughness shelf initiation temperature. t
- 1. Rccardelia. P. C. and Swediow J L. "A Combined A.,arytcabExpenmental Fracture Study of the Two Leading Theones of Elastm Plaste Fracture (Jantegral and Equivalent Energy)," HSST.TR-33. WCAP 8224. October 1973. l 42 1
. .,__s
Analysis should be performed using the J Integral or Equivalent Energy Concept.It2] Testing at temperatures characteristic of the fracture toughness upper shelf is not suggested due to the uncertainty of the point of crack initiation even when dynamic testing is performed. At these temperatures, static Je testing appears to be most indica-tive of conservative upper shelf fracture toughness properties. Research in this area is ] ~ currently being conducted by Westinghouse Research and Development Laboratory,
- ASTM E24, NRC, and others. Use of this technique will be fudher evaluated as it applies to surveillance specimen testing.
4 5. POSTIRRADIATION 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 W Hot cell tensile testing machine with:
- 1) pin type adapter for testing tensile specimens E Hot cell dynamic CT testing machine with clevis and appropriate measuring equipment associated with dynamic testing E Hot cell Charpy impact testing machine E Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters i
1 Rccardella. P C and Swediow, J. L "A Combined Analytical-Expenmental Fracture Study of the Two Leading Theones of Elashc Ptastic Fracture (J integral ano Equivalent Energy)." HSST.TR 33. WCAP-8224. October 1973
- 2. Buchalet C. and Mager, T. R , "Empernental Venfcahon of Lower Bound Kg Values Utihang the Equivalent Energy Concept." in Progress m Flaw Grovvth and Fracture Toughness Testmg. ASTM STP 536, pp 281296, Amencan Socety for Testrng and Matenais. Phdadetotva.1013 43
_p
APPENDIX A l' CATAWBA UNIT NO.1 REACTOR PRESSURE VESSEL SURVElLLANCE MATERIAL For the reactor vessel radiation surveillance program, Rotterdam Dockyard Company supplied Westinghouse with sections of SA508 Class 2 forging used in the core region of the Catawba Unit No.1 reactor pressure vessel, specifically, from the 8% inch intermediate shell course forging 05 of the pressure vessel. Also supplied was a weld-ment made from sections of intermediate shell forging 05 and lower shell course forging 04, using weld wire representative of that used in the original fabrication for the closing girth weld seam between the intermediate and lower shell courses. The surveillance weldment is identical to the closing girth seam weldment between forgings 04 and 05. The closing seam used weld wire heat no. 895075 with Type Grau L.O. (LW320) flux, lot P46, except for the 1 inch root pass at the ID of the vessel. This root pass used weld wire of heat no. 899680 with type Grau L.O. (LW320) flux, lot P23, with an as-deposited copper and phosphorous content of 0.03 and 0.009, respectively. The surveillance specimens were not removed from this root area. The forgings were produced by Klockner Werge AG. The heat treatment history and the chemical analysis of the pressure vessel surveillance materials are shown in tables A 1 and A 2, respectively. TABLE A-1 HEAT TREATMENT HISTORY Temperature Time (hr) Cooling Material (
- F) intermediate shell 1679-1697 3% Water-quenched forging 05 1220-1247 6 Air-cooled 1140 25 22 Furnace-cooled Weldment 1140 t 25 15 Furnace cooled A-1 l
_i
l' i TABLE A 2 ' CHEMICAL ANALYSIS OF MATERIALS Content in Indicated Material (weight %) Element Intermediate Shell Weld Forging 05(al Metall*1
.20 .049 C
.73 1.73 Mn
.014 .015 P
.004 .006 S
.35 .27 Si
.84 .71 Ni
.53 .56 Mo
.38 .036 Cr
.10 .066 Cu
.046 .025 Al
.024 .013 Co
.001 .003 Po
.010 .008 W
< ,001 .002 Ti
.002 .002 Zr
.003 .002 V
,001 Sn .005
.023 .011 As Cb <.002 <.002
.009 .008 N2 Zn .010 .010 Mg .002 .002 Ag .001 <.0005 B < 0005 <.0005
- a. Analysis conducted by Westinghouse A2 l
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