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Reactor Vessel Matl Surveillance Program for DC Cook,Unit 2,Analysis of Capsule Y
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{{#Wiki_filter:SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR DONALD C. COOK UNIT NO. 2 ANALYSIS OF CAPSULE Y l by E. B. Norris ](.'.- FINAL REPORT y SwRI Project No. 06-7244 002, J h:r pyy for indiana & Michigan Electric Company Donald C. Cook Nuclear Plant Bridgman, Michigan 49106 February 1984 Approved: / }{ U. S. Lindholm, Director Department of Materials Sciences l}gg6 ]@gM @3o 5 P

ABSTRACT Capsulo Y the second vossol material surveillanco capsulo removed f rom the Donald C. Cook Unit No. 2 nuclear power plant has boon tostod, and the results havo boon ovaluated. The analysis of the data indicates that the pressuro material will retain adequate shelf toughness throughout the 32 EFPY design lifetime. Heatup and cooldown limit curves for normal operation have been developed for up to 12 and 32 offective full power years of opera-tion. ii

TADLE OF CONTENTS .P,,a.gg LIST OF FIGURES iv LIST OF TABLES v I.

SUMMARY

OF RESULTS #40 CONCLUSIONS 1 II. DACKGROUND 3 III. DESCRIPTION OF MATERIAL SURVEILLN4CE PROGRAM 6 IV. TESTING OF SPECIMENS FROM CAPSULE Y 12 A. Shipment Opening, and Inspection of Capsulo 12 D. Noutron Oosimetry 13 C. Mechanical Property Tests 21 V.

1ALYSIS OF RESULTS 31 VI.

HEATUP M40 COOLDOWN LIMIT CURVES FOR NORMAL 38 OPERATI0tl 0F 00NALD C. COOK UNIT NO. 2 VII. REFERENCES 45 TENSILE TEST RECORDS APPENDIX A PROCEDURE FOR THE GENERATION OF ALLOWADLE N)PENDIX 0 PRESSURE-TEMPERATURE LIMIT CURVES FOR NUCLEAR POWER PLNIT REACTOR VESSEL iii

LIST OF FIGURES Figure Page 1 Arrangement of Surveillance Capsules in the Pressure 7 Vessel 2 Vessel Material Surveillance Specimens 10 .3 Arrangement of Specimens in Capsule Y 11 4 Radiation Response of Donald C. Cook Unit No. 2 Vessel 24 Shell Plate C5521-2 (Longitudinal Orientation) 5 Radiation Response of Donald C. Cook Unit No. 2 Vessel 25 Shell Plats C5521-2 (Transverse Orientation) 6 Radiation Response of Donald C. Cook Unit No. 2 Reactor 26 Vessel Heat-Affected Zone Material 17 Radiation Response of Donald C. Cook Unit No. 2 Reactor 27 'm-Vessel Weld Material 8 Effect of Neutron Fluence on RTNDT Shift, Donald C. Cook 33 ~ Unit No. 2 9, Dependence of Cv Upper Shelf Energy on Neutron Fluence, 36 Donald C. Cook Unit No. 2 10 - Reactor Coolant System Pressure-Temperature Limits 41 Versus 100*F/ Hour Rate Criticality Limit and Hydro-static Test Limit, 12 EFPY 11 Reactor Coolant System Pressure-Temperature Limits 42 Versus Cooldown Rates, 12 EFPY 12 Reactor Coolant System Pressure-Temperature Limits 43 Versus 100*F/ Hour Rate Criticality Limit and Hydro-static Test Limit, 32 EFPY 13 Reactor Coolant System Pressure-Temperature Limits 44 Versus Cooldown Rates, 32 EFPY l iv l t

LIST OF TABLES Table Page I Donald C. Cook Unit No. 2 Reactor Vessel Sur-8 ve111ance Materials [123 II Summary of Reactor Operations, Donald C. Cook 15 Unit No. 2 III Results of Discrete Ordinates Sn Transport 17 Analysis, Donald C. Cook Unit No. 2, 40' Capsules IV Summary of Neutron Dosimetry Results, Donald C. 19 Cook Unit No. 2, Capsule Y V Charpy Impact Properties of Longitudinal Plate 22 Donald C. Cook Unit No. 2, Capsule Y VI Charpy Impact Properties of Transverse Plate 22 Donald C. Cook Unit 2, Capsule Y VII Charpy Impact Properties of HAZ Material 23 Donald C. Cook Unit 2, Capsule Y VIII Charpy Impact Properties of Weld Metal 23 Donald C. Cook Unit 2, Capsule Y IX Effect of Irradiation on Capsule Y Surveillance 28 Materials, Donald C. Cook Unit No. 2 l X Tensile Properties of Surveillance Materials 29 Donald C. Cook Unit No. 2 XI Projected Values of RT f r Donald C. Cook 34 NDT Unit No. 2 1 XII Reactor Vessel Surveillance Capsule Removal 37 Schedule, Donald C. Cook Unit No. 2 1 I 8 v

I. SUMARY OF RESULTS AND CONCLUSIONS The analysis of the second material surveillance capsule removed from the Donald C. Cook Unit No. 2 reactor pressure vessel led to the following cenclusions: (1) Based on a calculated neutron spectral distribution, Capsule Y 18 received a fast fluence of 6.7 x 10 neutrons /cm2 (E > 1 MeV) at its radial c:nter line. (2) The surveillance specimens of the core beltline materials ex-parienced shifts in RT f 60*F to 100*F as a result of exposure up to NDT the 1982 refuelling outage. (3) The core beltline plate materials exhibited the largest shifts in RT Since the intemediate shell plate material has the highest int-NDT. tial (unirradiated) RTNDT, it will control the heatuo and cooldown limita-tions throughout the design lifetime of the pressure vessel. 18 2 (4) The estimated maximum neutron fluence of 1.88 x 10 neutrons /cm (E > 1 MeV) received by the vessel wall accrued in 3.24 effective full power years (EFPY). The projected maximum neutron fluence after 32 EFPY is 19 1.6 x 10 neutrons /cm2 (E > 1 MeV). This estimate is based on the assump-tion that the low leakage core configuration installed after Core Cycle 2 will be used throughout the remainder of the design life of the vessel. (5) Based on the analyses of Capsules T and Y, the projected values of RT f r the Donald C. Cook Unit 2 vessel core beltline region, at the NDT l 1/4T and 3/4T positions after 12 EFPY of operation, are 145'F and 105'F, re-i l spectively. These values were used as the bases for computing revised heat- [ up and cooldown limit curves for up to 12 EFPY of operation. l l

(6) Based on the analyses of Capsules T and Y, the values of RTNDT for the Donald C. Cook Unit 2 vessel core beltline region, at the 1/4T End 3/4T postions after 32 EFPY of operation, are projected to be 175'F and 135'F, respectively. These values were used as the bases for computing heatup and cooldown limit curves for up to 32 EFPY of operation. (7) The Donald C. Cook Unit No. 2 vessel plates, weld metal, and HAZ material located in the core beltline region are projected to retain suffi-cient toughness to meet the current requirements of 10CFR50 Appendix G throughout the design life of the unit. I 2

II. BACKGROUND The allowable loadings on nuclear pressure vessels are determined by applying the rules in Appendix G, " Fracture Toughness Requirements," of 10CFR50 [1]. In the case of pressure-retaining components made of ferritic materials, the allowable loadings depend on the reference stress intensity factor (KIR) curve indexed to the reference nil ductility temperature (RTNDT) presented in Appendix G, " Protection Against Non-Ductile Failure," of Section III of the ASME Code [2]. Further, the materials in the beltline region of the reactor vessel must be monitored for radiation-induced changes in RT Per the requirements of Appendix H, " Reactor Vessel Material Sur-NDT vst11ance Program Requirements," of 10CFR50. The RT is defined in paragraph NB-2331 of Section III' of the ASME NDT Code as the highest of the following temperatures: (1) Drop-weight Nil Ductility Temperature (DW-NDT) per ASTM E 208 [3]; (2) 60 deg F below the 50 ft-lb Charpy V-notch (C ) y temperature; (3) 60 deg F below the 35 mil C temperature. y The RT must be established for all materials, including weld metal and NDT h:at-affected zone (HAZ) material as well as base plates and forgings, which comprise the reactor coolant pressure boundary. It is well established that ferritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 10 neutrons per cm2 (E > 1 MeV) [4]. 17 Also, it has been established that tramp elements, particularly copper and 3

phosphorus, affect the radiation embrittlement response of ferritic mate-rials [5-7]. The relationship between increase in RT and copper content NDT is defined in Regulatory Guide 1.99. Estimates of shifts in RT in this NDT report are based on the current Revision 1 of Regulatory Guide 1.99 [83. In general, the only ferritic pressure boundary materials in a nuclear plant which are expected to receive a fluence sufficient to affect RT are NDT those materials which are located in the core beltline region of the reactor pressure vessel. Therefore, material surveillance programs include speci-mens machined from the plate or forging material and weldments which are located in the core beltline region of high neutron flux density. ASTM E 185 [9] describes the recommended practice for monitoring and evaluating the radiation-induced changes occurring in the mechanical properties of pressure vassel beltline materials. Westinghouse has provided such a surveillance program for the Donald C. Cook Unit No. 2 nuclear power plant. The encapsulated C specimens are y located on the 0.D. surface of the thermal shield where the fast neutron flux density is about three times that at the adjacent vessel wall surface. Therefore, the increases (shifts) in transition temperatures of the mate-rials in the pressure vessel are generally less than the corresponding shifts observed in the surveillance specimens. However, because of azimuth-al variations in neutron flux density, capsule fluences may lead or lag the maximum vessel fluence in a corresponding exposure period. The capsules also contain several dosimeter materials for experimentally determining the average neutron flux density at each capsule location during the exposure c i period. 1 i 4

The Donald C. Cook Unit No. 2 material surveillance capsules also in-clude tensile specimens as recommended by ASTM E 185. At the present time, irradiated tensile properties are used only to indicate that the materials tested continue to meet the requirements of the appropriate material spect-fication. In addition, the material surveillance capsules contain wedge opening loading (WOL) fracture mechanics specimens. Current technology limitations result in the testing of these specimens at temperatures well below the minimum service temperature in order to obtain valid fracture mechanics data per ASTM E 399 [10], " Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials." Currently, the NRC sug-gests storing these specimens until an acceptable testing procedure has been defined for determing the J fracture toughness [113. Ic This report describes the results obtained from testing the contents of Capsule Y. These data and those obtained previously from Capsule T are analyzed to estimate the radiation-ind'uced changes in the mechanical properties of the pressure vessel at the time of the refuelling outage as well as predicting the changes expected to occur at selected times in the future operation of the Donald C. Cook Unit No. 2 power plant. 5

III. DESCRIPTION OF MATERIAL SUP,VEILLANCE PROGRAM The Donald C. Cook Unit No. 2 material surveillance program is des-cribed in detail in WCAP 8512 [123, dated November 1975. Eight materials surveillance capsules were placed in the reactor vessel between the thermal shield and the vessel wall prior to startup, see Figure 1. The vertical center of each capsule is opposite the vertical center of the core. The neutron flux density at each 40' capsule location exceeds three times the maximum flux density on the vessel I.D. [13]. However, the peak vessel ex-posure rate has been significantly reduced since the introduction of a low leakage core configuration in Core Cycle 3. The capsules each contain Charpy V-notch, tensile, and WOL specimens machined from the SA533 Gr B, C1 1 plate, weld metal, and heat-affected zone (HAZ) materials located at the core beltline. The chemistries and heat treatments of the vessel surveillance materials are summarized in Table I. All test specimens were machined from the test materials at the quarter-thickness (1/4 T) location after performing a simulated postweld stress-relieving treatment. Weld and HAZ specimens were machined from a stress-relieved weldment which joined sections of the intermediate and lower shell plates. HAZ specimens were obtained from the plate C5521-2 side of the celdment. The longitudinal base metal C specimens were oriented with their y long axis parallel to the primary rolling direction and with Y-notches per-pendicular to the major plate surfaces. The transverse base metal C speci-y mens were oriented with their long axis perpendicular to the primary rolling direction and with V-notches perpendicular to the major plate surfaces. Tensile specimens were machined with the longitudinal axis perpendicular to 6

X (220') 270' / / l W (164*) Y (320') / Z (356') 180' / + / o' / / 5 (4*) V (176') l j - T (40') 90' U (140') Reactor Vessel Thermal Shield Core Barrel FIGURE 1. ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE PRESSURE VESSEL 7

TABLE I DONALD C. COOK UNIT NO. 2 REACTOR VESSEL SURVEILLANCE MATERIALS [123. Heat Treatment History Shell Plate Material: Heated to 1700 F for 4-1/2 hours, water quenched. Heated to 1600 F for 5 hours, water quenched. Tempered at 1250 F for 4-1/2 hours, air cooled. Stress relieved at 1150 F for 51-1/2 hours, furnace cooled. Weldment: Stress relieved at 1140 F for 9 hours, furnace cooled. Chemical Composition (Percent) Material C Mn P S St Ni Mo Cu Cr Plate C-5521-2(a) 0.21 1.29 0.013 0.015 0.16 0.58 0.50 0.14 Plate C-5521-2(b) 0.22 1.28 0.017 0.014 0.27 0.58 0.55 0.11 0.072 Weld Metal (b) 0.11 1.33 0.022 0.012 0.44 0.97 0.54 0.055 0.068 Weld Metal (c) 0.08 1.42 0.019 0.016 0.36 0.96 0.05 0.07 (a) Lukens Steel analysis. (b) Westinghouse analysis. (c) Chicago Bridge and Iron analysis. 8

the plate primary rolling direction. The WOL specimens were machined with the simulated crack parallel to the primary rolling direction and perpendi-cular to the major plate surfaces. All mechanical test specimens, see Figure 2, were taken at least one plate thickness from the quenched edges of the plate material. Capsule Y contained 44 Charpy V-notched specimens (8 longitudinal and 12 transverse from the plate material, plus 12 each from weld metal and HAZ material); 4 tensile specimens (2 plate and 2 weld metal); and 4 weld metal WOL specimens. The specimen numbering system and location within Capsule Y is shown in Figure 3. Capsule Y also was reported to contain the following dosimeters for d:termining the neutron flux density: Target Element Form Quantity Iron Bare wire 5 Copper Bare wire 3 Nickel Bare wire 3 Cobalt (in aluminum) Bare wire 2 Cobalt (in aluminum) Cd shielded wire 2 Uranium-238 Cd shielded oxide 1 Neptunium-237 Cd shielded oxide 1 Two eutectic alloy thermal monitors had been inserted in holes in the steel spacers in Capsule Y. One (located at the bottom) was 2.5% Ag and 97.5% Pb with a melting point of 579'F. The other (located at the top of the capsule) was 1.75% Ag, 0.75% Sn, and 97.5% Pb having a melting point of 5 90* F. i 9 a

W [' i Is,s .f gi \\'" l t 'i' I W ! '33 111-I v2s naos (c) Charpy v-notch impcet specimen I y,5 w i.e DQ Mh> D e- . i* f 1

f. 2

'89" L250 l L480 4 250 m e i g g in- .3 d (b) Tensile specimen t45 t 30 g\\ uzo teos '3C Y l ,,ai.4n , i t not**7 4 I 2 tg pm t i..us L.il i i Tl~l et QQ lso l .00i s.499l (c) Wedge opening loading specimen FIGURE 2. VESSEL MATERIAL SURVEILLANCE SPECIMENS 10

h Top Tensile .. i 4 m it WCL 1 2 10 WCL 29 WOL MW MW Tensile 13 14 T'T y h Charpy s..a $pecimen Cooe: MT - Plate C5521-2 (Transverse) W'T ML - Plate C5521-2 (Longitudinal) [ ["' Charpy W = Weld Metal MM - Weld Heat Affected Zone ^'* 216 Dos tmeter kk' Charpy 9 9 ..a M"W Charpy w= Charpy w.w 5 % Charpy . Z..a W'E M E Charpy W L w'% y y Charpy d ..u.. W"E-.f Charpy W'T 'Z. 3 Charpy E.b w-w Bottos y y Charpy 5.,1L FIGURE 3. ARRANGEMENT OF SPECIMENS IN CAPSULE Y 11

1 l l IV. TESTING OF SPECIMENS FROM CAPSULE Y The capsule shipment, capsule opening, specimen testing, and reporting of results were carried out in accordance with the Project Plan for Donald C. Cook Unit No. 2 Reactor Vessel Irradiation Surveillance Program. The SwRI Nuclear Projects Operating Procedures called out in this plan include: (1) XIII-MS-104-1, " Shipment of Westinghouse PWR Vessel Mate-rial Surveillance Capsule Using SwRI Cask and Equipment" (2) XI-MS-101-1, " Determination of Specific Activity and Analysis of Radiation Detector Specimens" (3) XI-MS-103-1, " Conducting Tension Tests on Metallic Specimens" (4) XI-MS-104-1, "Charpy Impact Tests on Metallic Specimens" (5) XIII-MS-103-1, " Opening Radiation Surveillance Cap-sules and Handling and Storing Specimens" Copies of the above documents are on file at SwRI. A. Shipment, Opening, and Inspection of Capsule Southwest Research Institute utilized Procedure XIII-MS-104-1 for the chipment of Capsule Y to the SwRI laboratories. SwRI personnel severed the capsule from its extension tube, sectioned the extension tube into several lengths, supervised the loading of the capsule and extension tube materials into the shipping cask, and transported the cask to San Antonio, Texas. The capsule was opened and the contents identified and stored in ac-cordance with Procedure XIII-MS-103-1. The long seam welds were milled off using a Bridgeport vertical milling machine. Before milling the long seam weld beads, transverse saw cuts were made to remove the capsule ends. After the long seam welds had been milled off, the top half of the capsule shell was removed. The specimens and spacer blocks were carefully removed and placed in indexed receptacles identifying each capsule location. After the disassembly had been completed, each specimen was carefully checked to 12

insure agreement with the identification and location as listed in WCAP 1 8512 [123. No discrepancies were found. The thermal monitors and neutron dosimeter wires were removed from the holes in the spacers. The thermal monitors, contained in quartz vials, were cxamined. No evidence of melting was observed, thus indicating that the I maximum temperature during exposure of Capsule Y did not exceed 579'F. All neutron dosimeters except for the Od-covered Co-Al were in the positions called out in WCAP 8512 and were correctly accounted for. Although the Co (Cd) dosimeters were apparently not included as a part of the capsule, there was no impact on the capsule fluence determination. B. Neutron Dosimetry The dosimeter wires were weighed on a Mettler microbalance, and the fission monitors were weighed on a Mettler digital balance after these mate-1 I rials had been doencapsulated. The gamma activities of the dosimeters were determined in accordance with Proc'edure XI-MS-101-1 using an IT-5400 multi-channel analyzer and a Ge(Lf) coaxial detector system. The calibration of l the equipment was accomplished with Mn, Co, and Cs radioactivity standards obtained trom the U.S. Department of Commerce National Bureau of Standards. All activities were corrected to the time-of-removal (TOR) at reactor shutdown. 4 Since a low leakage core configuration was placed in service during I core 3, the last cycle prior to the removal of Capsule Y, the flux density calculation must consider the specific spectrum-averaged reaction cross sec-tions (0,) as well as the average fraction of full power in each operating period (fP,), as follows: A/No

m=n I f5,P (I-g g)(g-At) m m

m=1 1 t 13

activity at TOR, dps; where: A = 4 energy-dependent neutron flux density, = n/cm2-sec; 5, spectrum-averaged activation cross section, = cm2; N, number of target atoms per mg; = decay constant for the activation product, day-1; A = t, decay time after operating period m, days; = T, operating days; = low leakage core peak flux factor; f = P, average fraction of full power during operating = period. The total neutron fluence is then equal to the product of the average neu- .. tron flux density during each fuel cycle and the corresponding reactor oper-Eting time at full power. The values of t,, T,, f and P,up to the 1982 refueling shutdown for Donald C. Cook Unit No. 2 are presented in Table II. The calculation of the neutronic factors is described below. As a part of the analysis of Capsule T [143, Southwest Research Insti-tute performed a two-dimensional discrete ordinates transport calculation uith the DOT 3.5 code, a 22-group neutron cross section library (CASK), a Py cxpansion of the scattering matrix, and an S rder of angular quadrature. 8 A one-eighth segment of a plane through the vertical axis was used to model the core, core barrel, thermal shield, surveillance capsules and holders, pressure vessel, and three water regions using R-9 coordinates. Two similar calculations, one for the original and the other for the low leakage core configuration, were made using a more recent version of the transport code (DOT-IV) and a neutron cross section library (47-group BUGLE-80) based on a newer version of the Evaluated Nuclear Data File ENDF/B-IV [153. 14

1 l TABLE II SUNNARY OF REACTOR OPERATIONS DONALD C. COOK UNIT NO. 2 4 Orerating Power Fraction of Decay Time Operating Dates Days Shutdown Generation Full Power After Period P:riod start stop (T ) Days (MWD ) (P,) (t.) t 1 03/22/78 04/19/78 33 23,% 9 0.2142 1678 04/20/78 05/02/78 14 2 05/03/78 05/19/78 17 27,264 0.4729 1647 05/20/78 05/31/78 12 3 06/01/78 06/03/78 3 5,510 0.5416 1632 .06/04/78 06/04/78 1 4 06/05/78 07/21/78 47 103,423 0.6489 1584 07/22/78 08/04/78 14 5 08/05/78 08/21/78 17 51,223 0.8888 1553 08/22/78 08/22/78 1 6 08/23/78 08/28/78 6 12,543 0.6167 1546 08/29/78 09/01/78 4 7 09/02/78 11/09/78 69 206,496 0.8826 1473 11/10/78 11/23/78 14 8 11/24/78 04/07/79 135 423,785 0.9257 1324 04/08/79 04/08/79 1 9 04/09/79 05/19/79 41 132,362 0.9520 1282 05/20/79 07/02/79,) 44 g 10 07/03/79 10/19/79 109 356,918 0.9656 1129 10/20/79 01/06/80 89 11 01/17/80 06/27/80 163 508.036 0.9191 877 06/28/80 07/12/80 15 12 07/13/ 80 10/18/80 98-321,325 0.9669 764 10/19/80 12/06/80 49 13 12/07/80 12/08/80 2 114 0.0168 7 13 12/09/80 12/09/80 1 14 12/10/80 12/14/80 5 9,668 0.5702 707 12/15/ 80 12/15/80 1 IDI 15 12/16/ 80 03/14/81 89 295,217 0.9782 617 03/15/81 05/17/81 64 16 05/18/81 07/03/81 47 124,407 0.7806 506 07/04/81 07/10/ 81 7 17 07/11/81 10/02/81 84 271,047 0.9516 415 10/03/81 10/23/81 21 18 10/24/81 10/25/81 2 1,147 0.1691 392 10/26/81 10/29/81 4 19 10/30/81 03/11/82 133 433,484 0.9612 255 03/12/82 03/29/82 18 20 03/30/82 08/01/82 125 405,618 0.9569 112 08/02/82 08/17/82 16 21 08/18/82 09/30/82 44 135,420 0.9076 52 10/01/82 10/01/82 1 22 10/02/82 11/21/82 51 164,672 0.9522 0 C' 4,013,648 (a) End of Core Cycle 1 (b) End of Core Cycle 2 (c) Equals 3.24 EFPY 8 3391 MW t 15

The resulting axial, radial, and azimuthal dependence of the fast neu-tron (E > 1.0 MeV) flux density and energy spectrum within the reactor ves-sal and surveillance capsules were used to calculate the spectrum-averaged cross sections for the threshold and the fission detectors as well as the lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsule. The pertinent factors obtained from these transport calculations are summarized in Table III. Although the peak ves-sal fluence rate for Cycle 3 is reduced approximately 20% from Cycles 1 and 2 by the change in core configuration for Cycle 3, the 40' capsule lead fac-tors are the same for both core configurations because of the proximity of the capsules to the peak flux azimuthal position (45*). Also, the reaction cross sections resulting from the use of BUGLE-80 are generally smaller than those generated using CASK and, as a result, the calculated capsule fluence rates (flux densities) are larger than previously reported [14]. However, since the vessel exposure rate is essentially the same as that calculated previously, the capsule lead factors for both core configurations are pro-portionately increased. In Capsule Y, the weld metal and the heat affected zone Charpy speci-mens were located in the specimen layer nearest to the vessel wall and the longitudinal and transverse shell plate Charpy specimens were located in the specimen layer nearest to the core. Since there is a radial dependence of the fast neutron flux in the vessel, the neutron expcsure received by the wold metal and heat affected zone Charpy specimens is expected to be lower than that received by the longitudinal and transverse shell plate Charpy specimens. The dosimetry program is capable of providing information on the radial dependence of the fast flux since the copper and nickel threshold d tectors were located on the radial centerlines of the Charpy specimen 16

TABLE III RESULTS OF DISCRETE ORDINATES Sn TRANSPORT ANALYSIS "^' hi:Sns "' A. Calculated Reaction Cr'oss Sections for Analysis of Fast Neutron Monitors (E > 1.0 Mov) 5(barns) Reaction Core Cycles 1 and 2 Core Cycle 3 54Fe(n p)54Mn .0694 .0697 58Ni(n,p)58Co .0933 .0938 63Cu(n, )60Co .000741 .000750 3 8 (n,f) .358 .359 U 237Np(n,f) 2.60 2.61 B. Calculated Capsule Lead Factors Position (" Location within Capsule Lead Factor I 211.7 cm Center of core-side Charpy layer 3.74 211.9 cm Center of capsule 3.57 212.2 cm Center of two specimen layers 3.33 212.7 cm Center of vessel-side Charpy layer 2.96 C. Low Leakage Core Peak Flux Factor Cycle 3 neutron flux density, E > 1.0 MeV = 0.806 Cycles 1 and 2 neutron flux density, E > 1.0 MeV (a) Distance from center of core (b) Capsule neutron flux density, E > 1.0 MeV Maximum neutron flux density at vessel I.D., E > 1.0 MeV 17

_ - - - _ - ~ l layers nearest to and farthest from the core, respectively, the iron thresh-old detectors were located at the radial position corresponding to the in-terface between two Charpy specimen layers, and the fission monitors were I at the radial centerline of the capsule. The activities of dosimeters and resulting flux values obtained from Capsule Y are presented in Table IV. A summary of the capsule Y and vessel I.D. fluxes calculated for full-power operation during Core Cycle 3 is as follows: 1 Dosimeter Measured Capsule Flux Lead Peak Vessel Flux at I.D. Type cm-2.sec-1, E > 1 MeV Factor cm-2.sec-1, E > 1 MeV 10 10 Copper 5.27 x 10 3.74 1.41 x 10 10 10 Fission 6.03 x 10 3.57 1.69 x 10 10 Iron 5.22 x 10 3.33 1.57 x 10 10 10 Nickel 5.03 x 10 2.96 1.70 x 10 The discrepancies in the peak vessel flux values determined from the j several dosimeter materials are attributed primarily to the uncertainties in i the calculated spectra and in the reaction cross sections. Other neutronic factors contributing to the estimated i 16.5% uncertainty (10) in a calcu-lated. flux value are the determination of disintegration rates and the cal-l culation of reaction rates (A /N )* SAT O Averaging the results obtained from the Capsule Y neutron dosimeters, the peak neutron flux incident of the I.D. surface of the pressure vessel 10 -2.sec-1, E > 1 MeV. during Core Cycle 3 is calculated to be 1.59 x 10 cm

If a fission-spectrum energy distribution is assumed at the capsule lo-cation, the cross section for the 54 e(n,p)S4 n reaction (E > 1.0 MeV)

F M tould be 98.26 mb [4], and the resulting value for fast flux at the cap-sule location during Core Cycle 3 would be 3.7 x 1010 cm-2.sec-1 This value is reported for reference only and has not been used in the analysis of results. 18 I

TABLE IV SUW4ARY OF NEUTRON DOSIMETRY RESULTS DONALD C. COOK UNIT NO. 2, CAPSULE Y Caesule Flux (em-2.see-l. E > 1 MeV(b) cosimeter Oosimeter Activation ATOR Position (a) Identification Reaction (dos /me) Cveles 1 & 2 Cycle 3 1 10 10 63Cu(n,a)60Co 9.36 x 10 6.46 x 10 5.20 x 10 211.7 cm Cu (Top Middle) 1 10 10 Cu (Middle) 9.59 x 10 6.62 x 10 5.33 x 10 1 10 10 Cu (Botten Middle) 9.46 x 10 6.53 x 10 5.26 x 10 10 5.27 x 1010 Average = 6.57 x 10 2 10 10 38 (n,f)U7C4 3.00 x 10 8.40 x 10 6.77 x 10 U 211.9 cm U-238 (Middle) U7 10 10 f Np-237 (Middle) Np(ne f) Cs 1.78 x 10 6.57 x 10 5.30 x 10 Average = 7.49 x 1010 6.03 x 1010 3 10 10 54Fe(n p)S4Mn 1.76 x 10 6.31 x 10 5.09 x 10 212.2 cm Fe (Top) 3 10 10 Fe (Top Middle) 1.80 x 10 6.44 x 10 5.19 x 10 3 10 10 Fe (Middle) 1.84 x 10 6.57 x 10 5.30 x 10 3 10 10 Fe (Botton Middle) 1.83 x 10 6.56 x 10 5.28 x 10 3 10 10 Fe (Botton) 1.81 x 10 f.49 x 10 5.23 x 10 Average = 6.18 x 1010 5.22 x 1010 4 10 10 58N1(nep)S8Co 2.87 x 10 6.18 x 10 4.98 x 10 212.7 cm Ni (Top Middle) 4 10 10 Ni (Middle) 2.88 x 10 6.22 x 10 5.01 x 10 4 10 10 Ni (Bottom Middle) 2.92 x 10 6.31 x 10 5.08 x 10 Average = 6.24 x 1010 5.03 x 1010 Thermal Flux 59Co(ney)60Co 1.02 x 10 (c) (d) 212.7 cm Co (Top) 7 Co (Bottom) 1.02 x 10 (d) (d) Co(Cd) (Top) (c) Co(Cd) (Bottom) Y (c) t (a) Distance from center of core. l (b) C21culated flux values subject to 116.5% uncertainty (le). l (c) Cd-covered dosimeters not in capsule. l (d) Not applicable fcr detenntning fast flux. 19

Dividing this result by the Low Leakage Core peak flux factor of 0.806 (see Table III), the peak neutron flux incident on the I.D. surface of the pres-10 -2.sec~1, sure vessel prior to Core Cycle 3 is calculated to be 1.98 x 10 cm 4 E > 1 MeV. The latter is within 15 of the value determined from the anal-10 ysis of Capsule T [14] and within 7% of the revised value of 2.12 x 10 ob-tained when the Table III constants are applied to the Capsule T dosimeter activities. The calculated full power neutron flux for the weld metal and i haat affected zone Charpy specimen layer is given by: 10 10 Core Cycles 1 and 2: 1.98 x 10 x 2.96 5.86 x 10 = 10 10 Core Cycle 3: 1.60 x 10 x 2.% 4.74 x 10 = i Similarly, the calculated full power neutron flux for the longitudinal i j plate, HAZ material and reference material Charpy specimens, the tensile specimens, and the WOL specimens are given by: l 10 10 1.98 x 10 x 3.74 = 7.40 x 10 (Long. Plate and HAZ C, Cycles 1 and 2) y 1.60 x 10 x 3.74 = 5.98 x 1010 (Long. Plate and HAZ C, Cycle 3) 10 y 1.98 x 10 x 3.33 = 6.59 x 1010 (Tensile Specimens, Cycles 1 and 2) 10 1.60 x 10 x 3.33 = 5.33 x 1010 (Tensile Specimens, Cycle 3) 10 j 1.98 x 10 x 3.57 = 7.07 x 1010 (WOL Specimens, Cycles 1 and 2) 10 10 1.60 x 10 x 3.57 = 5.71 x 1010 (WOL Specimens, Cycle 3) I Since Donald C. Cook Unit No. 2 operated for 730.70 effective full power days (EFPD) during Core Cycles 1 and 2, and 452.9 EFPD during Core l l Cycle 3, the calculated fluences for Capsule Y and the vessel up to the l 1982 refuelling outage are as follows: r 18 2 o Weld Metal and HAZ C Specimens - 5.55 x 10 n/cm y 18 2 o Long. and Trans. Plata C Specimens - 7.01 x 10 n/cm y 18 2 l o Tensile Specimens - 6.25 x 10 n/cm i l 20 i

l o WOL Specimens - 6.70 x 1018n/cm2 18 2 o Pressure Vessel ID Surface - 1.88 x 10 n/cm A C. Mechanical Property Tests The irradiated Charpy V-notch specimens were tested on a calibrated

SATEC Model SI-1K 240 ft-1b,16 ft/sec impact machine in accordance with Procedure XI-MS-104-1. The test temperatures, selected to develop the ductile-brittle transition and upper shelf regions, were obtained using a liquid conditioning both monitored with a Fluke Model 2168A digital thermo-meter. The Charpy V-notch impact data obtained by SwRI on the specimens centained in Capsule Y are presented in Tables V through VIII. The shifts in the Charpy V-notch transition temperatures determined for the vessel plate, the weld metal and the HAZ materials are shown in Figures 4 through 7.

The Capsule T results are included for comparison. A summary of the shifts.in RTNDT determined at the 30 ft-lb level as specified in Appendix G to 10 CFR 50 [1], and the reduction in C upper y shelf energies for each material, is presented in Table IX. Tensile tests were carried out in accordance with Procedure XI-MS-i 103-1 using a 22-kip capacity MTS Model 810 Material Test System equipped with an Instron Catalogue No. G-51-13A 2-in. strain gage extensometer and Hewlett Packard Model 7004B X-Y autographic recording equipment. Tensile tests on the plate material and the weld metal were run at 250*F and 550*F at a strain rate of 0.005 in/in/ min. through the 0.2% offset yield strength i using servocontrol and ramp generator. The results, along with tensile data rsported by Westinghouse on the unirradiated materials [123, are presented l in Table X. The load-strain records are included in Appendix A. l l w Inspected and calibrated using specimens and procedures obtained from the l Army Materials and Mechanics Research Center. l l 21 l

l l. TABLE V CHARPY IMPACT PROPERTIES OF LONGITUDINAL PLATE DONALD C. COOK UNIT NO. 2 CAPSULE Y Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance .i No. (*F) (ft-lb) (mils) (5 shear) ML-52 75 20.5 19 5 ML-55 100 24.0 25 15 ML-54 120 38.0 31 25 ML-51 160 69.0 60 60 ML-53 180 62.5 58 80 ML-50 210 102.5 77 100 ML-49 250 101.0 79 100 ML-56 300 106.5 83 100 TABLE VI CHARPY IMPACT PROPERTIES OF TRANSVERSE PLATE DONALD C. COOK UNIT NO. 2 CAPSULE Y Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance No. (*F) (ft-lb) (mils) (5 shear) MT-79 40 8.0 8 5 MT-83 60 10.5 9 2 MT-76 75 21.0 21 5 MT-84 100 23.0 21 5 i MT-78 120 21.5 24 10 MT-81 145 33.5 32 20 l MT-82 145 34.0 31 25 MT-75 160 46.5 42 25 MT-77 180 51.0 46 30 MT-74 210 63.5 56 100 MT-73 250 68.5 66 100 MT-80 300 74.5 66 100 22

d TABLE VII CHARPY IMPACT PROPERTIES OF HAZ MATERIAL DONALD C. COOK UNIT 2 CAPSULE Y Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance No. (*F) (ft-lb) (mils) (% shear) m-81 -25 7.5 8 5 MH-83 -10 24.0 21 15 MH-82 0 32.0 26 15 MH-77 20 28.5 27 10 MH-84 20 44.5 38 20 MH-78 40 55.5 46 30 MH-79 40 44.0 33 25 MH-76 75 90.0 67 100 MH-80 75 79.0 57 90 MH-75 160 65.5 65 100 MH-74 210 92.5 81 100 MH-73 250 114.5 90 100 TABLE VIII CHARPY IMPACT PROPERTIES OF WELD METAL DONALD C. COOK UNIT 2 CAPSULE Y Fracture Lateral Fracture Specimen Temperature Energy Expansion Appearance No. (*F) (ft-lb) (mils) (5 shear) W-83 -25 15.5 15 10 MW-84 0 15.5 17 10 MW-79 20 24.0 23 15 MW-80 40 31.5 30 20 MW-74 75 50.0 44 60 MW-82 75 27.0 25 25 l MW-76 120 41.5 39 25 MW-81 145 44.0 42 20 MW-73 160 74.5 70 100 MW-75 180 71.0 68 100 MW-78 210 63.0 61 100 MW-77 250 63.5 61 100 1 23

150 ~ Code: "~~- ' ~ ~~~ ~~~~~ Unirradiated Capsule T 120 - -- *-- Capsule Y E ge:;o'~~*~

h :.5 f;

l l j ~ ~ ~ ~- ' ' '. E0 = b ~ l--er,W.^ ? E ha995f 1 i t_ 3 -_,'b._.5 -l-j w

3

,-o, yh.00 F-- _. _.b _/-.c -1C0 0 1C0 200 300 400 533 100 .4.-m s 75 ^ f ~ *,., - E ~ZZ_T '~ j--/"

== / ,' c f 8 l j /-_ 3:5 f. 555 ~ c~ l' l-- a nl 1 ,f ^ &~~. BOF-o ~ 1 ~~, - - 3 l .-/ j, </ v _. _ _ _ ~. , _ _ ~. _. _, _. - _., _ 1C0 0 100 200 300 403 5:0 Test Temperature, 'F FIGURE 4. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 VESSEL SHELL PLATE C5521-2 (LONGITUDINAL ORIENTATION) 24

l00 ' toee: _T_-- Unirradiated __ - Caesale T 75 2 -- e- -Capsule Y _e2 g -g... w.- 3 ,r, ' _' ~ T r : =. _:~ ~.~~ i j .-. :== .1 105F t 50 = g ,r ,/ '- rL 9 5 __f a . = = -- ?c i y .1 1. -y i % )oor e A 'W ~ 25 s i ^e i .a / /-,--. ~ / 6E .5: ? s -100 0 100 200 300 400 5 ",0 100 ..ZZZZ 75 _ c'p~E^- e =~~ ~~~ ~ T :~-~:. ~-~ ,e - E, _-...a-g y ~, c c g. ?. "U 1 I x i .e w 1 A~- ~7'S F . TJ 12 r 3 .1 C ~~~~ ~ '

^ ~. ^

/ g 2 a, ~, o.:_ ' l ' - - ~ ~ -. /_- ~~~ .=_-_.-._.: @.. '-- -- i 0 -1:3 0 100 200 300 400 E00 Test Temperature. *F FIGURE 5. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 VESSEL SHELL PLATE C5521-2 (TRANSVERSE ORIENTATION) 25

1(0 ~~~-~ -~ . Csde- . :Z ~ ~ Untrradiated Capsule T ~~ 120 -- e-- Capsule Y e .c ,~ q /., .a 3.,

_' c.

. e, -.,c E' . -./ s EC t'

.~..

5 ~ ~ ~~

...'_."'~~

~ t -_. ' '...l e' /.~p4::;70i e 3 y_ 4C /_. j.*h10F

:.~._

=s. /,t_. ...,_, c e 103 0 100 200 300 400 500 _._g... _. _. 100 -_- w.....-.. pe..- ~ W ---- IE .e Ac. -- T ** --- ~ fy'*- - E g f_,.s. -f e- ,c r-4. t ^ c. e. = w -1 _ -e n m ,_ 65F ,t

2. an._-

a p ~ '. ~:.

. ~ _^. C

~ - _..:..C :. ~ ~, ~.. ~~ ^' /_~ ' ..'_ C. e-G ' , ~~-- ~ - ~ ' 1 300 4'3 500 100 0 100 200 Test Temperature 'F F'GURE 6. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 REACTOR VESSEL HEAT-AFFECTED ZONE MATERIAL 26

100 ~ Unirradiated Capsule T ~~~ n - --. -- capsul e Y

_ p -

__, s - - _ = 7 ,7,:-.-.._ e g _y = =. = - 50

)_ h 45F

_~ ~ [] w ./ y yg 5 p;'_*gd,_* y- /p / _r y. i i e - -- - - ( -100 0 100 200 300 400 E00 100 -. _.. ~. - ~_-~ _ _ _ Z- _. '.T Z s _ C '.: ~ ? M =z ,c.. s.-f: 7 ~~ -{o:(j f -~:~.: t d--~;-30F 3 p = - -. - -. jp. -.. J -100 0 100 200 300 400 500 Test Temeerature. *F FIGURE 7. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 REACTOR VESSEL WELD MATERIAL 27

TABLE IX EFFECT OF IRRADIATION ON CAPSULE Y SURVEILLANCE MATERIALS DONALD C. COOK UNIT NO. 2 Weld HAZ Trans. Plate Long. Plate Criterion (l) Metal (2) Material (2) C5521-2(3) C5521-2(3,5) Transition Temperature Shift 8 50 ft-lb 45*F 70*F 105'F 100*F 8 30,ft-lb 50*F 70*F 100*F 90*F 8 35 mil 30*F 65'F 75"F 95'F I' RT 50*F 70*F 100*F 90*F NDT C Upper Shelf Drop 70 ft-lb 31 ft-lb 18 ft-lb 24 ft-lb v (9%) (25%) (20%) (19%) (1) Refer to Figures 4-7. (2) Fluence = 5.55 x 1018 n/cm2, E > 1 MeV. (3) Fluence = 7.01 x 1018 n/cm2, E > 1 MeV. (4) Transition temperature shift at 30 ft-lb (46 ft-lb for longitudinal plate). (5) Transition temperatures at 77 ft-lb, 46 ft-1.b, and 54 mils [173.

TABLE X TENSILE PROPERTIES OF SURVEILLANCE MATERIALS I DONALD C. COOK UNIT NO. 2 l Fracture Fracture Uniform Total ~ l Test Spec. Temp. 0.2% YS UTS Load Stress Elongation Elongation R.A. l Condition Material No. (*F) (ksi) (ksi) (1b) (ksi) (5) (5) (%) Capsule Y(a) Plate C5521-2 MT-13 210 72.0 93.2 3447 150.1 15.2 19.3 53.1 (Transverse) MT-14 550 68.3 90.3 3367 146.6 13.4 18.6 53.1 Weld Metal W-13 210 75.8 93.0 3097 182.5 13.2 21.2 65.4 i W-14 550 73.6 91.8 3197 164.8 12.2 19.5 60.4 Room 67.4 87.3 3200 161.2 13.4 23.4 59.6 (b) Plate C5521-2 Room 65.4 85.9 2950 156.4 15.0 27.1 61.7 (Transverse) 300 58.8 78.6 2650 146.1 13.0 22.6 63.1 m 300 60.5 79.5 2675 157.6 10.6 19.8 65.4 550 57.5 83.0 3225 142.1 11.5 19.0 53.8 553 58.9 83.1 3150 145.6 12.7 20.5 56.0 Room 75.7 93.2 2850 173.4 13.9 25.7 66.8 Weld Metal Room 76.9 91.3 2950 17 8.8 12.2 22.6 66.6 300 70.7 88.0 2900 171.0 10.7 20.7 66.0 300 71.0 85.3 2875 179.0 10.3 21.2 67.5 550 70.0 87.2 3160 157.2 10.1 19.2 59.6 550 68.2 87.8 3050 166.0 9.3 20.2 62.8 18 2 (a) Fluence = 6.25 x 10 m/cm, E > 1 MeV. (b) Unirradiated [12].

Testing of the WOL specimens was deferred at the request of Indiana & Michigan Electric Company. The specimens are in storage at the SwRI radi-atton laboratory.. l I I i l i 30

i V. ANALYSIS OF RESULTS The analysis of data obtained from surveillance program specimens has the following goals: (1) Estimate the period of time over which the properties of the l vzssel beltline materials will meet the fracture toughness requirements of Appendix G of 10CFR50. This requires a projection of the measured reduction in C upper shelf energy to the vessel wall using knowledge of the energy y and spatial distribution of the neutron flux and the dependence of C upper y shelf energy on the neutron fluence. (2) Develop heatup and cooldown curves to describe the operational ) limitations for selected periods of time. This requires a projection of the crasured shift in RT to the vessel wall using knowledge of the dependence NDT cf the shift in RT n the neutron fluence and the energy and spatial dis-NDT tribution of the neutron flux. The energy and spatial distribution of the neutron flux for Donald C. Cook Unit No. 2 was calculated for Capsule Y with a discrete ordinates I transport code. This analysis, also applicable to Capsule T, predicted that l the lead factor (ratio of fast flux at the capsule location to the maximum i pressure vessel flux) was 3.57 at the capsule centerline, 3.74 for the core-side Charpy layer, and 2.% for the vessel-side Charpy layer (see Table l l III). This analysis also predicted that the fast flux at the 1/4T and 3/4T positions in the 8.5-in. pressure vessel wall would be 54% and 10%, respec-tively, of that at the vessel I.D. However, in this report the projection of Capsule Y results to the pressure vessel wall utilizes the more conserva-tive attenuation figures of 60% and 15% for the 1/4T and 3/4T positions to 31 w --rw --v-ww--r- --w-w,- w w ---w w

allow for the increased fraction of neutrons which might accrue in the 0.1 to 1.0 MeV range in deep penetration situations. A method for estimating the increase in RT as a function of neutron NDT fluence and chemistry is given in Regulatory Guide 1.99, Revision 1 [83. However, the Guide also permits interpolation between credible surveillance dsta and extrapolation by extending the response curves parallel to the Guide trend curves. The data from Capsules T and Y are deemed to be cred-ible because (1) the surveillance materials are judged to be controlling with regard to radiation damage, (2) the scatter in the transverse plate and cold metal Charpy data is small, and (3) the changes in yield strength are consistent with the Charpy curve shifts. Except for the longitudinal plate material, the slopes of the response curves constructed in Figure 8 are less than the square root of fluence utilized in Regulatory Guide 1.99. Although recent work [73 indicates that the square root of fluence dependence may be too high, the projected responses of the Donald C. Cook Unit No. 2 vessel bsitline materials are based on the trend curves of Figure 8 which were con-structed in accordance with Regulatory Guide 1.99 procedures. The Donald C. Cook Unit No. 2 vessel plate surveillance material is m:re sensitive than the weld metal and HAZ surveillance materials to irradt-Etion embrittlement. Since the unirradiated values of RT f r the inter-NDT mediate shell plate C5521-2 is higher than those of the weld and HAZ mate-rials [163, the beltline region plate material is projected to control the i adjusted value of RT throu2h the 32 EFPY design life of Donald C. Cook NDT for 12 and 32 EFPY l Unit No. 2. A summary of the projected values of RTNDT of cperation of Donald C. Cook Unit No. 2, assuming that the Cycle 3 low leakage core configuration will continue to be used, is presented in Table XI. 32

600 l l Ollll0 lll 0ll$ M l l l[ ~ ~ Tl i"l!h ll; ll Ih lU5 l '! l ~- l 11p I! lij

lll [i g.

-~ ~~~ i 1 n ji.l 400 J ['pF l - ; q% Reg. Guide 1.99 p j. u.

dn o

e l jj,l; 1 Upper Limit +._ 2 pli 3 .l. (, : : -lQ,I iy j!l i j q++: g i s. g, e t !! E H z: N l i Ill!E,W i..:jj 1 i L; .J .I I a I . ',,j-1 ? l l l E l I _ g4j; ' 'f f 7 j l 200 i q s l _ 00 ' O.g j_ 7

i I '

u 5 l 0 *\\d [,,, } j 1J i .Y_$ - g i j g l ill J ce v:$ j q [, 100 ,,_ (p ,-g ,f-y.. {.! .j, 7 w ,7,g;t,,..g ...;.g w a p .a I ;- l g,,, u fo M[ ~ ..! :;1. ..qh _.p, r a g k Il y r lf l . 0%]l-l 7, _ - i y u ,i y ' Z-p [1, 1 4 c g 60 '---- y = q ye - __;,, ~f

. p

.J. Code:. tB G f j! e =- 4-l O Trans. Plate I j ' -[}-,{ Il 'j ~ y .g 40 -jl ~ Elljll. T ~~ ~~~~ a Long. P1 ate j ~~ E.,' s PI<- e HAZ Material pd 7 4r Weld Metal = 5 3 llilj. i 2 E i, -{ 3 F5r--: l l l j j'[ i ~ 20 [ 2, 'a 5 c3 l9 - t 19 6 x 10 2 x 1017 1018 > 10 Neutron Fluence, n/cm2 (E > 1 MeV) FIGURE 8. EFFECT OF flEUTR0fl FLUENCE Oil RTflDT SHIFT, D0flALD C. COOK UNIT NO. 2

TABLE XI PROJECTED VALUES OF RT FOR DONALD C. COOK UNIT NO. 2 NDT Initial EFPY P.V. Material Location RTNDT Fluence (*) ARTNDT Adj. RTNDT 12 Plate C5521-2 I.D. 58'F(b) 6.3 x 10 98 156 18 18 1/4T 58'F 3.8 x 10 87 145 1 3/4T 58'F 9.4 x 10 47 105 HAZ Material I.D. 20*F(b) 6.3 x 10 74 94 18 18 1/4T 20*F 3.8 x 10 61 81 1 3/4T 20*F 9.4 x 10 33 53 Weld Metal I.D. 0*F(c) 6.3 x 10 53 53 18 18 1/4T 0*F 3.8 x 10 46 46 1 3/4T 0*F 9.4 x 10 26 26 32 Plate C5521-2 I.D. 58'F(b) 1.6 x 10 150 208 19 18 1/4T 58'F 9.8 x 10 117 175 10 3/4T 58'F 2.5 x 10 77 13 5 HAZ Material I.D. 20*F(b) 1.6 x 10 118 13 8 19 18 1/4T 20*F 9.8 x 10 92 112 18 3/4T 20*F 2.5 x 10 52 72 I Weld Metal I.D. 0*F(c) 1.6 x 10 85 85 19 18 1/4T 0*F 9.S x 10 66 66 8 3/4T 0*F 2.5 x 10 47 41 2 (a) Neutrons /cm, E > 1 MeV. (b) Reference 16. (c) Estimated per Reference 17. 34

A method for estimating the reduction in C upper shelf energy as a y function of neutron fluence is also given in Regulatory Guide 1.99, Revision 1 [8]. The results from Capsules T and Y are compared to a portion of Figure 2 of the Regulatory Guide 1.99, Revision 1, in Figure 9. Although the shelf energy responses of the plate and weld surveillance materials from Capsules T and Y fal~. below them, the predictive trend curves of Regulatory Guide 1.99, Revision 1, will be used in this analysis for conservatism. A 4 r sponse curve.has been drawn through the HAZ data since these results fall above the plate trend curve. Referring to the conservative trend curves for 0.05% Cu weld metal and 0.15% Cu plate, as well as the HAZ response curve, the projected C shelf y energies of the vessel materials are as follows: o Plate C5521-2 (Unirradiated Cv Shelf = 86 ft-lb) 32 EFPY at I.D. -- 63 ft-lb (27% reduction) 32 EFPY at 1/4T -- 65 ft-lb (24% reduction) 32 EFPY at 3/4T -- 71 ft-lb (17% reduction) o Weld Metal (Unirradiated Cv Shelf = 75 ft-1b) 32 EFPY at I.D. -- 59 ft-lb (21% reduction) 32 EFPY at 1/4T -- 61 ft-lb (19% reduction) 32 EFPY at 3/4T -- 64 ft-lb (14% reduction) o HAZ Material (Unirradiated Cv Shelf = 122 ft-lb) ~ 32 EFPY at I.D. -- 83 ft-lb (32% reduction) 32 EFPY at 1/4T -- 87 ft-lb (29% reduction) 32 EFPY at 3/4T -- 99 ft-lb (19% reduction) These projections indicate that the core beltline materials in the Donald C. Cook Unit No. 2 pressure vessel material will retain adequate shelf tough-ness throughout the 32 EFPY design lifetime. 35

60 hl. l 0l I 4 4 ! lN : I l h W

in i

i int ": h L -r-t! L-al ~ ~ W k ~ l l-! 40 - (- h ~ Reg. Guide 1.99 [ !!.s.;h "~ ! l. 1 ~~~~,,. .L k l ]lil: Upper Limit 4 22 y = I j $, _,,,:,.9--,__. d g c .i . + .l'[.I n g i i i :._ 7 l : r s. l1c n 'J f I.l 20 [ -[ j ( tar ~i = Reg. Guide 1.99 .:_1. :_: + I ---~ 2 r o, b 0.15% Cu Plate _- I,, 0 i c j [] w f,-- Lj j'lu! 0.05% Cu Weld 10 h'

p N 5 g

+ Code: = E ~ o Trans. Plate j I[!I[ I-5 { , = 4 H h"j ? _ i M 9 a Long. Plate { g. i N 6 H h y E 5 FEF i [ + Weld Metal [ l 9 HAZ Material ll 8 7 = ~ : {

[

11. 1 . f di 4 4 j l ;..;t I t = :.= = :

=.

n ]f j p}:j ~ ~ j j t_p i 1 :_ : :_ n it j 2 2 x 1017 1018 1019 6 x 10l9 Neutron Fluence, n/cm2 (E > 1 MeV) l FIGURE 9. DEPENDEllCE OF Cy UPPER SilELF EllERGY ON NEUTRON FLUENCE, DONALD C. COOK UNIT NO. 2

The current Donald C. Cook Unit No. 2 reactor vessel surveillance pro-gram removal schedule, revised to conform to ASTM E 185-79 [9], is summa-rized in Table XII. There are six capsules remaining in the vessel, of which three are standbys. TABLE XII REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE [16] DONALD C. COOK UNIT NO. 2 WOL Removal Equivalent Vessel C psule Material Time Fluence T Weld Metal 1.08 EFPY "' 3.4 EFPY at I.D. I Y Weld Metal 3.24 EFPY(b) 11 EFPY at I.D. X Trans. Plate 5 EFPY E.O.L. at 1/4T U Weld Metal 9 EFPY E.O.L. at I.D. .S Trans. Plate 32 EFPY E.O.L. at I.D. Y Trans. Plate Standby W Trans. Plate Standby Z Weld Metal Standby (a) Removed after core cycle 1. (b.) Removed after core cycle 3. 37

t VI. HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION OF DONALD C. COOK UNIT NO. 2 Donald C. Cook Unit No. 2 is a 3391 Mw Pressurized water reactor op-t carated by Indiana and Michigan Electric Company. The unit has been provided with a reactor vessel material surveillance program as required by 10CFR50, i' Appendix H. The second surveillance capsule (Capsule Y) was removed during the i 1982 refueling outage. This capsule was tested as described in earlier sec-tions of this report. In summary, these test results indicate that: (1) The RT f the surveillance plate material in Capsule Y in-NDT 18 creased 100*F as a result of exposure to a neutron fluence of 7.0 x 10 n utrons/cm2 (E > 1 MeV). 4 (2) Based on an analysis of the dosimeters in Capsule Y, the vessel eall fluence at the I.D. was 1.88 x 10 neutrons /cm2 (E > 1 MeV) at the 18 3 time of its removal. (3) The maximum RT after 12 effective full power years (EFPY) of NDT eperation was predicted to be 145'F at the 1/4T and 105'F at the 3/4T vessel tall locations, as controlled by the core beltline shell plate. These pro-jections are lower than those resulting from the evaluation of the data from capsule T [143 because of the introduction of the low leakage core. (4) The maximum RT after 32 EFPY of operation was predicted to be NDT i 175'F at the 1/4T and 135'F at the 3/4T vessel wall locations, as controlled 1 by the core beltline shell plate. These predictions are also lower than i those made earlier [143 for the reason given above. 38 i

The Unit No. 2 heatup and cooldown limit curves for 12 EFPY and 32 EFPY have been computed on the bases of (3) and (4) above. The following pressure vessel constants were employed as input data in this analysis: 86.50 in., including cladding Vessel Inner Radius, r1 = Yessel Outer Radius, ro 95.2 in. = Operating Pressure, Po 2235 psig = Initial Temperature, To 70*F = Final Temperature, Tf 550*F = 6 Effective Coolant Flow Rate, Q 134.6 x 10 lb,/hr = Effective Flow Area, A 26.72 ft = Effective Hydraulic Diameter, D 15.05 in. = The SwRI computer program calculates the allowable pressure over the temperature range 70*F - 550*F such that the reference stress intensity fac-tor, KIR, is always greater than the sum of twice Kyp (pressure induced) and kit (thermal gradient induced) as dictated by Appendix G of the Code [2]. The current version of the SwRI program incorporates the physical property data specified by Appendix I of the Code through the 1982 Summer Adenda. The changes in thermal conductivity code allowables made in the early 1980's r&duced the calculated allowable pressure at coolant temperatures below ab'out 200*F from that obtained when using the previously specified values. Heatup curves were computed for a heatup rate of 100*F/hr. Since lower rates tend to raise the curve in the central region, these curves ap-ply to all heating rates up to 100*F/hr. Cooldown curves were computed for c:oldown rates of 0*F/hr (steady state), 20*F/hr, 40*F/hr, 60*F/hr, and 100*F/hr. The 20*F/hr curve would apply to cooldown rates up to 20*F/hr; 39

the 40*F/hr curve would apply to rates up to 40*F/hr; the 60*F/hr curve could apply to rates up to 60*F/hr; the 100*F/hr curve would apply to rates up to 100*F/hr. The unit No. 2 heatup and cooldown curves for up to 12 EFPY are given ] in Figures 10 and 11; those for up to 32 EFFY are given in Figures 12 and 13. i 40

2600 O l ! !!!i REACTOR COOLANT SYSTEM HEATUP LIMITATIONS A."-j : ! ii !!!. ei 1 ib' CLt0ED FOR POSSIBE INSTRGENT ERROR) l i !! ' [j [jj! 'lj i !! i -i~' i7 .i ! l. . !]I[-""I r~ FOR FIRST 12 EFFECTIVE FULL POER j 'i~' ~" (/j l I ~ i l j ~j i i m ![1 i iii .PLICABLE(MARGINS OF 60 PSIG AND 10*F ARE IN-2400 i T !I . YEARS. b ti y ;H l S i ' " " " ' " " " " ' ' ' " ' " * " ' "y~ i ~ "l 1. I h 1 ')!!!!lj ]' i $j !i!' !" ' ili i "i l i l+h I l LEAK TEST LIMIT i ! c 1 - i i i Ill it -4 l !! ! jj ] l!! ii!: ! t! I 2 I I I i ti i -o 2000 { MATERIAL PROPERTY BASIS f [ h Fj y Il J j p [JE y 1800 L lj u.! L u.] ! 12 Ji ? i I ! j gj- [ti. j ] .u [][ g LBASE E TAL CU = 0.14t i l j N i I ACCH l TABWil ll ] n:. INITIAL RT DT = 58'r j i a 1600 12 EFPY RT DT (1/4T) = 145 'F' UNACCEPTABLE' hrn FL4' tr 1 7M f OPERATIW [~ i i IIII l N _/ j [ 0PERATION UII!i j [

' j o.

1400 l [ i iti g i }. j 1. 1 ?- l l i !i ! !! i j! !ll'!l! I ! I l ! ! l b 3 m 1200 -/ i j u t -y i j lj j p i jr ?; j i l l RA M

V U l j~

i l ll l l l / ~ ~ ~~ ' [j j i ' l 1000 LIMIT FOR HEATUP RATES i igf;j. t i LP TO 100*Pjtyt p -- 7: i b !L g [ h. /. ' N '/ u 800 p T i I lli. il I N p 4 1 6 l i j l l t i i t [3 600 ~N l 1 ~ ih:- i b sf { l LIMIT j ~l j I~l !I III l CRITICALITY l i. Li j j .j i j ll} ljj]hl iil i 400 ,li a -lii l

l il!! :!!il l I

i j m i IfI !! 1 I i !f h,f f h*'f I i I' i 200 ~ ~ hi !jij j l [ i ] ll i lj ell!! ![-" ! '[i I I ' i' "i' + l T l' 0 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR C00LNIT SYSTEM TEFFERATURE (*F) FIGURE 10. REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS 100 F/HR RATE, CRITICALITY LIMIT AND HYDROSTATIC TEST LIMIT,12 EFPY

2600 REACTOR COOLANT SYSTEM C00LDO N LIMITATIONS j i I. I !. !ll !!!! ....li ii - !j.; L! !! i "' 'l -' APPLICABE FOR FIRST 12 EFFECTIVE FULL PWERl I I I kL II i i ! li! k! I' h b U .! I I !.ll_ E 2400 CLUDED FOR POSSIBLE INSTRlfENT ERROR.) { l l ! . l ! 'i ! !ll i II O YEARS. (MARGINS OF 60 PSIG AND 10*F ARE IN,I i I !I I !' !. 2200 j j [ j l l j j jj j l E i;,g ! ;; ; i;

j; ;;
g
g

!;!!!I 'I I l ~i i }I hu j .j r ... L L,b

3

!}i!!!!iiii!il!IIIIIIII!!:1 IIl!!i Iii! i II :! I 2000 i ii n 3l j ! ii ii :l.!

jj

-1 i i .m MATERIAL PROPERTY BASIS

J9 I'.3 i !t

..!.!, li l. I i Eh 58'F OPERATION li !.:.I l l I BASE PETAL CU = 0.14% I .I' I ..}l l }j 'l!'! 1800 4 !!i UNACCEPTABLE I' W 12 EFPY RTIOT ((3/4T) = 105 'F

/

~j ! , l l !i i INITI AL RTNDT = /4T) = 145 'F i!! 1600 i 1 p q jj j { r j ; - t "'i ! j l " j$l! o t j- - r

n p

'l 1400 - '! hlMI" )I l --5 h r I I I h -{[!- ... i i ii .,!l[l l ['. l / il i li i i I /~~ ' 1. I 3 j ! L . l!!. a mg 1200 m i PRESSURE-T N / ACCEPTABLE ! l l 1000 A: - i I i LIMITS f ' OPERATION l l i ll!. b . i!.J N / j, -h i i; i' N l l [ I . li! ~ nr i i 800 .r H h' i 600 l lIR i 1.

[]

@.L-l mmc i F 0- m 400 } 20 " $M i l l - [ij-] j li: i tii g$ t I ( j l l ill: i40-i 200 1 dh I a + - l ~+ i < N I$ Y !! ! illlllI!!d! jI. il I l I,I h! !T kI L O!!n I L .i 0 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEW ERATLEE ( F) I FIGURE 11. REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS C00LDOWil RATES, 12 EFPY

2600 REACTOR COOLNIT SYSTEM 11EATUP LIMITATIONS AP- ;jil I : 'l: 4 'ly g,,ili g

hi a, a; a ji

PLICABLE FOR FIRST 32 EFFECTIVE FULL POWER

p E AK TESTI: i'! m' :!F !p, iii lN ".

[ !l iih 2400 } YEARS.(IMRGINS.0F 60 PSIG NJD 10*F ARE IN- ['i'i$ LIMIT 7 I b Iii fl! hl l!! 'I! !I! !!U 'fi Ul l!4 if

CLUDED FOR POSSIBLE INSTRtfENT ERROR) j j y q j pg ;'

[ j Ll ll l llj l :i'

!! yli ij[ j[ !i; ij jiij 2200 "a m

= P=i i ~! m e ta r"j l ! I II 'l y I i dj 4 m 4 --lI.i!;-liii.' i l] l l:li i I 2000 - (;.I

![iiiii bl hi {H!!!!: q ! !! j!!!iill j

'l i l l in ilil 7 % "k pi } ,1 ! l ~ l (( J '{l l ] i l [ i lj g i I !,..lt.: ! n,: t 77 J l i r! i ir irj l!! r ~t .r T ~! -T gi 'e r _o MATERIAL PROPERTY BASIS m I i l l I-' 1 ! l.l il! !} !

!!I:.. F

.!: h

:ii Y l

S 1800 BASE METAL CU = 0.14% 1 jj [f{' l pp I l !!"i[ ACCEPTABLE F T i !.j [if }j0PERATION' ~~ [li.~ Y ~ji i Il i }g ~ INITIAL RTNDT = 58"P 'i! 'j i 1600 32 EFPY RTNDT (1/4T) =$ 175'F d'! UNACCEPTABLE'~i l [L "i i S l:j F'; j' im e U d J d ~p i d IN"! L F-(3/4T) = 135'F - }- 9 0PERAT10N it: r m Id Il -" u 1400 4LU'I!M'"' i"l [l u !! !i: ! l ill l I II I l i ! jj 'j l I. 5 .'I lli fl ii Lii 'i Il if f i '. l l i I i i l i l' l i ! ih i l 'I I f i f i / i a- = j 7 ] ]r ! Q l F ~Tl.i!' b; 1200 ~ w j] l l l j,j h;i: j j -l l i j / - /' t 7 jp p p 3 ./ r I m i i /p [~ ~ ~ il [ e 1000 p p j ii ii i; qll [. j Iil PRESSURE-TEMPERATURE fj l 0 i h 3 ill l [1iml] jy ifi ] } [ ' jp LItilT FOR llEATUP RATES IFj [ 800 g !rhH ! } 4 i d E 1 :i t i 7; ! UP TO 100*FAR ! ! l / V i!

1 i /

N I tijj E l i n' I' 8 l"' i ! l l i i ! I'~ 3' \\ ll l j j i i ' }.lf l j 'CRITICALITYi!!! ijj} f.i h l y" i 600 .c p ~ ~ t l # b b l {h; l g[ l i l l i (g h 400 j; } [ [gl Jyi 71!7]i~ [' Tii! [Ji!- l r,,1 ~ j j j.; j j[ l b EN i I

!I !

b lh b N l N! b U M E N 200 r ] L I l0 'b Il:b i t-i 0 60 100 1 50 200 250 300 350 400 450 AVERAGE REACTOR C00UWF SYSTEM TEffERATURE (*F) i FIGURE 12. REACTOR C00LAftT SYSTEN PRESSURE-TEftPERATURE LIMITS VERSUS 100 F/llR RATE, CRITICALITY LIMIT AtlD llYDR0 STATIC TEST LIMIT, 32 EFPY

2600 I !!!. d !![.h<lOi!kh

REACTOR' COOLANT SYSTEM C00LDOWN' LIMITATIONS.

i !il !!I! I!!I l '.i l ! '!d.!' !!!! Hi . APPLICABLE FOR FIRST 32 EFFECTIVE FULL POWER Li !.! i! li! illd. 1 (iAL Hl' I!l! :!!' !U :ll: H!i.I I.l!?;!t f;il 2400 7 YEARS. (MARGINS OF 60 PSIG AND 10'F ARE IN-l i!ll i! l !* HH Hl il l ![!i L CLU0ED FOR POSSIBLE INSTRlNENT ERROR.) i if f I ' i i'. F. !!L I l j i 'i I ! li jl['! l! {h d[! ! I!; I . l[ i l l 'd '{ }ij i 2200 j "~ 'T ". n i -li ;id E!!!!NiiT[l'jljij!d}}O!!!!!!l!!!j..I..!!!!4!Ifl.i [ ! h m i!Ii l" !R #i "!i '!II ii! fI !!'i l! 1 i E ' L iJ i I I I I 2000 t! Il!i Illll! (V. L f%TERIAL PROPERTY BASIS 1 i I I i !. I i ~ I i ! IU I i l T E jif 1800 " S BASE N TAL CU = 0.14% l { l

INITIAL RTNDT = 58'F

[ [] / l j jiji E 1600 1 32 EFPY RTNDT ((1/4T) = 175 'F r ~ -! O!, ~ 3/ T).=J3 'F l LNACCEPTABLE . ] ! .I J 1400 M J a i: j 1! !J I I m t j j l I I / ,rn-} l [~~ f{' i i 4 a ^ t;. 1200 i

b-,I: h ni li I

{b r ACCEPTABLE ~ l i 3 [ l / - OPERATION !!p j~ k t b t f i 1 i F PRESSURE-TEffERAllJRE 8 1000 L f l (! !N l !!!! {- l lO IMITSg'. L i -[ E 800 h pjl 1 J 4 ~T "" N'" H E 7 ll - 'I l l ! il [3 !!!!!!l!Ul!!!!!!!! ili j. i i: L -f $ h I I N i I L 2 ! qh:, y !!- 600 .C00LDOWN 3 I .h$ I f i RATE..WHR N y i I i l l i M 0-h - s c 1 400 H - n+M 2 i il r T 7 i t I-3 I 4 gj] q ji i '. 3 -i 11! 200 100%- j' ih. lNNI!00Ohi i.i i, I l.li ! L I . I I i t .i i E I 4 i,. 0 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR C00LAllT SYSTEM TEffERATURE (*F) FIGURE 13. REACTOR C00LAtlT SYSTEM PRESSURE-TEf1PERATURE LIi1ITS VERSUS C00LDOWil RATES, 32 EFPY

'i i i VII. REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, " Licensing of Production and Utilization Facilities." 2. ASME Boiler and Pressure Vessel Code, Section III, " Nuclear Power Plant Components." 3. ASTM E 208-81, " Standard Method for Conducting Drop-Weight Test to Determine Ni-Ductility Transition Temperture of Ferritic Steels," 1982 j Annual Book of ASTM Standards. i 4. Steele, L. E., and Serpan, C. Z., Jr., " Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970. 1 5. Steele, L. E., " Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels," International Atomic Energy Agency, Technical Reports i Series No. 163, 1975. 6. ASME Boiler and Pressure Vessel Code, Section XI, " Rules for Inservice Inspection of Nuclear Pcler Plant Components," 1974 Edition. i m 7. Randall, P. N., "NRC Perspective of Safety and Licensing Issues Regarding Reactor Vessel Steel Embrittlement - Criteria for Trend j Curve Development," presented at the American Nuclear Society Annual l Meeting, Detroit, Michigan, June 14, 1983. l 8. Regulatory Guide 1.99, Revision 1, Office of Standards Development, l U.S. Nuclear Regulatory Commission, April 1977. 9. ASTM E 185-79, " Standard Recommended Practice for Surveillance Tests } for Nuclear Reactor Vessels," 1981 Annual Book of ASTM Standards. 10. ASTM E 399-81, " Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials," 1982 Annual Book of ASTM Standards. 11. ASTM E 813-81, " Standard Test Method for JIc, a Measure of Fracture Toughness," 1982 Annual Book of ASTM Standards." 12. "American Electric Power Service Corporation Donald C. Cook Unit j, No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8512, j November 1975. 13. Letter. F. Noon of Westinghouse to J. R. Jensen of the American { Electric Power Service Corporation, Document AEP-80-528, March 19, 1980. l 14. Norris, E. B., " Reactor Vessel Material Surveillance Program for j Donald C. Cook Unit No. 2; Analysis of Capsule T," SwRI Report j 06-5928, September 16, 1981. i l 45 i -n-n-

i 15. Magurno, B. A., "ENDF/B-IV Dosimetry File," BNL-NCS-50446 (ENDF-212), Brookhaven National Lab, April 1975. 16. Donald C. Cook Unit No. 2 Technical Specifications. 17. US NRC Standard Review Plan, NUREG-75/087, Section 5.3.2, Pressure-Temperature Limits, November 24, 1975. 4 4 e d i 1 46 i ~_-y __._---e.,__, ._y__.___e_---n-- -w.m_+__-,_ -e,--,__p -,re._,--%_n

i { APPENDIX A TENSILE TEST RECORDS 1 \\ i l l 47

0 T e i' N 3 1 Southwest Research Institute ' Department of Materials Sciences TENSILE TEST DATA SHEET 3 1 Test No. T-I Es t. U. T. S. psi Project No. Of -77 ud f Spec. No. MT-#3 Initial G. L. /,oo o in. Machine No. 22 <, P Temperature 2. I O 'F Initial Dia. .2 Co in. Date 7- / 5~- S ? / ,00fi At/,vkInitial Thickne ss Strain Rate in. Initial Area _OeM imY Initial Width in. r Top Temperature 2 f '2 'T Maximum Load TM lb / Bottom Temperature 210 'F 0.2% offset Load 3G7 lb Final Gage Length 1, t 43 in. 0.02% offset Load Ib Final Diameter . I"I I in. Upper Yield Point Ib Final Area .O24 % in. 2 ( / kb P81 km. U. T. S. = d Ibj *p' fa Initial Area F,. d a L ee a G W Si

0. 25. Meet Md =)l SO psi

0. 2% Y. S.

= Initial Area

0. 02r. Nset bad

0. 02% Y. S.

p,g = Initial Area Upper Y.S. poi = = Initial Area = x 100 = 14.3 % Elongation ,g g / \\ pil,1 ^ N* * * * * *

A '" x 100 =

O *! % R. A. = mL S,g..t.r.. l ~

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