ML20010C838

Union Electric Co,Callaway Unit 1,Reactor Vessel Radiation Surveillance Program

ML20010C838
Person / Time
Site:	Callaway
Issue date:	05/31/1981
From:	Mager T, Singer L WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20010C835	List: ML20010C834 ML20010C838
References
WCAP-9842, NUDOCS 8108210092
Download: ML20010C838 (42)

Text

_ _ _ _ _ _ _

i WESTINGHOUSE CLASS 3 l*

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UNION ELECTRIC COMPANY CALLAWAY UNIT NO.1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM f

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L. R. Singer May 1981 M

APPROVED:

T. R. Mager, thnager Metallurgicaland NDE Analysis Work Performed Under SNP-106 WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 855 Pittsburgh, Pennsylvania 15230 i

8108210092 810819 PDR ADOCK 05000483 A

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PREFACE This report has been technically reviewed and checked by E. H. Williams of Metallurgical and NDE Analysis.

W

. law E. H. Williams Date: May 13,1981 l

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I A pressure vessel steel surveillance pmgram per ASTM E 185-73 has been developed l

fcr the Union Electric Company Callaway Unit No.1 to obtain,information on the effects of radiation on reactor pressure vessel material under operating conditions. The radiation j

surveillance program for the Callaway Unit No.1 is designed to, and in compliance with, j

federal government regulations identified in appendix H to 10CFR, part 50, entitled j

" 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 fabrication, and the preirradiation test results.

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TABLE OF CONTENTS i

S 3ction Title Page 1

PURPOSE AND SCOPE 1-1 2

CAPSULE PREPARATION 2-1 21.

Pressure Vessel Material 2-1 2-2.

Machining 2-1 2-3.

Charpy V-notch Impact Specimens 21 2-4.

Tensile Specimens 23 2 5.

1/2T Compact Specimens 23 2 6.

Dosimeters 2-3 2-7.

Thermal Monitors 2-3 2-8.

Capsule Loading 29 1

3 PREIRRADIATION TESTING 3-1 3-1.

Charpy ' *-notch Tests 31 3-2.

Tensile Tests 3-1 3-3.

Dropweight Tests 32 4

POSTlRRADIATION TESTING 4-1 4 1.

Capsule Removal 41 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 t

4 5.

Postirradiation Test Equipment 4-3 Appen' dix A CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A-1 O

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LIST OF ILLUSTRATIONS Figure Title Pagt 22 2-1 Charpy V notch impact Specimens 22 Tensile Specimen 2-4 23 Compact Specimen 25 24 Irradiation Capsule Assembly 2 7/2-8 25 Dosimeter Block Assembly 2 10 2-6 Specimen Locations in the Callaway Unit No.1 Reactor Surveillance Test Capsules 2-13/2-14 3-1 Preirradiation Charpy V-notch Impact Energy for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 3-8 32 Preirradiation Charpy V-notch Impact Energy for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 3-9 3-3 Preirradiation Charpy V notch Impact Energy for the Callaway Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-10 3-4 Preirradiation Charpy V-notch Impact Energy for the Callaway Unit No.1 Reactor Pressure Vessel Core Region Weld Heat-Af fected-Zone Material 3 11 35 Prcirradiation Tensile Properties for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R27081 (Longitudinal Orientation) 3-12 36 Preirradiation Tensue Properties for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 3-13 3-7 Preirradiation Tensile Properties for the Callaway 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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LIST OF TABLES Table Title Page 21 Type ard Number of Specimens in the Callaway Unit No.1 Surveillance Test Capsules 29 22 Quantity of Isotopes Contained la the Dosimeter Blocks 2 11 3-1 Preirradiation Charpy V-notch Impact Data for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 33 3-2 Preis radiation Charpy V notch impact Data for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 3-4 3-3 Preirradiation Charpy V-notch Impact Data for the Callaway Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-5 3-4 Preirradiation Charpy V-notch Impact Data for the Callaway Unit No.1 Reactor Pressure Vessel Core 4

Region Weld Heat Affected-Zone Material 36 3-5 Preirradiation Tensile Properties for the Callaway Unit No.1 Reactor Pressure Vessel Lower Shell Plate R2708-1 nnd Core Region Weld Metal 3-7 4-1 Surveillance Capsule Removal Schedule 41 i

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SECTION 1 PURPOSE AND SCOPE The purpose of this program is to monitor radiation effects on the reactor vessel materials of the Union Electric Company Ca!!awa9 Unit No.1, a four-loop, 3425-megawatt plant, under actual operating conditions. Evaluation of the radiation effects is based on preirradiation testing of Charpy V notch, tensile, and dropweight specimern and postirradiation testing of Cha:py V-notch, tensile, and compact specimens.

Current reactor pressure vessel material test requirements and acceptance standards NDT s determined i

utilize the reference nil-Juctility temperature, RTNDT, as a basis. RT from the dropweight nil-ductility transition temperature (TNDT) per ASTM E208 and the weakD1 direction 50 ft Ib Charpy V-notch temperature (or the 35-mil lateral expansion NDT s defined as the dropweight TNDT or the temper-temperature if it is greater). RT i

ature 60*F less than the 50 f t ib (or 35-mil) Charpy V notch temperature, whichever is greater.

Therefore RTNDT = TNDT. If TNDT&T50t35)-60 F and RTNDT = T50(35) - 60 *F, if T50(35)- 60*F > TNDT where RTNDT

= Reference nil-ductility temperature TNDT

= Nil-ductility transition temperature per ASTM E208

= 50 f t lb temperature from Charpy V-aotch specimens oriented in T50(35) the weak direction (or the 35-mit temperature if :t is greater) a

1. Lorytudinal axis of the specernen or: entad normal to the m,3er wormng 1rection of toe ofate 11

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 lli 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 mechanicmconcepts. Appendix G defines an acceptacle method for calculating these limitations.

It is known that radiation can shift the Charpy V-notch impact energy curve to higher NDT ncrease with radiation temperatures.[1,21 Thus, the 50 ft Ib temperature and RT i

exposure. The extent of the shif t in the impact energy curve, that is, radiation embrittle-ment, is enhanced by certain chemical elements (such as copper) present in reactor vessel steels.l3 Al l

The 50 f t Ib temperature or RTNDT ncrease with service can be monitored by a surveil-i lanco program involving periodic checking of irradiate'd reactor vessel surveillance J

specimens. The surveillance program is based on ASTM E185-73 (Standard Recom-mended Practice for Surveillance Tests for Nuclear Reactor Vessels.) Compact frac-ture mechanics specimens will be used in addition to Charpy V notch specimens to evaluate the effects of raoktion on tha fracture to';ghness of reactor vessel mater-i ials. [5,6,7,8,9,10,11]

1.

Porter. L. F.

9adetion Effects in Steet." in Metenars rn Nuclear Applications. ASTM-STP 276. pp 147-i95. Amencan Society for Testing and Matenals. Philadelphia. t 960.

2. Steele. L. E. and Hawthorne, J R., "New Information on Neutron Embnttlement and Embnttirnent Relief of Reactor Pressure Vessel Steels " NRL-6160. August 1964.

3 Potapovs. U. and Hawthorne. J R.. "The Effect of Residual Elements on 550' F trradiation Response of Selected Pressure Vessel Steels and Weldrrents." NRL-6803.. September 1968

4. Steele. L. E. Structure and Composition Effects on irradiation Sens*vity of Pressure Vessel Steels." interadration Effects on Structural Alloys for Nuclear Reactor AppI> cations. ASTM STD 484. Op 164115. Amencan Society for Testing and Matenats.

i Philadelphia.1970.

5. I enderman. E. Yanichko. S E and Hazetton. W. S.. "An Evaluation of Radiation Damage to Reactor Vessel Steels Using Botn Iransition Temperature and Fracture Mechanics Approaches." in The Effects of Rad.ation on Structural Metals. ASTM STP-426, pp 260 277. Amencan Society for Test.ng and Matenais. Philadelphia 196 7.

4 6 Mantoine M. J "B amal Dnttle Fracture Tests." Trans. Am. Soc Mech. EngrF 6 7. Senes D. 293-298 t1965)

7. Porse. L, " Reactor Vessel Design Considering Radiat on Effects." Trans. Am Soc. Mech. Engrs 86. Senes D. 743 749 i

(1964) 8 Johnson. R E,' Fracture Vechanics A Base for Br ttle Fracture Prevention."WAPD-TM-505. Noverrber 1965 9 Wessel E T. and Pryte. W H. 'Irwestigatron $t the Applicabil'ty Cf the Bianal Br ttle Fractu e Test for Determining Fracture r

Toughness."WERL 884411. August 1965 10 Wtison W K. " Analytic Determinaton of Stress intens4ty Factors for the Mar.joine Bnttte Fracture Test Specimen." WERL-0029 3 August 1965

11. Johnson. R E and Pas erb E. J ~ Fracture Toughness of irradiated A302 8 Steet as in'tuenced by M.crostructure ' 7rans.

Amer Nuct Soc 9 390-392(1966).

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s Postirradiation testing of the Charpy impact specimens 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

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vessel during startup and cooldown (the Charpy) specimens are most nearly indicative l

of the radiation exposure experienced by the vessel. This will allow a check of the l

predicted shift in the reference temperature. The postirradiation test results of the compact specimens will provide actual fracture toughness properties of the vessel material. These properties may be used to establish allowable stress intensity factors for subsequent analyses.

Six material test capsules, located in the reactor be+ ween the neutron shielding pads and I

vessel wall, are positioned opposite the center of the core. The test capsules are located in guide tubes attached to the neutron shielding p;.ds. The capsules contain test specimens from a plate from the reactor vessel lower shell course adjacent to the core region, representative weld metal, and heat affected-zone (HAZ) metal.

'The thermal history or heat treatment given these specimens is similar to the thermal history of the reactor vessel material with the exception that the postweld heat treatment received by the specimens has been simulated (appendix A).

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1 SECTION 2 CAPSULE PREPARATION 2-1.

PRESSURE VESSEL MATERIAL Reactor vessel material was supp?ied by Combustion Eagineering, Inc. from lower shell plate R2708-1, Heat No. C4499-2. Combustion Engineering, Inc., also supplied a weldment which joined sections of material of the lower shell plate R2708-1 and the adjoining intermediate shell plate R2707-1, Heat. No. C43441. Data on the limiting core region plate, weld, and weld heat affected-zone material are provided in Appendix A.

22.

MACHINING Test material obtained from the lower shell plate (after the thermal heat treatment and

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forming of the plate) was taken at least one plate th'J< ness from the quenched ends of the plate. All test specimens were machined from the 1/4 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 waldment joining lower shell plate R2708-1 and adjoining intermediate shell plate R2707-1. All heat-affected-zone specimens were obtained from the weld heat affected-zone of lower shell plate 1

R27081.

2-3.

Charpy V-notch impact Specimens Charpy V notch impact specimens (figures 21) from lower shell plate R2708-1 were machined in both the longitudinal orientation (long;tudinal axis of specimen parallel to major rolling direction) and transverse crientation (longiicdinal axis of specimen normal to major rolling direction). The core region weld Charpy impa t specimens were machined from the weldment such that the long dimension of the Chrpy soecimen was normal to the weld direction. The notch was machined such that the dir3ction of crack propagation in the specimen was in the welding direction.

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Tensile Specimens l

Tensile specimens (figure 2 2) from shell plate R2708-1 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld were oriented normal to the welding direction.

25, 112T Compact Specimens Compact test specimens (figure 2 3) from shell plate R2708-1 were machined in both the longitudinal and transverse orientations. Compact test.specirnens 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.

DOSitiETERS 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 and unshielded) anc cadmium shielded NP237 and U2as 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 Pyrox tubes and then inserted in spacers located as shown in f

figure 2-4. The two eutectic alloys and their melting points are the following:

l 2.5 percent Ag,97.5 percent Pb Melting point: 579 *F 1.75 percent Ag,0.75 percent Sn,97.5 percent Pb Melting point: 590 F l

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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 in each capsule is shown in table 21.

TABLE 21 TYPE AND NUMBER OF SPECIMENS IN THE CALLAWAY UNIT NO.1 SURVEILLANCE TEST CAPSULES Capsules U, V, W, X, Y, and Z Material Charpy Tensile CT i

Plate R2708-1 1-Longitudinal 15 3

4 Transverse 15 3

4 Weld Metal 15 3

4 HAZ 15 l

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 4

capsule positions shoivn in figure 2-4. Each capsule also contains a dosimeter block t

(figure 2-5) located at the center of the capsuie. Two cadmium oxide-shielded capsules, each containing isotopes of either of U238 or Np:37, are located in the dosimeter block.

The double containment afforded by the dosimeter assembly prevents loss and contamination by the U2aa and Npaar and their activation products. Each dosimeter block contains approximately 12 milligrams of U228 and 17 milligrams of Np237 (table 2-

2) held in a 3/8 inch long by 1/4 inch outside-diameter sealed stainless steel tube, 1

respectively. Each tube was placed in a 1/2 inch-diameter hole in the dosimeter block 29

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I MATER!AL NO ITEM TITLE SPECIFCATION REO D.

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l-I (one U238 and one Np237 tube per block), and the space around the tube was filled with cadmium oxide. Af ter placement of this material, each hole was blocked with two 1/16-inch thick aluminum spacer discs and an outer 1/8 inch thick steel cover disc welded in place.

l-l The numbering system for the capsule cpeciment and their locations is shown in figure i

2 6. The specimens are seal welded into a square capsule of austenitic stainless steel to prevent corrosion of specimen surfaces during irradiation. The capsules were hydro-statically tested in domineralized water to collapse the capsule on the specimens so that l

optimum thermal conductivity between the specimens and the reactor coolant is obtained. The capsules were helium-seak tested as a fini inspection procedure. Fabri-cation details and testing procedures are listed in figure 2 4.

TABLE 2 2 l

OUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOOKS j

isotope Wolght(mg)

Compound Weight (mg)

Np2a7 17 1

NpO2 20 1

UO8 14.25 U238 12.0 3

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Cw?

Cw4 Cw3 Cw?

Cwl CW14 CH14 Cwit CH11 CW8 CH8 Cl4 R3 CL2 CL1 CW5 CH5 CW2 Cwi CW13 CH13 Cwie Ono CW7 CH7 CW4 CH4 Cwt LEGEND. CL LOWER SHELL PLATE R2708-1 (LONGITUDIN AL)

CT - LOWER SHELL PLATE R2708-1 (TRANSVERSE)

CW - WELD MET AL CH HEAT-AFFECTED ZONE MATERIAL l

l

P

'J s4 Egu e1 s

lu -

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p3 3

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a I

T T1 T

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T1 T1 T

T T

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(

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7 6

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3 2

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3 2

s i

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t

l-SECTION 3 PREIRRADIATION TESTING 31.

CHARPY V NOTCH TESTS Charpy V-notch impact tests were performed according to ASTM E23 with specimens from the vessel Lwer shell plate H27081. Specimens of both longitudinal and transverte orientations were tested at various tes, temperatures in the range from l

-100 to 225*F, yielding a full Charpy V notch transition curve in both orientations (tables 3-1 and 3 2 and figures 31 and 3 2). Tests were also pcrformed on weld metal I

and HAZ metal at various temperatures from -180 to 225 F. The results are shown in tables 3-3 and 3 4 and figures 3-3 and 3-4.

l The specimens were tested on a Sontag SI-1 iriipact 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 l

Table 3-5 and figures 3-5,3 6, and 3-7 show the results of tensile tests (per ASTM E 8 l

and E 21 test criteria) from vessellower shell plate R2708-1 and from the weld metal.

Specimens from plate R2708-1 and the weldment were tested at room temperature, 300

• F, and 550

• F in both the longitudinal and transverso 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-stiain curve for each specimen. The chart recorder was calibrated to the Class B-1 extensometer. The measurement and control of speeds in the tests I

conformed to ASTM A370 68 (Mechanical Testing 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.

l9 3-1

~

c 3-3.

DROPWElGHT TESTS The nil ductility transition temperature (TNDT) was determined for plate R27081 and

~

the core region weld metal and heat-affected-zone by dropweight tests (ASTM E-208) performed at Combustbn Engineering, Inc. The following results were obtained:

Material TNDT(* F)

Plate R2708-1 O

l Weld Metal

- 50 HAZ

- 80 O

O 32

1 TABLE 31 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CALLAWAY UNIT NO.1 REACTOR l,

PRESSURE VESSEL LOWER SHELL PLATE R27081 (LONGITUDINAL ORIENTATION)

Test Temperature impact Energy Shear Lateral Expansion

('F)

(ft ib)

(%)

(mils)

- 100 2

O O

- 50 9

14 4

50 12

-3 5.5 4

O 39 23 25 0

40 30 26

^

0 41 14 24 25 39 40 27 25 41 30 29 25 58 42 37 50 52 42 38 50 70 50 46.5 50 83 66 5'

75 103 71 62.5 75 107 76 68 150 121 100 75.5 150 129 100 76 225 125.5 100 75 225 127 100 78 4

3-3 l

O

TABLE 3-2 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE R2708-1 (TRANSVERSE ORIENTATION)

Test Temperature impact Energy Shear Lateral Expansion

(* F)

(ft ib)

(%)

(mils)

- 100 6

0 0

- 50 12 5

5.5

- 50 12 14 4.5

- 10 30 18 18.5

- 10 35 10 23 0

45 25 30 0

46 34 31 0

48 25 32 40 55 37 37.5 40 55 55 41 75 56 54 41 75 72 73 51 110 75 94 53 110 80 90 59 150 95 100 62 150 114 100 68 225 98.5 100 67.5 225 107 100 67.5 e

34

1 TABLE 3-3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Test Temperature impact Energy Shear Lateral Expansion

(

• F ';

(ft Ib)

(%)

(mils)

- 150

-9 5

3.5

- 100 12 14 6

- 100 13 14 5

- 50 34 38 24

- 50 39 48 26.5

- 25 31 43 24

- 25 46 48 36 0

59 67 45 O

66 62 47.5 0

69 67 49 30 74 '

75 57 30 82 93 61 75 93 96 68 75 98.5 95 68 150 102 100 72 150 115 100 75 225 116.5 100 81 225 117 100 78 3-5 n

y __ _._ m TABLE 3-4 PRE:RRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD

{

HEAT-AFFECTED-ZONE MATERIAL l

Test Temperature impact Energy Shear Lateral Expansion

(*F)

(ft ib)

(%)

(mlis)

- 180 7

2d 3

~

- 180 36 18 16.5

- 100 25 30 15

- 00 35 18 18

- 75 29 38 19 I

- 75 34 52 26

- 50 50 30 32

- 50 66 35 36 0

67 72 48.5 0

80 58 53 30 78 82 51 30 79 92 58 75 93.5 100 62 75 100 100 61 75 104 100 73 150 116 100 70.5 150 124 100 78.5

~

210 100 100 69 O

36

TABLE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE R2708-1 AND CORE REGION WELD METAL O.2 %

Ultimate Test Yield Tensile Fracture Fracture Uniform Total' Reduction Temp. Strength Strength Load Stress Elongation Elongation in Area Material

• F (ksi)

(ksi)

(16)

(ksi)

Plate R2708-1 70 67.0 89.0 2750 184.0 16.0 26.3 69.7 (Longitudinal 70 67.0 90.0 2700 181.0 15.4 26.3 69.7 Orientation) 300 61.0 82.0 2575 175.0 14.1 24.6 70.1 300 61.0 83.0 2G75 169.0 13.2 22.7 67.9 550 59.0 87.0 2920 165.0 14.5 23.6 64.1 550 60.0 87.0 2800 155.0 14.5 23.9 63.5 Plate R2708-1 70 67.0 89.0 2875 164.0 15.4 25.2 64.4 9 (Transverse 70 68.0 89.0 3000 176.0 17.5 28.7 65.3 Orientation) 300 61.0 83.0 2850 156.0 14.0 23.6 63.0 300 61.0 83.0 3180 143.0 13.0 21.1 55.1 550 59.0 86.0 3150 125.0 13.2 20.6 48.7 550 60.0 87.0 2925 134.0 13.8 21.0 55.4 Weld Metal 70 72.0 89.0 2875 176.0 14.0 24.7 67.0 70 73.0 89.0 2905 184.0 14.0 24.4 68.0 300 66.0 82.0 2700 159.0 11.4 21.5 65.6 300 69.0 85.0 2875 166.0 11.8 21.5 65.0 550 80.0 90.0 3200 167.0 11.4 20.3 62.0 550 80.0 91.0 3155 172.0 12.4 21.6 63.0 k

14o O

gs 120 -

O 100 lii

_J p

80 E

0 o

eo O

5 O

40 Q@

20 C

I I

I I

o

-100 o

100 200 300 TEMPER ATU RE (

• F)

FIGURE 3-1 PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL PLATC R2708-1 (LONGITUDINAL ORIENTATION) 38

l

'120 O

y l

0 100 O

g 80

.J I

F 2

o l

LL w

l 60 0

0 0 E

/

l Z

tu 40

[

8 l*

2 l

20

/

O l

0 I

I I

I i

-100 0

100 200 300 1

1 TEM PER ATU RE ( ' F) l FIGURE 3-2 PREIRRADIATION CHARPY V-NOTCH IMPACT ENERGY FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE R2708-1 (TRANSVERSE ORIENTATION) 3-9 l

l

~

.c 140 O

O-

}

120 -

O 100 o

l 3

l-80

?

P 8

o b

60 d,

Z j

W 40 o/

O 20 O'

I O

-200

-100 0

100 200 300 TEM PER ATU RE (

• F)

FIGurlE 3-3 PREIRRADIATION CHARPY V-NOlCH IMPACT ENERGY FOR THE CAL!.AWAY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3 10

140 O

120 -

0 100 b

O G1 l

J j

t so 06 O

a l

W 60 l

Z l

W O

40 O

OO l-d 20 0

I I

I 0

-200

-100 0

100 200 300 TEM PER ATU RE (

• F)

FIGURE 3-4 PREIRRADIATION CHARPY V NOTCH IMPACT ENERGY FOR THE CALLAWAY UNIT NO.1 l*

REACTOR PRESSURE VESSEL CORE REGION WELD HEAT-AFFECTED-ZONE MATERIAL l

3-11 O

__ h a

100 2

ULTIMATE TENSILE STRENGTH O'

ON s

g x

80

_b 2

\\

2 y

O-

~

~

I-60

_A W

U 0.2% YlELD STRENGTH I

I 40 so 2

6

^6 60 REDUCTION IN AREA w

l-Q' 40 - 2 l-O 3

TOTAL ELONGaTICN C

0 20 O-0-

O UNIFORM ELONGATION 2

I I

I I

I o

o 100 200 300 400 500 eco TEMPERATURE (

• F)

FriURE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL PLATE R2708-1 (LONGITUDINAL ORIENTATION) 3-12

100

~

2 i

N

-2 N

-0

^

g W

ULTIMATE TENSILE STRENGTH W

IU

~

2 W

60 0.2% YlELD STRENGTH 40 80 60 m

6 O

O REDUCTION IN' AREA O

2 40 l-O TOTAL ELONGATION b

20 6-e e

UNIFORM ELONGATION l

I I

I I

n O

100 200 300 400 500 600 TEM PER ATU RE (

• F)

FIGURE 3 6 PREIRRADIATION TENSILE PROPERTIES FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSUR2 VESSEL LOWER SHELL PLATE R2708-1 (TRANSVERSE ORIENTATION) 3-13 m

100 2

ULTJ ' ATE TENSILE STRENGTH

/

=

I so 0

0N E

6~

2 g*

O.2% YlELD STRENGTH E

g 60 I

i i

i i

.o I

80 2

REDUCTION IN AREA 60 a

v

~

40 sO 2

' ' ^ ' " ' " ^ ' ' "

'3 0

20

-2

-C h-e-

e UNIFORM ELONJATION l

I I

I I

o O

100 200 300 400 500 600 TEM PER ATU RE (

• F) i FIGURE 3 7 PREIRii.31ATION TENSILE PROPERTIES FOR THE CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL 3 14

X ts e

T leisne T

ro f

ev ruCn N

ia I

A r

R t

T S

S

-ss e

r t

S lac ip y

T C.

-3 e

ru ig F

m0mG a

i cac f

SECTION 4 POSTIRRADIATION TESTING 4-1.

car $ULE REMOVAL The first capaule (Capsule U) should be removed at the end of the first core cycle (1st ref t.eling) as shown in Table 4-1. Subsequent capsules should be removed at 5, 9 and 15 EFPY (Effective Full Power Years) as indicated. Each specimen capsule,3 moved after radiation exposure will be transferred to a postirradiation test facility for dis-assembly and testing of all the specimens.

TABLE 4-1 SURVEILLANCE CAPSULE REMOVAL SCHEDULE a

0:!untation Lead Removal Expected Capsule Capsule I

Factorlbl Time Fluence (nicm2)

Identification Capsulesfal U

58.5 4.00 1st Refueling 4.3x 1018 Y

241*

3.69 5 EFPY 1.3x 1019[c]

V 61 3.69 9 EFPY 2.4 x 1019[d]

X 238.5 4.00 15EFPY 4.3 x 1019 W

121.5 4.00 Stand By Z

301.5' 4.00 Stand-By

a. Reference Irradiation Caps 0le Assembly Drawing, Figure 2-4.

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

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

d. Fluence at vesselinner wall at End of Life.

4-1

)

4-2.

CHARPY V NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the lower shell pl ate R27081, 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 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 cLmpared with preirradiation test data. The test temperature for the remaining specimens should then be adjusted l

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 representative of the upper end of the Charpy energy transition region.

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

4 4.

FRACTURE TOUGHNESS TESTS ON 1/2T COMPACT SPECIMENS a

in light of current requirements of 10CFR, Part 50, Appendix G and applications of ASME Section lil, Appendix G and Section XI, Appendix A, the 1/2 inch thick compact specimens should be tested in such a manner as to determine both static, crack ini-tiation, propagation and arrest fracture toughness parameters throughout the tempera-ture range of interest with emphasis on the sharp fracture toughness transition regions consistent with specimen availability. The specimens should thus be statically tested in accordance with ASTM E399-74 procedures modified to account for the size of the specimens available.[1] Data obtained in the sharp transition region should produce and K ; on the shelf, the values may be quite conservative. Clear values equal to Kic fa definitions of the transition regions should have first priority since other test results such l

l as Charpy shelf energy allow engineering estimates of shelf toughness values. Initiation,

(

propagation and arrest values may not be obtained from all specimens tested in this I

1 Witt. F J Fracture Toughness Parameters Cttaced from S<ngie Smat1 Specimen Tests

..C!.D-9397. October 1973 4-2

'. O region in contrast to the shelf region. Specific test procedures should include unloading compliance and data interpretation should utilize the Equivalent Energy and J Integral and dJ/da cr concepts.[1,2A41 Fracture toughness data so obtained will be K, K a, Jic ic l

engineering estimates thereof. Advantages should be taken of the Charpy impact and tensile data in the selection 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 foliows:

5 Milling machine or special cutoff wheel for opening capsules, dosimeter blocks and spacers B Hot cell tensile testing machine with pin type adapter for testing tensile specimens.

5 Hot cell static CT testing machine with clevis and appropriate measuring equipment modified to account for the size of the specimens E Hot cell Charpy impact testing machine E Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters.

1. Witt. r J ' Fracture f oughness Parameters obtamed trom Single sm.ot Specimen Te.ts" WCAP.939 7. octoter 19 78 2 Duchaiet. C and Mager. T. R. Emper rrental VerAcation of Lowet Sound Kg Values Utd+r.o the Eauhaient Enerm Concect ^c hopress en flaw Growth and Fracture Tnqqhness Testing. ASTM STP 530 pp 281 200. American Soc.ety for Testing and Materials. Phdadelphra.1973 e

3 Landes. J D and Beg!cv J A. Recent Deweicon ents in J Test ng ', in Devecoments e Tractu e ecr'a" es Test 48cthods r

Standard ration ASTM STP 632. op 57 81. Americar. %cie.y for Testing and MateN!s Phdadeich a t 97 7 4 McCaba. O E 'Determ.nat cn of R Curves for Structu a: Mater.ais Ustng Naniencar Mechanics Ut.inods

.n Flaw Greath and r

Fractge ASTM STP 631 pp 245-2t'6 Amencan Socie'y for Test.ng and Materiais Phdadeicht.1 1977 43

o APPENDIX A CALLAWAY UNIT NO.1 REACTOR PRESSURE VESSEL SURVE!LLANCE MATERIAL For the reactor vessel radiation surveillance program, Combustion Engineering, Inc.,

supplied Westinghouse with sections of A533 Gr B CL 1 plate used in the core region of the Callaway Unit 1 reactor pressure vessel, specifically, from the 9 5/8-inch lower shell plate R2708-1 of the pressure vessel. Also supplied was a weldment made from sections of lower shell plate R2708 1 and adjoining iptermediate shell plate R2707-1.

t This weldment was fabricated using 3/16 inch Mil B-4 weld filter wire, heat number 90077 and Linde 124 flux, lot number 1061 and is identical to that used in the actual fabrication of the reactor vessel. The plates were produced by Lukens Steel Company.

l*

The heat treatment history and the chemical analysis of the pressure vessel surveillance materials are shown in tables A-1 and A-2, respectively.

i l

TABLE A-1 HEAT TREATMENT HISTORY I

Tem erature ime Material Cooling p

Austenitizing:

4 Water-quenched 1600 25 Lower Shell Tempered:

cooled Plate R2708-1 1225 25 ress Relief:

St 1150 + 50 Weldment 1150 t 50 7.75 Furnace-cooled A-1

c TABLE A 2 CHEMICAL ANALYSIS OF MATERIALS Content in Indicated Material (weight %)

Element Lower Shell Weld Plate R2708-1[a]

Metal [a]

~

C

.22

.15 Mn 1.47 1.37 P

.006

.005 S

.014

.008 Si

.25

.44 Ni

.59

.07 Mo

.57

.54 Cr

.05

.04 Cu

.07

.06 Al

.025

.003 Co

.013

.011 Pb

[b]

<.001 W

<.01

<.01 Ti

<.01

<.01 Zr

<.001

< 001 V

.003

.004 Sn

.002

.003 As

.001

<.001 Cb

<.01

<.01 N2

.008

.007 Sb

.0015 B

<.001

.001

a. Analysis conducted by Combustion Engineering,Inc.

b. Not detected,

.e A2 l

., _.,