ML17329A567

American Electric Power Co DC Cook Unit 2 Reactor Vessel Radiation Surveillance Program.

ML17329A567
Person / Time
Site:	Cook
Issue date:	11/30/1975
From:	Chirigos J, Davidson J, Yaninchko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17329A566	List: ML17329A565 ML17329A567 ML17334B426
References
WCAP-8512, NUDOCS 9207280261
Download: ML17329A567 (38)

Text

WESTINGHOUSE CLASS 3 AMERICAN ELECTRIC POWER COMPANY DONALD C. COOK UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J. A. Davidson S. E. Yanichko J. H. Phillips November 1975 APPROVED:

J. N. Chirigos, Manager Structural Materials Engineering Wor o Order No. AMP-106 j

pER~Q HRC DAT- 0 24 0 92072802bi 920713 Nuclear Energy Systems PDR ADOCK 050003i5 P. O. Box 355 P . PDR Pittsburgh, Penns Ivan'

ABSTRACT A pressure vessel s teel ope for surveillance program was developed or tthee Ameri American Electric Power Company, Donald C. C oo k U nit No. 2, to obtain information on the effects of radiation e reactor vessel material under operating conditions. The pro ram corn e ects ased on comparison preirradiation testin of a ermine toughness properties of the reactor pressure vessel. ontinuous e se . Con re'ug ens witt in t e reactor res sure vessel provides data on the inteegririty oof th e vessel in terms of adequate toughness ro ertie A description of the sur-ness properties.

veillance capsules and preirradiation test results is also included apsu es an

TABLE OF CONTENTS Section Title Page PURPOSE AND SCOPE SAMP LE P R EPA RATI ON 2-1 2-1. Pressure Vessel Material 2-1 2-2. Machining 2-1 2-3. Charpy V-Notch Impact Specimens 2-1 2-4. Tensile Specimens 2-1 2-5. Wedge Opening Loading Specimens 2-4 2-6. Monitors 2-4 2-7. Dosimeters 2-8. Thermal Monitors 2-4 2-9. Surveillance Capsules 2-4 2-10. Capsule Preparation 2-4 2-11. Capsule Loading 2-4

'PREIRRADIATION TESTING 3-1 3-1 ~ Charpy V-Notch Impact Tests 3-1 3-2. Tensile Tests 3-1 3-3. Dropweight Tests 3-14 POSTI R RADIATION TESTING 4-1 4-1. Capsule Removal 4-1 4-2. Charpy V-Notch Impact Tests 4-2 4-3. Tensile Tests 4-2 4.4, Wedge Opening Loading Kid Fracture Toughness Tests 4-2 4-5. Postirradiation Test Equipment 4-2 APPENDIX A DONALD C. COOK UNIT NO. 2 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A-1

LIST OF ILLUSTRATIONS Fig'ure Title Page 2-1 Charpy V-Notch Impact Specimen 2-2 2-2 Tensile Specimen 2-3 2-3 Wedge Opening Loading Specimen 2-5 2-4 Irradiation Capsule Assembly 2-6 2-5 Dosimeter Block Assembly 2-8 2-6 Location of Specimens in the Reactor Surveillance Test Capsules 2-9 3-1 Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Inter-mediate Shell Course Plate C5521-2 (Longitudinal Orientation) 3-4 3-2 Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Inter-mediate Shell Course Plate C5521-2 (Transverse Orientation) 3'-5 3-3 Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-8 3-4 Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat-Affected Zone Material 3-9 3-5 Preirradiation Tensile Properties for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Course Plate C5521-2 (Transverse Orientation) 3-11 3-6 Preirradiation Tensile Properties for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-12 3-7 Typical Tensile Test Stress-Strain Curve 3-13

LIST OF TABLES Table Title Page 2-1 Type and Number of Specimens in the Donald C. Cook Unit No. 2 Surveillance Test Capsules 2-7 2-2 Quantity of Isotopes Contained in the Dosimeter Blocks 2-?

3-1 Preirradiation Charpy V-Notch Impact Data for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 3-2 3-2 Preirradiation Charpy V-Notch Impact Data for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 3-3 3-3 Preirradiation Charpy V-Notch Impact Data for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-6 3-4 Preirradiation Charpy V-Notch Impact Data for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat-Affected 2one Material 3-7 3-5 Preirradiation Tensile Properties for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate C5521-2 and Core Region Weld Metal 3-10 4-1 Schedule for Removal of Specimen Capsules 4-1 A-1 Heat Treatment History A-1 A-2 Quantitative Chemical Analysis (Weight-Percent) A-2

SECTION 1 PURPOSE AND SCOPE The purpose of the American Electric Power Company, Donald C. Cook Unit No. 2, surveil-lance program is to obtain information on the effects of radiation on the reactor vessel materials of the reactor during normal operating conditions. Surveillance material is selected as the most limiting material based on surveillance selection procedures which are outlined in ASTM E185-73, Annex A1. Evaluation of the radiation effects is based on the preirrad-iation testing of Charpy V-notch, tensile, and dropweight specimens, and postirradiation testing of Charpy V-notch, tensile, and wedge-opening-loading (WOL) specimens.

Current reactor pressure vessel material test requirements and acceptance standards use the reference nil-ductility temperature, RTNDT, as a basis. RTNDT is determined from the drop-weight nil-ductility transition temperature, NDTT, or the weak (transverse oriented) direction 50 ft Ib Charpy V-notch impact temperature (which ever value is greater) as defined by the following equation:

RTNDT = NDTT, if NDTT > T50(35) 60'F OI'TNDT =

T50(35) 60'F, if T50(35) 60'F ~ NDTT where RTNDT = Reference nil.ductility temperature NDTT = Nil-ductility transition temperature as per ASTM E208 T50(35) 50 ft Ib temperature from transverse oriented Charpy V-notch impact specimens (or the 35 mil-temperature, if it is greater(" j )

1. In the case where at least 35 mils lateral expansion is not obtained at the 50 ft Ib temperature, the temperature at which 35 mils lateral expansion occurs is used.

1-1

An empirical relationship between RTNDT and fracture toughness for reactor vessel steels has been developed and is presented in appendix G, section III, of the ASME Boiler and Pressure Vessel Code (Protection Against Non-Ductile Failure). The relationship can be em-ployed to set allowable pressure-temperature relationships, based on fracture mechanics con-cepts, fo'r the normal "operation of reactors. Appendix G of the ASME Boiler and Pressure Vessel Code defines an acceptable method for calculating these limitations.

It is known that radiation can shift the Charpy impact energy curve to higher temperatures.~ '

Thus, the 50 ft-Ib temperature, and correspondingly, the RTNDT, increase with radiation exposure. The extent of the shift in the impact energy curve that is, the radiation em-brittlement is enhanced by certain chemical elements, such as copper, present in reactor vessel steels. ~ ~ )

The 50 ft Ib temperature, and correspondingly the RTNDT, increase with service and can be monitored by a surveillance program which consists of periodically checking irradiated reactor vessel surveillance specimens. The surveillance program is based on ASTM E185-73 (Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels). WOL fracture mechanics specimens are used in addition to the Charpy impact specimens to evaluate the effects of radiation on the fracture toughness of the reactor vessel materials. ( ~ r r ~ ~" ~ "j

1. L. F. Porter, "Radiation Effects in Steel," in Materials in lyuclear Applications, ASTM STP.276, pp. 147.195, American Society for Testing and Materials, Philadelphia, 1960.

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

3. U. Potapovs and J. R. Hawthorne, "The Eftcct ot Residual Element on 550 F Irradiation Response of Selected Pressure Vessel Steels and Weldmcnts," NRL-6803, September 1968.

4. I.. E. Steels, "Structure and Composition Effects on Irradiation Sensitivity of Pressure Vessel Steels," in irradiation Effects on Structural Alloys for Nuclear Reacror Applicarions, ASTM STP484, pp. 164-175, American Society for Testing and Materials, Philadelphia, 1970.

5. E. Landerman, S. E. Yanichko, and W. S. Haze!ton, "An Evaluation of Radiation Damage to Reactor Vessel Steels Using Both Transition Temperature and Fracture Mechanics Approaches," in The Effects of Radiarion on Structural Meta/s, ASTM STP<26, pp. 260277, American Society for Testing and Materials, Philadelphia, 1967.

6. M. J. Msnjoine, "Biaxial Brittle Fracture Tests," Trans. Am. Soc. Mech. Engrs. 87, Series D, 293.298 (1965).

7. L. Porse, "Reactor-Vessel Design Considering Radiation Eftects," Trans. Am. Soc. Mech. Engrs. 86, Series D,

743.749 f1964).

8. R. E. Johnson, "Fracture Mechanics: A Basis for Brittle Fracture Prevention," WAPD-TM-505, November 1965.

9. E. T. Wessel and W. H. Pryle, "Investigation ot the Applicability of the Biaxial Brittle Fracture Test tor Determining Fracture Toughness," WERL.8844.11, August 1965.

10. W. K. Wilson, "Analytic Determination of Stress Intensity Factors for the Manjoine Brittle Fracture Test Specimen,"

WERL.0029-3, August 1965.

11. R. E. Johnson and E. J. Pasierb, "Fracture Toughness of Irradiated A302.8 Steel as Influenced by Microstructure,"

Trans. Amer. Nucl. Soc. 9, 390.392 f19661.

1-2

Postirradiation testing of the Charpy impact specimens provides a guide for determining pressure-temperature limits on the plant. A temperature shift in the reference temperature will occur in the irradiated Charpy impact specimen test data as a result of radiation exposure at plant temperatures. These data can then be reviewed to verify or establish new pressure-temperature limits of the vessel during start-up and cooldown. This allows a check of the predicted shift in the reference temperature. The postirradiation test results on the WOL specimens provide actual fracture toughness properties for the vessel material. These properties may be used for subsequent evaluation as per the methods outlined in the ASME Code, appendix G.

Eight material test capsules are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are located in guide tubes attached to the thermal shield. The capsules contain Charpy impact, WOL, and tensile specimens from the limiting core region plate. This plate is the reactor vessel inter-mediate shell plate adjacent to the core region and is 8 3/4 inches thick. Charpy impact, WOL, and tensile specimens obtained from the representative core region weld metal, and Charpy impact specimens from the weld material heat-affected zone (HAZ), are also located in the capsules. In addition, dosimeters to measure the integrated neutron flux and thermal monitors to measure temperature are located in each of the eight material test capsules.

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

1-3

SECTiON 2 SAMPLE PREPARATiON 2-1. PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by The Chicago Bridge and Iron Company from the vessel intermediate shell plate C5521-2. A submerged arc weldment which joined sections of material from this plate and lower shell plate C5592-1 was also supplied by The Chicago Bridge and Iron Company. Data on the pressure vessel material are presented in appendix A.

2-2. MACH IN I NG Test material was obtained from the intermediate shell course plate when the thermal heat treatment was complete and the plate formed. All test specimens were machined from the 1/4-thickness section of the plate after a simulated postweld stress-relieving treatment on the test material was performed. The test specimens represent material taken at least one plate thickness (8 3/4 inches) from the quenched ends of the plate. Specimens were machined from weld and heat-affected zone (HAZ) material of a stress-relieved weldment which joined sections of the intermediate and lower shell plates. All HAZ specimens were obtained from the weld HAZ of intermediate shell plate C5521-2 2-3. Charpy V-Notch Impact Specimens (Figure 2-1)

Charpy V-notch impact specimens from intermediate shell plate C5521-2 were machined in both the longitudinal orientation (longitudinal axis of specimen parallel to major working direction) and transverse orientation (longitudinal axis of specimen perpendicular to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of the Charpy was normal to the weld direction; the notch was machined such that the direction of crack propagation in the specimen was in the weld direction.

2-4. Tensile Specimens (Figure 2-2)

Tensile specimens were machined with the longitudinal axis of the specimen perpendicular to the major working direction of the plate.

2-1

O.OI IR

0. 009 90'n50'.395 0.393 89 0 395

0. 393 I. 063 I . 053

2. I25

2. I05 ALL OVER UNLESS OTHERWISE SPECIF IED Figure 2-1. Charpy V-Notch Impact Specimen

GAGE LENGTH 0.995 0.25I DIA DIA-.B. 0. 2iI9"A" 0.395 0 IA "B" 16 0.393 l I I

NOTE:

I I I I "B" OIA IS TO BE ACTUAL "A" OIA +0.002 i0 0.005

0. 198 "A" 0.250 R" B" TAPERING TO AT THE CENTER 0.255 TYP 0. I 97 NOTES:

l.250 REDUCED I. 495 I 260 SECTION I .480 I. LATHE CENTERS REQUIRED Ij. 250 2- ~ALL OVER UNLESS OTHERWISE SPEC IF IEO 4.2 IO 0.630 "B" BLEND LINE FOR R 0 620 I6 A I6 0.790 0 786 0 395 0 375 SECT I OH A-A 0 377

" ( )

rL OF HOLES TO BE WITHIN 0.002 OF TRUE OF SPECIHEN g

Figure 2-2. Tensile Specimen

2-5. Wedge Opening Loading Specimens (Figure 2-3)

Wedge opening loading (WOL) test specimens were machined along the transverse orientation so that the specimen would be loaded perpendicular to the major working direction of the plate and the simulated crack would propagate along the longitudinal direction. All specimens were fatigue precracked according to ASTM E399-70T.

2.6. MONITORS 2-7. Dosimeters Eight capsules of the type shown in figure 2-4 contain dosimeters of copper, iron, nickel, and aluminum-cobalt wire (cadmium-shielded and unshielded), neptunium-237, and uranium-238.

The dosimeters are used to measure the integrated flux at specific neutron energy levels.

2-8. Thermal Monitors The capsules contain two low-melting-point eutectic alloys so that the maximum temperature attained by the test specimens during irradiation can be accurately determined. The thermal monitors are sealed in Pyrex tubes and then inserted in spacers (figure 2-4). The two eutectic alloys and their melting points are as follows:

2.5% Ag, 97 5% Pb Melting point 579'F 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point 590'F 2-9. SURVEILLANCE CAPSULES 2-10. Capsule Preparation The specimens were seal-welded into a square austenitic stainless steel capsule to prevent corrosion of specimen surfaces during irradiation. The capsules were then hydrostatically tested in demineralized water to collapse the capsule on the specimens so that optimum thermal conductivity between the specimens and the reactor coolant could be obtained. The capsules were helium-leak tested as a final inspection procedure. Finally, the capsules were coded S, T, U, V, W, X, Y, and Z. Fabrication details and testing procedures are listed in the notes in figure 2-4.

2-'I1. Capsule Loading Upon receipt, the eight test capsules are positioned in the reactor between the thermal shield and the vessel wall at the locations shown in figure 2-4. Each capsule contains 44 Charpy V-notch specimens, 4 tensile specimens, and 4 WOL specimens.

2-4

I' ALL OVER

2. NOTCH DEPTH TO BE EXTEHDED BY 0.09-0.I56 BY FATIGUE CRACKING I.I30'OTES: 3. SLOf 10 4.

BE PARALLEL TO SURI SLOT TO BE PERPENDICULAR fo ACf'A" WITHIN 0. 005 SURFACES "8" 6 "C" 101AL 1.45 WITHIN 0.005 I.43 5. DIM.S MARKED THUS 'UST HOT HAVE A T. R.O. EXCFEDING

+ O.OOI OH OPPOSITE fACES

0. 375 D IA. I. I20 I . 005 THRU

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0. 995 O.I27 . I27 SEE NOTE 2 0. I23 0. I 23 0.279

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0. 00 I 0.50I 0.500-20 THD CLASS "3B" 0.499 DEEP (l4)

DIM."X" DIM."Y" 0.375 DEEP 0 I23 0.0463 0.023I 0.0473 0.0236 SPECIMEH IDEHTIF ICATIOH (2) LOCATIOHS Figure 2-3. Wedge Opening Loading Specimen

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b t>>W>>>>t >>CCCV>>>>V Figure 2-4. Irradiation Capsule Assembly 2>>6

The relationship of the test material to the type and number of specimens in each capsule is shown in table 2-1.

Dosimeters of pure iron, nickel, and copper, aluminum-0.15 percent cobalt, and cadmium-shielded aluminum-0.15 percent cobalt, wires are secured in holes drilled in spacers located in the capsule positions shown in figure 2-4. Each capsule also contains a dosimeter block (figure 2-5) which is located at the center of the capsule. Two cadmium-oxide shielded cap-sules, each containing isotopes of either U or Np (both 99.9 percent pure) 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 activation products. The amounts of each are presented in table 2-2. Both of them are held in a 3/8-inch long by 1/4-inch OD sealed brass tube and stainless steel tube, respectively. Each tube is placed in a 1/2-inch-diameter hole in the dosimeter block (one U and one Np tube per block), and the space around the tube filled with cadmium oxide. After placement of this material each hole is blocked with two 1/16-inch-thick aluminum spacer discs and an outer 1/8-inch-thick steel cover disc welded in place.

The numbering system for the capsule specimens and their locations is shown in figure 2-6.

TABLE 2-1 TYPE AND NUMBER OF SPECIMENS IN THE DONALD C. COOK UNIT NO. 2 SURVEILLANCE TEST CAPSULES Capsule S,V,W, and.X Capsule T,U,Y, and Z Material Charpy Tensile WOL Charpy Tensile WOL Plate C5521-2 (longitudinal)

Plate C5521-2 (transverse) 12 12 Weld Metal 12 12 HAZ 12 12 TABLE 2-2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS Weight Weight Isotope (mg) Compound (mg) 237 12+ 20+1 1 Np02 U238 12 14.25 U308 2-7

MATERIAL NO.

I TEM Tl TLE SPECIFICATION RE/'.

I BLOCK

/ I COVER I

I SPACER NEPTUNIUM SEALED CAPSULE STAINLESS (0. 250 OD x 0.375 LG) STEEL URANIUM23B SEALED CAI SULE BRASS (0.250 OD x 0.375 LG)

CADMIUM OXIDE AS RE '0 TYP.

0.06 CJJ CJl CJl CJl I

CJJ I

Figure 2-5. Oosimeter Block Assembly

Q1 RS SR 8$ RR RRERRRRR RR RR SIR SE RE SIR RW RR RR RE ER RR RR RR RR QR RR RR RR 5$ RE D

RR RW RW RR RR ES RR RE KRl IRK RR RK RR RR RR RR RR 1IE SIR EE RR RR RR RR RR RIW RK RR

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SECTION 3 PREIRRADIATION TESTING 3-1. CHARPY V-NOTCH IMPACT TESTS Charpy V-notch impact tests were performed on the vessel intermediate shell plate C5521-2, at various temperatures from -50'o 210'F to obtain a full Charpy V-notch transition curve in both the longitudinal and transverse orientations (tables 3-1 and 3-2, and figures 3-1 and 3-2). Charpy impact tests were performed on weld metal and HAZ material at various tempera-tures from -100'o 300'F. The results are reported in tables 3-3 and 3-4 and figures 3-3 and 3-4, respectively.

The Charpy impact specimens were tested on a Sontag Sl-1 impact machine which is inspected and calibrated every 12 months using Charpy V-notch impact specimens of known energy values. These impact specimens are supplied by the Watertown Arsenal.

3-2, TENSILE TESTS Tensile tests were performed on the vessel intermediate shell plate C5521-2 (in the transverse orientation) and the weld metal at room temperature, 300'F, and 550'F. The results are shown in table 3-5 and figures 3-5 and 3-6.

Tensile tests for the intermediate shell plate and weld metal were performed on an Instron TT-C tensile testing machine using the standard Instron gripping devices. A full stress-strain curve was obtained for each specimen using a Baldwin-Lima-Hamilton Class B-1 extensometer and chart recorder, the latter calibrated to the extensometer. The method of'measuring and controlling speeds for tensile tests on the Instron TT-C are governed by ASTM A370.68 (Mechanical Testing of Steel Products).

The lnstron TT-C tensile testing machine 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 3-7.

3-1

TABLE 3-1 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE DONALD C. COOK NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE C5521-2 (LONGITUDINALORIENTATION)

Test Lateral Temp Energy Shear Expansion

('F) (ft Ib) (%) (mils) 15 18 12 18 '19 13 15 18 25 26 30 20 25 31 25 23 25 25 29 50 52 35 38 50 47 35 37 50 46 35 70 65 47 70 42 49 70 76 55 100 91 65 66 100 98 70 76 100 90 62 67 125 126 85 78 125 114 77 79 125 103 70 210 122 100 83 210 132 100 86 210 128 100 84 3-2

TABLE 3-2 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE DONALD C. COOK NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE C5521-2 (TRANSVERSE ORIENTATION)

Test Lateral Temp Energy Shear Expansion

('F) (ft Ib) (i) (mils)

-50 5.5

-50 6.0

-50 6.0 10 39.0 29 27 10 29.0 25 17 10 25.0 30 18 70 43.0 40 32 70 42.0 43 33 70 39.0 28 100 66.0 60 100 71.5 63 53 100 68.0 65 49 120 67.5 58 56 120 76.0 65 60 120 75.0 72 59 210 81.0 100 210 88.0 100 63 210 90.0 100 66 3-3

8556-22 l40 l30 0

I20 0 IIO IOO 90 80 70 60 50 40 0

30 20 IO

-300 -200 -IOO 0 IOO 200 300 TEMPERATURE ( F)

Figure 3-1. Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 3-4

8556-17 IOO 90 80 70 0 0

60 I

50

%0 0 0 30 0 0

20 IO 0

-IOO IOO 200 300 TEMPERATURE (oF)

Figure 3-2. Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Reactor Pressure Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 3-5

TABLE 3-3 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE DONALD C. COOK NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Test Impact Lateral Temp Energy Shear Expansion

('F) (ft Ib) (%) (mils)

-25 20.0 48 20

-25 22.0 30 17

-25 31.5 40 25 20 32.0 38 20 35.0 47 28 20 33.0 50 27 60 58.0 74 48 60 47.0 65 37 60 39.0 50 29 100 74.0 95 63 100

'6.0 85 47 100 65.0 95 53 210 72.0 100 210 70.0 100 63 210 77.0 100 300 72.0 98 ~

66 300 79.0 100 71 300 81.0 100 70 3-6

TABLE 34 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE DONALD C. COOK NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT-AFFECTED ZONE MATERIAL Test Impact Lateral Temp Energy Shear Expansion

('F) (ft Ib) (%)

-100 21.0 30 (mils)'50 100 5.0 12

-100 14.0 29 34.0 16

-50 23.0 27 21

-50 70.5 53 39

-25 89.0 65 52 25 70.0 60 43

-25 90.0 60 52 95.0 70 59 76.0 65 52 130.0 100 75 50 84.0 90 55 50 67.0 85 48 50 136.0 100 76 125 95.0 66 125 104.0 99 75 125 82.0 90 71 210 147.0 100 77 210 113.0 100 80 210 86.0 100 3-7

8556-IS 90 80 0 0 70 0 0 50 0-40 30 0 20 IO

-100 l00 200 300 400 TEMPERATURE (oF Figure 3-3. Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-8

I60 l50 0

ISO l30 0 I20 IIO 0 IOO 0

0 90 80 0 0 70 0 0 60 50 40 30 20 0 0 IO

-300 -200 -IOO 0 IOO 200 300 TEMPERATURE ( F)

Figure 3-4. Preirradiation Charpy V-Notch Impact Energy Curve for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat-Affected Zone Material 3-9

TABLE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE DONALD C. COOK UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE C5521-2 AND CORE REGION WELD METAL Ultimate Test 0.2% Yield Tensile Fracture Fracture Uniform Total Reduction Temp Strength Strength Load Stress Elongation Elongation In Area Vessel Material ('F) (psi) (psi) (Ib) (psi) (0/) (0/) ('/)

Plate C5521-2 ROOM 67400 87350 3200 161200 13.4 23.4 59.6 (Transverse Orien- ROOM 65450 85900 2950 156400 15.0 27.1 61.7 tation) 300 58800 78600 2650 146100 13.0 22.6 63.1 300 60500 79500 2675 157600 10.6 19.8 65.4 550 57500 83000 3225 142150 11.5 19.0 53.8 550 58950 83150 3150 145600 12.7 20.5 56.0 Weld Metal ROOM 75750 93250 2850 173400 13.9 25.7 66.8 ROOM 76900 91300 2950 178800 12.2 22.6 66.6 300 70750 88000 2900 171000 10.7 20.7 66.0 300 71000 85350 2875 179000 10.3 21.2 67.5 550 70000 87250 3160 157200 10.1 19.2 '9.6 550 68200 87800 3050 166000 9.3 20.2 62.8

8556-20 l 00 90

~ So CO CD ULTIMATE TEHSILE STREHGTN 70 60 O. 2g YIELD STRENGTH 50

%0 80 70 60 cv 50 REDUCTION IN AREA o- 00 30 0 TOTAL ELONGATION C5 0 20 0 UNIFORM ELONGATIOH l0 loo 200 300 %00 500 600 700 EgpEgATuliE (oF)

Figure 3.5. Preirradiation Tensile Properties for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Intermediate Shell Course Plate C5521-2 (Transverse Orientation) 3-11

8556>>2I lpp 90 80 ULTIMATE TENSILE STRENGTH C) 70 0.2% YIELD STREHGTH 60 50 40 80 70 60 I

REDUCTION IK AREA 50 40 I

30 TOTAL ELOHGATIOH 20 UNIFORM ELONGATIOH Ip lpp 200 300 400 500 600 700 TEMPERATURE ( F)

Figure 3-6. Preirradiation Tensile Properties for the Donald C. Cook Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-12

STRAIN Figure 3-7. Typical Tensile Test Stress-Strain Curve 3-13

3-3. DROPWEIGHT TESTS The NDTT was determined for plate C5521-2, the core region weld metal, and HAZ material by dropweight tests (ASTM E-208) performed at The Chicago Bridge and Iron Company. The following results were obtained:

Material NDTT ('F)

Plate C5521-2 +10 Weld Metal -40 HAZ -10 3-14'

SECTION 4 POSTIRRADIATION 4-1. CAPSULE REMOVAL Specimen capsules are removed from the reactor only during normal refueling periods.

The recommended schedule for removal of capsules is presented in table 4-1.

TABLE 4-1 SCHEDULE FOR REMOVAL OF SPECIMEN CAPSULES Multiplying Factor By Capsule Which the Capsule Leads Identification Vessel Maximum Exposure Removal Time 2.9 End of first core cycle 2.9 9 years 2.9 18 years 2.9 30 years 1.0 Standby V 1.0 Standby W 1.0 Standby 1.0 Standby Each specimen capsule is removed after radiation exposure and transfered to a post-irradiation test facility for disassembly of the capsule and testing of all specimens within that capsule.

4-1

4-2. CHARPY V-NOTCH IIVIPACT TESTS The testing of the Charpy impact specimens from the intermediate shell course plate, the weld metal, and HAZ material in each capsule can be done singly at approximately five different temperatures. The extra specimens can be used to run duplicate tests at test temperatures of interest.

The initial Charpy specimen from the first capsule removed should be tested at room temperature. The impact energy value for this temperature should be compared with the preirradiation test data; the testing temperatures for the remaining specimens should then be raised and lowered as needed. The test temperatures of specimens from capsules exposed to longer irradiation periods should be determined by the test results for the previous capsule.

4-3. TENSILE TESTS The tensile specimens for each of the irradiated materials should be tested at test temperatures identical to the WOL fracture toughness test temperatures of the material.

'4-4. WEDGE OPENING LOADING Kid FRACTURE TOUGHNESS TESTS

. In light of current requirements of 10CFR, Part 50, ASME Code, appendix G, the WOL specimens should be tested dynamically to adequately characterize the fracture toughness properties of the reactor vessel. The WOL specimens for each of the irradiated materials should be tested in accordance with ASTM E399-70T with appropriate modifications necessary for dynamic tests. Test temperatures which are recommended are the irradiated 50 ft Ib temperature, 212'F, and temperatures representative of the irradiated Charpy V-notch upper shelf region if the 212'F test temperature occurs in the transition region. When the material fracture toughness at these temperatures is too high to be valid according to ASTM E399-70T, test data can then be interpreted by either the J Integral Concept~ j or the Equivalent Energy Concept~ j.

4-5. POSTIRRADIATION TEST EQUIPMENT The following minimum equipment is required for the postirradiation testing operations.

Milling machine or special cutoff wheel for opening capsules, and dosimeter blocks and spacers

~

~

1. Fracture Toughness, ASTM STP-514, American Society for Testing and Materials, Philadelphia, 1972.

2. T. R. Mager and C. Buchalet, "Experimental Verification of Lower Bound K Values Utilizing the Equivalent Ic Energy Concept, I in Progress in Flaw Growrh and Fracrure Toughness Tesring,

~

ASTM STP 536, pp. 281.296, American Society for Testing and Materials, Philadelphia, 1973.

4-2

~ Hot cell tensile testing machine with pin-type adapter for testing tensile specimens

~ Hot cell dynamic WOL testing machine with clevis and appropriate displacement measuring equipment associated with dynamic testing Hot cell Charpy impact testing machine

~ Sodium iodide scintillation detector and pulse height analyzer for gamma counting of the specific activities of the dosimeters 4-3

i' APPENDIX A DONALD C. COOK UNIT NO. 2 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL The Chicago Bridge and Iron Company supplied the Westinghouse Electric Corporation with sections of A533 Grade B, Class 1 plate used in the core region of the Donald C. Cook Unit No. 2 reactor pressure vessel for the Reactor Vessel Radiation Surveillance Program.

The sections of material were removed from the 8 3/4-inch intermediate shell plate C5521-2 of the pressure ve'ssel. The Chicago Bridge and Iron Company also supplied a weldment made from sections of plate C5521-2 and adjoining lower shell plate C5592-1 using weld wire representative of that used in the original fabrication. The heat treatment history and quantitative chemical analysis of the pressure vessel surveillance material are presented in tables A-1 and A-2, respectively.

TABLE A-1 HEAT TREATMENT HISTORY Material Temperature Time

('F) (hrs) Coolant Intermediate Shell 1650/1750 4 1/2 Water (Plate C5521-2) quenched 1550/1650 Water quenched 1200/1300 4 1/2 Air cool 1150 + 25 51 1/2 Furnace cooled Weld 1140 + 25 Furnace cooled

D