Text

Reactor Vessel Irradiation Surveillance Program Analysis of Capsule 8, Final Rept
	ML20100N109
Person / Time
Site:	Dresden, Quad Cities, 05000000
Issue date:	08/31/1984
From:	Lindholm U, Norris E; SOUTHWEST RESEARCH INSTITUTE
To:	;
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ML17195A753	List: ML20100N109; ML20100N115;
References
	SWRI-7857, NUDOCS 8504180318
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{{#Wiki_filter:SwRI-7857 {l i QUAD CITIES NUCLEAR POWER STATION UNIT 1 REACTOR VESSEL IRRADIATION l SURVEILLANCE PROGRAM ANALYSIS OF CAPSULE NO. 8 by E. B. Norris FINAL REPORT SwRI Project No. 06-7857 l for l Commonwealth Edison Company Nuclear Stations Division P. O. Box 767, Room 1230 Edison Chicago, Illinois 60690 August 1984 = SOUTHWEST RESEARCH INSTITUTE ,[g// SAN AN(ONIO HOUSTON ?E""E80888888,, PDR

SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 QUAD CITIES NUCLEAR POWER STATION UNIT 1 REACTOR VESSEL IRRADIATION SURVEILLANCE PROGRAM ANALYSIS OF CAPSULE NO. 8 by E. B. Norris i FINAL REPORT SwRI Project No. 06-7857 1 for Commonwealth Edison Company Nuclear Stations Division P. O. Box 767, Room 1230 Edison Chicago, Illinois 60690 i 1 August 1984 Approved: f U. S. Lindholm, Director Department of Materials Sciences

TABLE OF CONTENTS Page LIST OF TABLES iii l LIST OF FIGURES iv I.

SUMMARY

OF RESULTS 1 II. 'BACKROUND 2 III.- DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM 4 I V.' TESTING OF IRRADIATED SPECIPENS 5 . A. . Shipment, Opening, and Inspection of Capsule 5 B. Neutron Dosimetry 5 C. Mechanical Property Tests 15 D. Check Chemical Analyses 18 VI. HEATUP AND C00LDOWN LIMIT CURVES FOR OPERATION OF l QUAD CITIES UNIT NO.1 27 VI. REFERENCES 29 ' APPENDIX A - SURVEILLANCE SPECIPEN DRAlIINGS AND CAPSULE PHOTOGRAPHS 31 APPENDIS B-PHOTOGRAPl!S OF TESTED SPECIMENS AND TENSILE TEST RECORDS 34 i c t t l

LIST OF TABLES L Table Page I Operations Summary - Quad Cities 1 7 II Results of Discrete _ Ordinates Sn Transport Analysis, Quad Cities 1 and 2 12 III Summary of Neutron Dosimetry Results, Quad Cities 1, Vessel Wall Surveillance Basket No.8-(215*) 14 IV Hardness Properties of Surveillance Materials, Quad Cities 1 17 Y Charpy V-Notch Impact Data on Surveillance Specimens Removed from Quad Cities Unit 1 20 VI Effect of Irradiation on the Charpy V-Notch Upper Shelf Energies of the Quad Cities Unit 1 Vessel Surveillance Materials Basket No. 8 (215* ) 24 4 VII Tensile Properties of Surveillance Materials Basket No. 8 (215'F), Quad Cities 1 25 VIII Check Chemical Analysis Results, Basket No. 8, Quad Cities 1 26 1 ] fii i

~.. _~ I LIST OF FIGURES Figure Page 1. Discrete Ordinates Calculational Models, Quad Cities Unit 1 11 2-Calculated ~ Fast Flux (E > 1 MeV) Distribution, Quad Cities Unit 1 13 3 Vessel Wall Fluence as a Function of Operation of Quad Cities 1 16 i '4 Charpy Y-Notch Impact Properties of Quad Cities L Unit 1 Surveillance Base Plate 21 5 Charpy V-Notch. Impact Properties of Quad Cities 2 Unit -1 Surveillance Electroslag Weld Metal 22 6 ' Charpy V-Notch Impact Properties of Quad Cities ]. Unit 1 Surveillance Electroslag Weld Heat-Affected Zone Material 23 l i 7 - Pressure-Temperature Limits for Levels 'A & B Service for Up to 10 EFPY 28 i i f r 4 9 l l I i i -l IV

I.

SUMMARY

OF RESULTS The analysis of the vessel material surveilla' nce capsule basket No. 8 removed from the.215* position in the Quad Cities 1 pressure vessel during the 1982 refueling outage led to the following results. (1) Based on a calculated neutron spectral distribution, 16 the capsule received a fast fluence of 5.5 x 10 2 n/cm, E > 1 MeV, in 6.64 Effective Full Power Years (EFPY)ofoperation. (2) There were small positive or negative shifts in RTNDT of the surveillance materials as a result of the above exposure. These data are consistent with the level of the exposure and may reflect data scatter rather than real changes in the RTNDT* (3) The Charpy upper shelf energy of the electroslag weld and heat-affected zone (HAZ) surveillance materials l decreased, but that of the base plate material increased i after exposure. This may also reflect data scatter, i (4) Based on a calculated neutron spectral and spatial distribution, the maximum fluence rate (flux) inci-dent on the pressure vessel wall is 3.46 x 108 n/cm? sec, E > 1 MeV. Therefore, the projected maxi- / mum vessel wall neutron fluence after 32 EFPY of operation is 3.5 x 1017, E > 1 MeV. f l l l

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 duc-tility temperature (RTNDT) presented.in Appendix G, " Protection Against Non-ductil'e Failure," of Section' III of the ASME Code [2].

Further, the materials in the beltline region of the reactor vessel must be moni-tored for radiation-induced changes in RTNDT per the requirements of Appendix H, " Reactor Vessel Material Surveillance Program Requirements,"

- of 10CFR50.- The RTNDT is defined in Paragraph NB-2331 of Section III of the - ASME Code as the highest of the following temperatures: (1) Drop-weight nil ductility temperature (Dti-NDT) { per ASTM E 208 [3]' (2)- 60 deg F.below the 50 ft-lb Charpy V-notch (Cy) temperature (3) . 60 deg F below the 35 mil Cy temperature. 1 -The RTNDT must be established for all materials, including weld metal and heat-affected zone (HAZ) material as well as base plates and forgings, which comprise the reactor coolant pressure boundary. t It is well established that-ferritic materials undergo an increase in strength and hardness and a decrease in dtetility and toughness when exposed to neutron fluences in excess of 1017~ neutrons per cm2 (E > 1 MeV) [4]. Also, it has been established that tramp elements, particular-g -- ly copper.;adriKkel ci~n affect the radiation embrittlement response of -ferritic materials [5-7].- The relationship between. increase in RTNDT_and - copper content is defined in Regulatory Guide 1.99. Although this docu-l' ment -is being revised by the NRC to reflect a more.recent evaluation of ~ ~ neutron embrittlement data from reactof's~ rvleillinciipfogriisi,' estimates u l_ of shifts in.RTNDTLin this report are based on the current Revision 1 of Regulatory Guide 1.99 [8]. s en v-- n ---,w, s

3 In general, the only ferritic pressure boundary materials in a nu-clear plant which are expected to receive a fluence sufficient to affect RTHDT are those which are located in the core beltline region of the re-l actor pressure vessel. Therefore, material surveillance programs include specimens machined from or representative of the plate or forging material and weldments which are located in the core beltline region of high neu-tron flux density. ASTM 'E 185 [9] describes the recommended practice for monitoring and evaluating the radiation-induced changes occurring in the mecharical properties of pressure vessel beltline material. General Electric has provided such a surveillance program for the Quad Cities 1 Nuclear Power Station. Six baskets of test specimens were located near the I.D. surface of the pressure vessel where the fast neu-tron flux density is slightly higher than that at the adjacent vessel wall surface. However, because of azimuthal variations in neutron flux density, these vessel wall basket-fluences may lead or lag the maximum ~ vessel fluence in a corresponding exposure period. Three sets of accel-erated-exposure baskets were included-to augment the surveillance pro-gram. These baskets, located near the core, would have much larger lead factors to provide early infonnation on icng-term behavior of the sur-veillance materials. The baskets also contain several dosimeter materials for experimentally detennining the average neutron flux density at each capsule location during the exposure period. The Quad Cities-1 mechanical property surveillance capsules also in-clude tensile specimens as recomended by ASTM E 185. At the present time, irradiated tensile properties are used to indicate that the materials . tested continue to meet the requirements of the appropriate material speci-fication and to judge credibility of the surveillance capsule Charpy data. This report describes'the results obtained from testing the contents of Basket No. 8 from the 215' position in Quad Cities' Unit 1 (the second. ~ I vessel wall surveillance capsule tested). The results were analyze _d to dd-- termine the radiation-induced chanaes in the mechanical properties of'the-l surveillance inaterials at the time of the refueling outage and the fluence l rate atfthe vessel wall. l w. l c6 .l / 1 9 m m -.w. ,__w,

f 4 III. DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM The G.E. vessel material surveillance program is described in de-tail in NED0-10115 [10]. Six vessel wall baskets, each located opposite the vertical. center of the core and containing encapsulated Charpy V-notch and tensile specimens, were placed in the Quad Cities 1 vessel at the pres-sure vessel wall prior to startup. In addition, three baskets were placed near the core. One vessel wall basket (No. 3 ) and two near-core baskets (Nos. 2 and 4) had been previously removed, tested, and the results re-ported [11,12]. The vessel wall Basket No. 8-, removed during the 1982 refueling outage from the 215* position, contained two impact capsules (each hold-ing 12 Charpy V-notch specimens plus an iron, a nickel, and a copper flux wire) and three tensile capsules (each holding two tensile specimens). The capsules did not contain thermal monitors. Drawings of the impact and tensile specimens, and photographs of the impact and tensile capsules and the surveillance basket are snown in Appendix A. According to NED0-10115 [10],-the Quad Cities 1 base metal specimens were made from flat slabs cut parallel to and one-quarter plate thickness from both of the plate surfaces, and were machined with their longitudinal axes parallel to the plate rolling direction. The electroslag weld (ESW) metal and ESW heat-affected zone (HAZ) specimens were cut from a test weld representing a vessel welded joint which had been fabricated from vessel base material, the weld and HAZ Charpy V-notch specimens being oriented with the long axis transverse to the weld direction and the weld HAZ 8 ensile specimens with the long axis taken parallel to the weld direction. The notches of all Charpy V-notch specimens were perpendicular to the orig-inal plate surfaces. L The mechanical properties of unirradiated (baseline) surveillance specimens for the Quad Cities 1 vessel were determined and reported at the time of the testing and analysis of the specimens from the first baskets [13]. These data were used to determine the shifts in RTNDT exp' rienced by the materials in the 215* basket No. 8. e t

1: 5 t IV.. TESTING OF IRRADIATED SPECIMENS The capsule shipment, capsule opening, specimen testing, and report-ing of results were carried out in accordance with the Project Plan for the Quad Cities 1 Nuclear Power Plant Reactor Vessel Irradiation Surveil-l lance Program. Each of these activities is discussed below. . A. Shipment, Opening, and Inspection of Capsule SwRI loaded the impact and tensile capsules into a radioactive mate-rial shipping cask and transported them to San Antonio, TX. After' visually _ } inspecting and photographing the surveillance capsule basket, SwRI per- ~ sonnel-opened the basket and photographed the contents. The~re was no evid,ence of damage 'to the capsules, see Appendix A. The capsules were opened and the contents identified and stored in accordance with SwRI Procedure XIII-MS-103-1. The end plugs were cut from each capsule with a band saw set up in the hot cell, then the test s specimens and dosimeter wires were removed from the shell and placed in indexed receptacles. Each mechanical test specimen was inspected for identification num-ber, which was checked against the master list in NED0-10ll5 [10], and no discrepancies were found. The neutron dosimeter wires we"e identified 4 and placed in tagged containers. B. Heutron Dosimetry The gama activities of the dosimeters were determined in accordance with SwRI Procedure-XI-MS-101-1 using an IT-5400 multichannel analyzer and l a Ge(Li) coaxial detector system. The calibration of the equipment was ac-137 s radioactivity standards obtained from 54 n, 60 o, and C M C complished with the.U.S. Department of Commerce National Bureau of Standards. All activ-ities were corrected to' the time-of-removal (TOR) at reactor shutdown. The dosimeter wires were weighed on a balance having a sensitivity of 1 microgram. Infinitely dilute saturated activities (A AT) were S _ ~

H 6 I calculated for each of the dosimeters because ASAT is directly related to the product.of the energy-dependent microscopic activation cross section I and the neutron flux density. The. relationship between ATOR and ASAT is given by: TOR, **7 " p ),,- AT -Atm m e SAT m=1 fraction of full power (2511 nth) for the where: .P = m operating period m; decay constant for the activation product, A = day-1; number of operating days for the period m; Tm = tm decay time after the operating period m, = days. The Quad Cities 1 operating history up to the 1982 refueling outage, which was used in the calculation of-ATOR, is presentied in Table I. The primary result desired from the dosimeter analysis is the total fast neutron fluence (> 1 MeV) which the surveillance specimens received. The average flux density at full power is given by: ^ SAT 4=Na g -dependent neutron flux density, where: - c energy /sec; = n/cm ASAT_ = saturated activity, dps/mg. target element; spectrum-averaged activation cross section, .= cm2; and . NO number of target atoms per eg. = The total neutron fluence is.then equal to the product of the average neu-tron flux density _and the equivalent reactor operating time at full power. A discrete ordinates Sn transport analysis for the Quad Cities 1 reactor vessel was performed using the DOT-IV code to determine the radial, a w w y

e 7 TABLE I OPERATIONS SU.9ARY - QUAD CITIES 1 -No. of. Effective Fraction Decay l Operating Days' Full Power Full Power Days Period (Tm) Days (Pm) (tm) .04/72 30 2.62 0.0870-3781 05/72 31 8.16 0.2630 3750 I 06/72' 30 5.95 0.1980 3720 07/72-31 10.08 0.3250 3689 08/72 31 14.74 0.4750 3658 09/72 30-17.65 0.5880 3628 10/72 31 .21.12 0.6810 3597 11/72 30 22.11 0.7370 3567 '12/72 31 23.66 0.7630 3536 01/73 31 23.93 0.7720 3505 02/73 28 20;34 0.7260 3477 -03/73 31 23.40 0.7550 3446 04/73 30 15.36 0.5120 3416 05/73 31 14.10 0.4550 3385 06/73 30 25.81 0.8600 3355 07/73 31 26.30 0.8480 3324 i 08/73 31 21.23 0.6850 3293 09/73 30 18.67 0.6220 3263 10/73 31 21.02 0.6780 3232 11/73 30 20.86 0.6950 3202 12/73 31 25.88 0.8350-3171 01/74 31 18.73 0.6040 3140 02/74 28 23.58 0.8420 3112 03/74 31 24.54 0.7920 3081 0.0000 3051 04/74 30 0.0000 3020 05/74 31 0.0000 2990 06/74 30 07/74 31 3.54 0.1140 '2959 i 08/74 31 25.45 0.8210 2928 09/74 -30 26.31 0.8770 .2898 '10/74 31 15.24 0.4920 2867 l-11/74 30 23.22 0.7740 2837 12/74 31 26.17 'O.8440 2806-01/75-31 7.70 0.2480 2775 02/75 28 3.11 0.1110 2747 03/75 31 27.74 0.8950 2716 .04/75 30 28.00 0.9340-2686 l 05/75 31 19.11 0.6160 2655 06/75 30 24.32 0.8110 2625. [ 07/75 31 23.17 0.7470-2594 l 08/75 31 20.68 0.6670 2563 l 09/75-30. 23.23-0.7740 2533-1 1 4 .-w +

8 TABLE I CONT. -OPERATIONS

SUMMARY

- QUAD CITIES 1 No. of Effective Fraction Decay Operating'_

Days Full Power Full Power Days Period (Tm) Days (Pm) _(tm)_ 10/75 31 22.51 0.7260 2502 11/75 30 19.97 0.6660 2472 12/75 31 19.11 0.6160 2441 01/76 31 1.13 0.0360 2410 0.0000 2381 02/76 29 03/76 31 6.86 0.2580 2350 04/76 30 20.92 0.6970 2320 05/76 31 21.86 0.7050 2289 06/76 30 14.93 0.4980 2259 07/76 31 8.98 0.2900 2228 08/76 31 16.60 0.5350 2197 09/76 30 26.06 0.8690 2167 10/76 31 25.24 0.8140 2136 11/76 30 19.82 0.6610 2106 12/76 31 27.10 0.8740 2075 01/77 31 24.96 0.8050 2044 02/77 28 18.59 0.6640 2016 03/77 31' 15.45 0.4980 1985 0.0000 1955 04/77 30 05/77 31 12.26 0.3950 1924 06/77 30 21.01 0.7000 1894 07/77 31 22.24 0.7170 1863 08/77 31 11.60 0.3740 1832 09/77 30 9.05 0.3020 1802 10/77 31 17.55 0.5660 1771 11/77 30 23.72 0.7900 1741 12/77 31 25.00 0.8070 1710 01/78 31 28.99 0.9350 1679 02/78 28 20.26 0.7240 1651 03/78 31 28.41 0.9160 1620 04/78 30 17.57 0.5860 1590 05/78 31 19.74 0.6370 1559 06/78 30 21.56 0.7190 1529 07/78 31 25.60 0.8260 1498 08/78 31 27.26 0.8790 1467 09/78 30 23.82 0.7940 1437 10/78 31 21.89 0.7060 1406 11/78 30 19.35 0.6450 1376 12/78 31 19.84 0.6400 1345 01/79 31 8.49 0.2740 1314 l 02/79 28 0.11 0.0040 1286 03/79 31 24.43 0.7880 1255 r

9 . TABLE I CONT. i OPERATIONS

SUMMARY

- QUAD CITIES 1

l No. of Effective Fraction Decay: Operating '

Days Full Power Full Power Days Period (Tm) Days-(Pm) (tm) 04/79 30 26.85 0.8950 1225 05/79 31 21.1 3 0.6810 1194 n 06/79 30 28.33 0.9440 1164 07/79 31 - 29.13 0.9400 1133 08/79-31 28.40 0.9160 1102 09/79 30 - 17.02 0.5670 1072 10/79 31 29.55 0.9530 1041 11/79 30 26.31 0.8770 1011 12/79 31~ 21.36 0.6890 980 01/80 31: 28.99 0.9350 949 02/80 29 25.43 0.8770 920 03/80 31 28.37 0.9150 889 04/80 30 27.43 0.9140 859 05/80 31 24.17 0.7800 828 '06/80 30 19.79 0.6600 .798 07/80 31 17.02 0.5490 767 08/80 31 17.88 0.5770 736 i 0.0000 706-09/80 30 0.0000 675 l 10/80 31 0.0000-645 1 11/80 30 12/80 31 5.03 0.1620 614-01/81 31 24.95 0.8050 583 02/81 28 26.20 0.9360 555 03/81 31 21.56 0.6950 524 i 04/81 30 29.30 0.9770 494 } 05/81 31 26.93 0.8690 463 06/81 30 28.05 0.9350 433 f 07/81 31 21.04 0.6790 402 08/81-31 26.20 0.8450 371 09/81 6 27.31 0.9100 341 10/81 31 29.35 0.9470 310 11/81 30 22.19 0.7400 280 -12/81 31 29.23 0.9430 249 01/82 31 28.93 0.9330-218 02/82 2:8 26.17 0.9350-190 03/82-31 27.98 0.9030 159 04/82 30 23.66 0.7890 129 05/82-31 19.63 0.6330 98 06/82 30 19.26 0.6420 68 07/82 31 18.50 0.5970 37 lo 08/82 31 16.04 0.5180 6 l. 09/82 6-2.36 0.3930 O Total irradiation time = 6.64 EFPY

10 vertical, and azimuthal dependence of the fast neutron (E > 1.0 MeV) flux l density and energy spectrum incident on the reactor vessel and surveil-l lance basket capsules. The R-e and R-Z calculations were made using 15 of the 22 neutron group DLC-23/ CASK cross section library, a P1 expansion of the scattering matrix, and an S8 order of angular quadrature. The R-0 model utilized a one-eighth segment, and the R-Z model was represented as a one-half segment (see Figure 1) because of the symmetry involved. The l .various material zones (e.g., core, coolant, internals, pressure vessel, I etc.) were described by homogenizing the major elements within each zone. The resulting spectra were used to calculate the spectrum-averaged reaction cross sections for the threshold detectors and the lead factors needed to relate the neutron exposure of the pressure vessel to that of the surveillance basket capsules. The pertinent factors obtained from this transport analysis are summarized in Table II. The capsule flux is less than the peak fast flux incident on the vessel I.D. because of azimuthal and vertical flux variations related to the core geometry and power distribution. The calculated azimuthal and vertical flux maps are shown in Figure 2. The calculated cross section for the 54 e(n,p)54 n reaction is some-F M iihat less than[the 236 millibarn'(mbi valuelalculated for vessel wal{_ ~ ~~ ~ ~ Basket No. 3 [11]. It is believed that the values given in Table II are reasonable because SwRI-computed reaction cross sections for Browns Ferry Unit 3, which used the same methodology, agree well with those measured by General Electric during the same core cycle [14]. The dosimetry results obtained with the calculated spectra are pre-sented in Table III. If a fission-spectrum energy distribution is assumed 54Fe(n.p)S4Mn reaction I at the capsule location, the cross section for the (E > 1.0 MeV) would be 98.26 mb [4], and the resulting value for fast flux n/cm/sec._Thisvalue_isre-2 at the capsule location would be 4.2 x 108 ported for reference only and has not been used in the analysis of results. 1 ( The discrepancies in the peak vessel flux values determined from the several dosimeter materials are attributed primarily to the uncertainties j in the calculated spectra, in the spatial distribution, and in the reaction-f i

0 II 45 Shroud Capsule Jet '- ,um t P.V. -_-,____l__ H0 2 HO i 2 i l l ..__u._m_._._..._i._

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__..,u____._ ._ u _ _._i_._._p_ i l l )/9, _.. 1_.. _ l_ _ __ f _. _ _ l;.... _ _ _ _ _ l_.... _ '.g[ W o o (a) R-0 Model g l [ f Shroud l I i l l }N l i i l l l l i e i i i i e i e e i i i i 'Jete P.V. i apsde l l Core' i i ( i r e i l l Y I l l i l l l i i i f I i e e i e e i i .i ,i i i i i e i e e i j 1 ' NO e i 2 ,i .i l .I e e i. e i e _q g i l s* (b) R-7 Model FIGURE 1. DISCRETE ORDIflATES CALCULATI0flAL f10DELS, QUAD CITIES UNIT 1

12 r l: l-TABLE II RESULTS OF DISCRETE ORDINATES Sn TRANSPORT ANALYSIS QUAD CITIES 1 AND 2 t .14. Calculated Reaction Cross Sections for Analysis of Fast heutron i ' Monitors (E > 1.0 MeV) l Reaction 5 (barns) 54 e(n.p)54pm 0,197 F p [. 58Ni(n.p)58o 0.237~ C 63Cu(n,a)60 o 0.00330 C l s B. Calculated Basket Lead Factors (E > 1 MeV) l- } Location within Position (a)- Vessel Wall Lead Factor (b) l i 1: j 215' I.D. Surface 0.755 215' 1/4T 1.07 j. l 21 5' 3/4T -4.75 i l-l . (a) Azimuthal position of surveillance Basket No. 8~ Capsule neutron flux density E > 1.0'MeV (b) Maximum neutron flux density at vessel I.D., E.> 1.0 MeV t 4 ? i t' f h I. ~,

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!!!i i h e 1- !j! Jtt +a+w.a,7-t i _e im 1 o "m dit p o i i i '1 !!p m a-.. -250 1x107 2 4 6 1 x 108 2 4 x 108 Neutron Flux, E > 1 MeV (b) R-Z Calculation FIGURE 2. CALCULATED FAST FLUX (E > 1 MeV) DISTRIBUTION, QUAD CITIES UNIT 1

~- TABLE III SUmtARY OF NEUTRON DOSIMETRY RESULTS QUAD CITIES 1, VESSEL WALL SURVEILLANCE-BASKET NO. 8 (215*) Dosimeter I.0. Activation Weight ATOR ASAT Flux, E > 1 MeV(a) Material Capsule Reaction (mg) (dps/mq) (dps/mg) (n/cm2 sec) / 54 e(n.p)S4 n 112.8 1.813 x 101 2.518 x 101 2.08 x 108 F M Fe G1 -G2 112.0 1.859 x 101 2.582 x 101 '2.13 x 108 Average = 2.10 x 108 63 u(n,a)60C0 297.3 3.309 x 100 6.700 x 100 3.10 x 108 C Cu G1 o 290.9 3.351 x 100 6.785 x 100 3.14 x 108 'G2 Average = 3.12 x 108 Ni G1 58Ni(n.p)58 o 210.2 2.046 x 102 3.141 x 102 1.89 x 108 C r G2 o 21 3.0 2.091 x 102 3.210 x 102 1.93 x 108 Average = 1.91 x 108 (a) Calculated flux values subject toa i 16.5% uncertainty (lo).

15 cross sections. Other neutronic factors contributing to the estimated 16.5% uncertainty (le) in.the calcu.lded fJuxeJ are thi determinatioi: ~ of disintegration rates and the calculation of reaction rates (ASAT/NO). For example, the iron monitors effectively measured the capsule flux for the last three years of operation, and the copper monitors effectively measured the capsule flux for the entire exposure period. Averaging the results obtained from the iron. aid coppe'r tHfesholC f neutron dosimeters, the fast neutron flux at the surveillance basket ~ 8 n/cm2/sec, _I _.$ I ' location during full power operation would be 251_x J0 E > 1 MeV, and the peak value incident on the pressure vessel I.D. would [ be 3.40' x 10. Since Quad Cities 1 operated for 6.64 effective full 8 power years (EFPY) up to the September 1982 refueling outage, the cal-culated basket and vessel fluences to that time are as follows: 5.47 x 1016 n/cm2 i - Surveillance Basket 7.24 x 1016 n/cm2 Pressure Vessel I.D. Surface 5.11 x 1016 n/cm2 Pressure Vessel 1/4T 1.15 x 1016 n/cm2 Pressure Vessel 3/4T f i The vessei wall fluence as a function of plant operation is shown in l Figure 3. C. tiechanical Procerty Tests Hardness tests were run in accordance with ASTM Method E 18 [15] on one Charpy V-notch specimen selected from each material group. The re-sults are given in Table IV. The irradiated Charpy V-notch specimens were tested on a calibrated

t 240-ft-lb,16-ft/sec SATEC impact machine (Model SI-lK) in accordance with l

SwRI Procedure XI-MS-104-1. The test temperatures, selected to develop ( the ductile-brittle transition and upper shelf regions, were obtained l using a liquid conditioning bath monitored with a Fluke Model 2168A

Inspected and. calibrated using specimens and procedures obtained from l

the Army Materials and Mechanics Research Center. r

4 x 1017

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+ ~ >== 16 =u ~. .ji ~ p- '_ y i I

I '

~ i.. .I 11' .I l w' + / ;: ' ;l; q u.." t ntat c m 1 ~ g . I ( i I f I 'I II+: il i1I.. l !T I 1 I I .ti i . e i i ] I I' I i r 'I e f I e ..r, .f' /i I i t I.1 i i i i if . :lli P i f i i. + ii a. i .ili e i i I* / / Il i!! /, e u ti i on m, ! i ; H O / i / i i I i i lid po I i 6 ljH M I l I!h M / i /l l I l l l i fl i 1 x 1017 i i i es i c y v ..m... g. w.. me .m ... =H...w s. &,.. -= a emm==m..== . -. _-.A i _=. m . y g. g .t .,g J,.e,.e-=a[w ._...~ g==e e e-o /[ ye 4 .e4+ & e+==n Q d% 4= h V -'i = 6 -"v 4 RP

-e -

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I

1 !i e I I Iil !!I' I I! ! 't!' l I O 'm H P l i i I i I , IF I !! ! 9! l Ilj. 1 x 1015 i i i i i u i, - i usi a i un 1 2 4 6 10 20 40 Plant Operation, EFPY FIGURE 3. VESSEL WALL FLUENCE AS A FUNCTION OF OPERATION OF QUAD CITIES 1

17 J TABLE IV HARDNESS PROPERTIES OF SURVEILLANCE MATERIALS QUAD CITIES 1 1 (Fluence.= 5.0 x 1016, E > 1 MeV) Charpy . Test Spec 1mgn Material' No. (,a 1 Hardness (HRB) Base M16 94.0 94.0 { -94.5 95.0 94.5 Average = 94.4 i ESW M7L 88.0 90.0 90.0 90.5 90.5 Average = 89.8 ESHAZ MPD 89.5 89.5 90.0 91.0 90.5 Average = 90.1 i (a) Specimen identification code given in Reference 10. i i s .y.

~ 18 i r. i ' digital thermometer. -The Charpy V-notch impact data obtained on the ir-I tradiated ~ materials are presented in Table V, and photographs of the frac- {' ture faces are included in Appendix B. The Charpy.V-notch transition curves for the irradiated plate material, weld metal, and HAZ material contained in the 215' capsules are compared to the unirradiated values i [ - in ' Figures 4 through 6. jihe, shif ts,1(the,50{ft;1b,[30 ft-1b,Jnd]S 'mlT lateral expansion transition temperatures were small, see Table VI. The l "Charpy upper shelf energy behavior, also shown in Table VI, was not en-tirely consistent with the transition temperature changes. For example, the decrease in the transition temperature of the weld metal may result i from data scatter rather than a real change in the material property. However, the small changes which were measured are consistent with the I-level of fluence received by capsule basket No. 8. i Tensile tests were carried out on the irradiated materials in ac- . cordance with SwRI Procedure XI-MS-103-1 using a 22-kip capacity.MTS I ~ Model~ 810 tester equipped with an Instron Catalog No. G-51-13A 2-in. l strain gage extensometer and Hewlett Packard Model 7004B X-Y auto-graphic recording equipment. Tensile tests were run at room tempera-l ture and 550*F. The results are presented in Table VII. Each tensile load-strain record and photographs of the tested specimens are included j-in Appendix B. i- 'D. Check Chemical Analyses i 4 Check chemical analyses were run on' samples cut from the fracture end of selected tested Charpy specimens. All of the weld metal speci-f mens were tested for copper and nickel content at SwRI using an X-ray fluorescent technique. One base plate and two weld specimens were l { then sent to. Westinghouse Advanced Reactors Division Analytical Labo- - ratory for complet'e analyses using gravimetric.(Si), combustion (C and S), and ICP P1asma:(remainder of elements). methods of analyses. The re-sults are summarized in Table VIII. The differences in the copper and nickel results can be' attributed to at least three factors: 2

19 (1) The ICP Plasma method measures the chemical content of the full volume of sample material, while the .X-ray fluorescent method looks only at the surface i-of the sample. (2) The I square centimeter cross sectional area avail-able from the Charpy samples is smaller than desired for detecting small amounts of residuals with an X-ray fluorescent technique. . (3) The gamma activity of the sample increases the diffi-cul_ty in accurately measuring the fluorescent peak. 4 1 4 i 1 1

20 TABLE V CHARPY V-NOTCH IMPACT DATA ON SURVEILLANCE SPECIMENS REMOVED FROM QUAD CITIES UNIT 1 (Fluence = 5',5 x 1016, E > 1 MeV) Test Impact Lateral Spec mp Temperature Energy Expansion Shear Material No. aj (*F) (ft-lb) (mil) (%) Base ~Ml B -80 8.0 6 0 HlD -40 -16.5 16 5 HlJ -20 42.5 35 10 Mll 0 44.0 34 15 M16 72 92.0 71 50 ~M13 120 104.0 79 100 MIK 160 109.5 81 100 M12 200 106.0 84 100 l-I ESHAZ MPA -80 5.5 6 0 ML6 -40 20.5 18 5 MJB. -30 36.0 31 5 MJU -20 41.5 36 10 MKB 0 40.0 37 10 -MPD 72' 117.0 55 100 MJ1 120 107.0 88 100 i MLU 160 112.5 89 100 ESW MBU -80 2.0 2 0 i MAD -40 . 2. 5 5 0 MA4 -20 .36.0 31 5 MBY 0 55.0 43 10 M7L 72 79.0 65 50 M6D 120 83.0 ' 63 65 MDC 160 93.0 65 90 i MD4 200 105.0 75 100 (a) Specimen identification code given in Reference 10.

21 lu. C U ~^ 144. E " tae. E 8 m e R o see. y 30.0 t F T u.0 O s e# --C>-- eJAD CITIES 1 SASE RETAL UNIRRADIATO - L 4e.g B /o A* 55 44 8* 49 13 71

8.3.

72 * $1.68 g / ~ --G-- 0JAO CITIES 1 BASE METAL TRA*DIATED I 84 0 /e' ' ~' A* 52 15 Be 52.39 ft

4.49 12 * $1.57 s

0,7 I I I f .00

1M.

.40 SM. 3 0. 300. 400. TUPDATleE (DEC F3 see. i e a s C u e , yj y S j.- - ( e .O s. O A f Y f / n p O/ p,....... 4,g e4 /: n / +C--@ RAT.Ew.tATA - ect Bast RTAl. - LM1ARASIATU I C ., /.. 36.47 3* 25.30 ft * -0.73 TE

es.4
L y,,

s uT.w..ATA - SC1.Aet Em - le. Asia,o. 8 A* 48.5. B* 37.TT T1

8.00 f3
es.95 I

I f f .e0 -100. .00 See. See. 300. 400. i TWWERMtsE (354 F3 l FIGURE 4. CHARPY V-NOTCH IMPACT PROPERTIES OF QUAD CITIES UNIT 1 SURVEILLANCE BASE PLATE I l

22 164. a a a a C Y gg, g M 33, E

O C see.

v s '. . e. .p g g l / / l r

7

~' T $e.4 / /

/

O L . /: + 0unD CITIES 1 ELECTROSLAG WLD FETAL tm!AAADIATED q,, / A= S4.54 B* S4.55 f t

49.24 72
52.85

/ g i / 8 8 *e ~' ' --e--GuaD CITIts t ELECTROSt.AG WLD IETAL IMADIATED ~ 45.12 TI

S4.96 C

A * -42.85 3,* ft * ~6.,44 w d* Ce t e .se

lee.

.00 100. 300. 300. 400. TDFEMTUIE iBEC F3 100. 4 5 4 C y 8

O L

e 30.4 / 4.Q 9....*........ ...y p ,/ M P p / ........p.a. 4,, I n e l l l f t --C--C-V 1AT.DF.94TA - OC1 ESW IETAL - LSeleM01ATES I / ~ L g, A* 27.88 3* 30.40 ft

35.12 TE
47.81 s'

~8--C-4P LAT.DF. DATA - GC3 ESW #ETAL - IneA08ATES. p# I O 8 A* 44.98 3* 30.Se ?! = 19.98 78

N.50 4P 9

e i e e .00 -tee. .00 100. 300. 300. 400. TUFERAftM (8ES F) l r l FIGURE 5. CHARPY V-NOTCH IMPACT PROPERTIES OF QUAD CITIES UNIT 1 SURVEILLANCE ELECTROSLAG WELD METAL

23 8 4. C u g M ~' f* $2e. ,,,w_,,,,,_,,,,,,,,,,,,_ ~ e g /. g .j / Y / /

7 Se.4

( ,j F / O T Ee.4 .Oi f s = / O# + oump CITits 1 H42 miratAL UNtaneetetta _ L ,.g,. i <e.e A= $3.57 5= 57.04 ft. 19.66 72

St.34.

8 5 3 88.4 -+ - Que0 CITIES I se42 mitRIAL tmRASIATED " g A* G0.42 3= 55.74 ft = 9.41 72

57.44 *

,e f f f f .00 -100. .00 800. 300. 300. 400. TDFERATUIC (DEO F3 e s a tee. e. C /, v .O.....,..... L A S 7 ,,, - - - - - - -. - - - - = = - - - - - - e' ': e O. s s. / f ~ ~ O' / l p

/

(. ....................j. /. R /l / O + C-v 1AT. DIP. gate - ett Em mz mf. - intaguetAtp g / _....../............ A. ,0..e ,1. -l.. 0.. l L,,,, l dI e' --S-- CW TAT.DP. DATA - SC1 EW m2 mf. a. IMIATEB - 3 ~, I l A* 35.3F,3

33. 5 ft * -1.98 ft o SS.m e

j .Se -les. .Se leg. See. 3es. 400. TUFERumsE (384 Ft I FIGURE 6. CHARPY V-NOTCH IMPACT PROPERTIES OF QUAD CITIES UNIT 1 SURVEILLANCE ELECTROSLAG WELD HEAT-AFFECTED ZONE MATERIAL

24-TABLE VI EFFECT OF IRRADIATION ON THE CHARPY V-NOTCH UPPER SHELF ENERGIES OF QUAD CITIES UNIT NO.1 VESSELSURVEILLANCEMATERIA(SBASKETNO.8(215*) (Fluence = 5.5 x 10 o, E > 1 MeV) i ESW ESW Base Weld HAZ 4 Criterion (l) Plate Metal Material TransitionTemperatureShift(2) 0 50 ft-lb 1F -40*F -6*F 0 30 ft-lb 4*F -28'F -4 F 0 35 mil 7*F -32*F 5*F ARTNDT(3) 4ep (4) (4) C Upper Shelf Drop nil 3 ft-lb 6 ft-lb y (3%) (5%) (1) Refer to Figures 4-6 (2) Cy parameter determined by hyperbolic tangent fit where C parameter = A + B tanh((T - T )/T ) 1 2 y (3)Transitiontemperatbreshift@30ft-lb[1] (4) Apparent negative shift in RTNDT .r L l

TABLE VII TENSILE PROPERTIES OF SURVEILLANCE MATERIALS RASKET NO. 8 (215*),QUADCITIES1 (Fluence =.5.5 x 1016, E > l MeV) Fracture Fracture Uniform Total -Test Spec Temp' O.2% YS UTS Load Stress Elongation Elongation R.A. Material No.ta) ( F) (ksi) (ksi) (lb) (ksi) (%) (%) (%) 8ase .PDE 72 71.3 93.5 2860 194 14.8 23.7 69.9 PDC 550 58.3 82.7 2650 160 12.6 21.5 66.4 ESHAZ-PL5 72 65.2 84.3 2550 178 11.7 19.8 70.4 PLJ 550 59.8 76.8 2530 160 10.2 17.1 67.2 ESW PJS 72 65.6 85.0 2670 178 14.1 24.9 69.5 PJ7 550 58.0 78.7 2780 153 12.0 18.1 62.4 (a) Specimen identification code given in Reference 10. I y

_m.___._____ . _ _ _ _.. _. _ _ _ _.. _ _ _ _ _ _. _. _. _ _ _. ~.. _... TA8LE VIII . r CHECK CHEMICAL ANALYSIS RESULTS BASKET NO. 8. Qun0 CITIES-1 Specimen. Weight Percent of Element Identification (a) Source (b) C S P Si Cu Ni Cr Mo V l ' M16 W .210 .010 010 .16 182~ 48 .092 45 <.001 S. .25 48 3 j MR4 W .158 .006 012 .05. 136 25 .063 45 <.001 1 i S 19-26 MAD W. .236 .004 012 . 08 _.138 26 .063 45 <.001 -38 5 19 l 20 .30 2 IRL .S l 24 .33 peu 5 16 .39 MED S 21 .40 l IOC -5 19 ..33 4 } 994 5 21 .40 fey S 27 .56 990 S 1 i t (a) Cut from tested C, specimen, see Table V for material I.D. (b) W = Westinghouse Atomic Power Division; S = Southwest Research Institute. l I i d i i N I [ 4 i 4 t 3, 1 .r. i r-y .,,w -<<r

-vy i--- -&

-ew.,,r,, 4-

27 V. HEATUP AtlD C00LDOWil LIMIT CURVES FOR OPERATION OF QUAD CITIES UNIT fl0. 1 Unit No.1 is one of four similar 251-in. I.D. boiling water reactors (Dresden Units 2 & 3 and Quad Cities Units 1 & 2) operated by Comonwealth Edison Company. Each of the similar units has been provided with a reactor vessel material surveillance program as re-quired by 10CFR 50, Appendix H [1]. Based on the capsule analyses, heatup and cooldown limit curves for Level A and B service and for up to 10 EFPY of operation have been computed. The limit curves developed are a generic set for all four vessels listed above. The curves, given in Figure 7, are based on the worst case for vessel fluence rate, copper / nickel con-They were computed in accordance with tent, and initial RTNDT. Regulatory Guide 1.99, Revision 1 [8] and NUREG-75/087 [15] using the following bases: 1/4T RT = 60'F NDT 3/4T RT = 49 F NDT Cooling / heating rate = 100'F/hr. 4 Vessel inner radius = 125.5 in. Vessel outer radius = 131.8 in. Operating pressure = 1030 psig Initial temperature = 60'F Final temperature = 550'F Effective coolant flow rate = 1.08 x 108 lbm/hr. Effective flow area = 96.0 ft2 Effective hydraulic diameter = 24.0 in. I I

I l ! l I ! ll.!!! !!!! !! 4.I T !![ !_! i.l!. !}[j j.! !II !ll '.[ l! Ul ~}I I ,!!! !!U.'iI!!I L ! !!!! :ii! !!! ! i' !! Il i 11! !!!! i fi / l 1 ! II / I l t l l f i lii !!i! 3 ENh $!N!$ $!!! ! !.l! i[. N !N ! l . I ! I}I. L l I! I !!! !ll! I !i i !! ! !? ! l il r i ll! i l l I i ,{ ( [l. i i I ( I !id{ Ul ! .. Ll.!L { ! illdi lli OI! I L ! !? = i j; {.{' i a j 't E ii! x c . ii t- - t, C +- ..a a g a g I!! !.i! !!:! l ! ij j-{l. i i I[ ~ /)! a TV i I Ul TIlii lli! ill i9 I. !I!! ! j j j j j 800 y j;; ; }j}j j; '; i f l l; I 'I [ l j ^ l E

!i! !Ill i l i!!! liil i

i i 1 ~

p L

{ r 3 f .p.. .J .. -! b :; l I l 3 pi!L ! Ll : j l $ 600 7 f'T 'j j y . T UB'i i l 1 j t .'d*' lill ' 1. i I j i i ! Ti'i 3 l i j i jj ' ig l[ j' [{l !!hI l lj [. [ Curve A - Inservice Pressure Tests Limits g} i Curve B - Heatup - Cooldown Limits For j ii j j j-j a 400 t.~*j d j .. ! [. t!!.plj Nonnal and Upset Operation l j I h 1J; [li L,J - r Tj e Curve C - Critical Core Operation Limits bl U I.l l i

0 1

11 ....j.i .j j ! { Bases - }!..![ L g dI Beltline RTNDT = 60F01/4t and 49F03/4t [. f ! i = l 60F 200 j j Closure Flange Region RT nT =hr Max. gj gj j; j [l N 1 [~ lleating/ Cooling Rate = 100F/ ) j I (( 11 l l l I l l [j i Ii !!illllillll}ll!lfil!lillilll}}}}} !lil}}}{llillllIlli,illlllllIllllllllli !lll}!IlllililllllIllill} ! l ll tl li il I L ' !j i ti l! ! li O 60 100 150 200 250 300 350 400 Vessel Wall Temperature (deg F) FIGURE 7. PRESSURE-TEWERATURE LIMITS FOR LEVELS A & B SERVICE FOR UP T010 EFPY l Is

2_9, V, I. 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," 1980 Edition. 3. ASTil E 208-69, " Standard Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," 1975 Annual Book of ASTM Standards. 4. L. E. Steel and C. Z. Serpan, Jr., " Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970. 5. L. E. Steele, " Neutron Irradiation Embrittlement of Reactor Pres-sure Vessel Steels," International Atomic Energy Agency, Techni-cal Reports Series No. 163, 1975. 6. ASME Boiler and Pressure Vessel Code, Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," 1980 Edition. 7. 'J. C PeFrind,3,_tiullaert2G',_R._0dittCaiidP.M.LombrozoT ~ ~ ~ " Physically _ Based Reoression Correlations of EmbrittisseHt"Fata~ i WorOie~ actor Pressure Vessel SurveilliicD~FograinsN EPRT NP 3317, ~~ ~ '~ ~ -~ 7fhiiaW,1,98(,. 8. Regulatory Guide 1.99, Revision 1, Office of Standards Develop-ment, U.S. Nuclear Regulatory Consnission, April 1977. 9. ASTM E 185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1975 Annual Book of ASTM Standards. 10. J. P. Higgins and F. A. Brandt, " Mechanical Property Surveillance of General Electric BWR Vessels," NEDO-10115, July 1969. 11. J. S. Perrin et al, " Quad Cities Nuclear Plant Unit No.1 Reactor Pressure Vessel Surveillance Program: Capsule Basket No. 2 and Capsule Basket No. 3 " Battelle Columbus Laboratories Report, March 1, 1975. 12. T. R. Mager et al, " Analysis of the Third Capsule from the Common-l wealth Edison Company Quad Cities Unit 1 Nuclear Plant Reactor Ves-sel Radiation Surveillance Program," Westinghouse Electric Corpora-l tion WCAP-9920. Electric Power Research Institute Research Project l< 1021-3 Topical Report. August 1981. l l

~_ .~ 30 l J. S. Perrin and-L. M. Lowry, " Quad Cities Nuclear Plant Unit No.1 13.- and Unit No. 2 Reactor Pressure Vessel Surveillance Programs: Unir-radiated Mechanical Properties,"~Battelle Columbus Laboratories Final ' Report to Commonwealth Edison Comapny,' February 15, 1975. 14. .E. B. Norris, " Spectral Analysis of a BWR Vessel," Proceedings of the Fourth ASTM-EURATOM Symposium on Reactor Dosimetry," NUREG/CP-0029, - Vol. 2, pp. 1043-1050. 15. US NRC Standard Review Plan, NUREG-75/087, Section 5.3.2, Pressure-Temperature Limits,' November 24, 1975. i i 1 t i e i d ) l l l

31 4 b APPENDIX A SURVEILLANCE SPECIMEN ORAWINGS AND CAPSULE PHOTOGRAPHS i l

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34 APPENDIX B PHOTOGRAPHS OF TESTED SPECIMENS AND TENSILE TEST RECORDS I l e

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7 . Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET e Test No. T-f Est. U. T. S. psi Project No. O S '78 C'7-M Spec. No. PDE Initial G. L. I, ozz. in. Machine No. 22v.pMTs Temperature ~12 'F-Initial Dia. .2sr in. Date s-r r -g<t Strain Rate.ocS'a/;a /m.: Initial Thickness in. Initial Area,O'J906 ta E L Initial Width in. Top Temperature 'F Maximum Load 4590 lb I Bottom Temperature 'F 0.2% Offset Load 3N lb Final Gage Length l,2Gf-in. 0.02% Offset Load Ib Final Diameter O. 13 7 in. Upper Yield Point Ib Final Area

0. 0 i O Y in. 2 e

/ U.T.S. *= hk 607 psi = Initial Area ~/l, ~b01 psi 04 2% Y. S. = = Initial Area j l

0. 029. Mset had 2,g

0. 02% Y. S.

= Initial Area l = pper' Yield Point U psi Upper Y.S. = 1,,, t G. L. a G % Elongation = x 100 = 2'5. ~/ ef, 3 I

^#**~

A"** = x 100 = $9.9 % R.A. Initial Area I g Sign:ture: (. y I t I l e O.

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n

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u m =-:!!:.p:m

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t,-

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rt: .-

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t.

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3. E..

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re.

t.... nn

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... ut:

m m:

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:.I.~a::

s.

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r =

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gj+n" -M

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m-m : uu m

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.:: m..a.n-in1 n.a-... :..a...u.};t.a.ncu.t

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r:

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mu.

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n n,c

ne n m

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m

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n.

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: =-

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Southwest Research Institute Department of Materials Sciences TENSILE TEST DATA SHEET ? Test No. T-2 Est. U. T. S. psi Project No. OG-78S7-ool k7E Initial G. L. /, O'20 in. Machine No. 2Z k'E AM Spec. No. 6-~ l b M Tamperature I 'F Initial Dia. .UO in. Date OD6If a!wk Initial Thickneas in. Initial Area M 9 Db8 I Strain Rate l Initial Width in. Top Temperature 'F Maximum Load N I70 lb Bottom Temperature

F 0.2% Offset Load 32EO lb in.

0.02% Offset Load Ib Final Gage Length Final Diameter

I in.

Upper Yield Pcint Ib Final Area .OI4W in.2 g?g# f H um L ad psi U.T.S. = Initial Area / = h 6 M psi

0. 2% Y. S.

= 3 Initial Area

0. 027. Maet had p,i

0. 02% Y. S.

= Initial Area l l Upoer' Yield Point. Ps2 j Upper Y.S. = 1 Area x 100 = 2 *

*~

~ ' % Elongation = Initial G. L. ""*A"**~ = x 100 = % R. A. Initial Area y Sign;ture: P i

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m. e

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m...dua u.-: 2uu ant:2.:ca= mipun F

+ t;;;.

s aum;m

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t

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n tm:mun_

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t Ma t

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m: am r

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t: ut:

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n :

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"** T

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j{CO

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n :nu

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=

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h:tur:.n..== - -=c s.r..n=:_. _:

=:

m:

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:= :mjuu= uma

nag-=

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m un :.

an* =n i := :- i a. t: : u: :nd:

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n_t

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=

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5:uthwact Reciarch Instituta Department of Materials Sciences TENSILE TEST DATA SHEET Tset No. T-3 Est. U. T. S. psi Project No. OC= ~76S7'COI Pbb Inidal G. L. I, Ol9 in. Machine No.82 MIE hfPS Spec. No. Temperature ~/I-

F Initial Dia.
Z46 in.

Date 6-Ib -bY Strain Rate,006M,hm.h Initial Thickness in. Initial Area. 049 30S ^ 2 ~ Initial Width in. Top Temperature 'F Maximum Load 4070 lb TO Bottom Temperature 'F 0.2% Offset Load lb l j Final Gage Length l.2E l in. 0.02% Offset Load Ib !W Final Diameter in. Upper Yield Point Ib Final Area

O I DI 4 in. 2 E

I E %'g gg / um ad U. T. S. = Psi Initial Area / . o set L ad, [pS* Zll Psi

0. 2% Y. S.

= Initial Area

0. 025. Mut had

0. 02% Y. S.

P,i = Initial Area Unner Yield Point I Upper Y.S. = = psi 1 j i / Final G. L. - Initial G. L. % El dm = x 100 = Ib'b Initial G. L. / ^""~ ^#** x 100 = 10* N % R.A. = Initial Area f/ w y Y' k nl Si

t... t F

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F Initial Dia.

2Oh in. Date $~~ Ik ~ bf-Strain Rate.OO6sa/r /m.'o Initial Thickness in. Initial Area. 04 NO6'M Initial Width in. l Top Temperature ____

F Maximum Load 3 7 l O lb I

Bottom Temperature

F 0.2% Offset Load 2690 lb i

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0. 2% Y. S.

= = p Initial Area set ad

0. 02% Y. S.

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Scuthwast Rosacrch In::tituta Department of Materials Sciences TENSILE TEST DATA SHEET Test No. T-I Es t. U. T. S. psi Project No. 06-7967-OO l - Spec. No. - Initial G. L. I,O2 O in. Machine No. 22 kiE MTS 2do in. Date 5-l4-64 / ~ Tamperature 550 F Initial Dia. Strein Rate #00$i*/Nm,A Initial Thickness in. Initial Area, OM O@7,$Z Initial Width in. l Top Temperature

F Max 2 mum Load MOhD lb Bottom Temperature
F 0.2% Offset Load 2 9 G0 lb Final Gage Length I
D9 in.

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= dh

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F Initial Dia.

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