& . — | | ORNL~TM-3063 [ ()‘ S 5 ) Contract No. W-7405-eng-26 METALS AND CERAMICS DIVISION AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT HASTELLOY N SURVEILLANCE SPECIMENS — FOURTH GROUP H. E. McCoy, Jr. el %N < MARCH 1971 _ _LEGAL NOTICE——— i This " report - was prepared a3 an account. of work | sponsored by the United States Government, Neither | 1 the United States nor the United States Atomic Energy -] Commission, nor any of their employees, nor any -of { | their contractors, subcontractors, or their employees, | : .| makes any warranty, express or implied, or assumes any - | legal liability or responsibility for the accuracy, com- | pleteness or usefulness of any information, apparatus, -product or ‘process disclosed, or represents that its use wouid not infriflge priva;flely ow@ed rights. ok | | | OAK RIDGE NATIONAL LABORATORY * _ , , ~ Oak Ridge, Tennessee L o - operated by o I .~ UNION CARBIDE CORPORATION P _ for the ; ‘ j ' | U.S. ATOMIC ENERGY COMMISSION ; n ” | | ~ DISTRIBUTION OF THIS DOCUMENT IS UN | o ifl % 4 %, %, oy (e LN 4, ay 7 C iii - CONTENTS Abstract . . . . . . . . . .. 0 o0 e L Introduection . . . . . . . Experimental Details . Surveillance Assemblies e e e e e e e e Materials . . . . . . . . . Test Specimens . . . . . . . . . . Irradiation Conditions . . . . . . . . . . . Testing Techniques . Experimental Results . . . . . . . . « « « & Visual and Metallographic Examination Mechanical Property Data — Standard Hastelloy N Mechanical Property Data — Modified Hastelloy N Metallographic Examination of Test Samples . Discussion of Results . . . . . . . . . . « . . . Sumary and Conclusions 7 Acknowledgments . . . . . . . Page 0 08 I vt »t WD L) o g OB 2 Wn ~l 1 MM 2 &) i ¢ ;A “) “) . AN EVALUATION OF THE MOLTEN-SALT REACTOR EXPERIMENT HASTELLOY N SURVEILLANCE SPECIMENS — FOURTH GROUP H. E. McCoy, Jr. - ABSTRACT Two heats of standard Hastelloy N were removed from the core of the MSRE after 22,533 hr at 650°C and exposed to a thermal fluence of 1.5 X 1021 neutrons/cm? and a fast fluence > 50 kev of 1.1 X 10%! neutrons/cm®. The mechanical proper- ties have systematically deteriorated with increasing fluence. However, the change in ropertles is due to the helium produced by the 1 B(n,x)”1Li transmutation and can be reduced - by changes in chemical composition. Some of these modified heats have been exposed to the core of the MSRE and show improved resistance to irradiation. The corrosion of the Hastelloy N has been largely due to the selective removal of chromium. The rates of removal are much as predicted from the measured diffusion rate of chromium. Other superficial structure modifications have been observed, but ‘they likely result from carbide precipitation along slip bands that were formed during machining. INTRODUCTION The Molten~Salt Reactor Experiment (MSRE) is a single region reac- tor that is fueled by a molten fluoride salt (65 LiF—29.1 BeFo—5 ZrF,— 0.9 UF;, mole %), moderated by unclad graphite, and contained by - Hastelloy N (Ni~16 Mo-7 Cr—4 Fe-0.05 C, wt %). The details of the reac- tor design and constructidniban_be'fbund elsewhere.® We knew that the neutron environment would produce some changes in the two structural mate- - rials - graphlte and Hastelloy N. Although we were very confldent of ~ the compatiblllty of these materlals with the fluoride salt,'we needed to keep abreast'ofrthe_p9581ble;development of-corros1on.problems within | 1R c. Roblnson, MSRE De31gn and Operations Report ' Pt. 1, Descrlp- tion of Reactor Design, ORNL-TM-728 (1965). the reactor itself. For these reasons, we developed a surveillance ( program that would allow us to follow the property changes of graphite S0 v and Hastelloy N specimens as the reactor operated. O The reactor went critical on June 1, 1965. After many small prob- lems were solved, normal operation began in May 1966. We removed four groups of surveillance samples. The results of tests on the first three groups were reported.2_4 This report deals primarily with the results of tests on samples removed with the fourth group. The fourth group included two heats of standard Hastelloy N used in fabricating the MSRE and three heats with modified chemistry that had better mechanical proper- ties after irradiation and appear attractive for use in future molten- salt reactors. The respective history of each lot was (1) standard Hastelloy N, annealed 2 hr at 900°C and exposed to the MSRE core for 22,533 hr at 650°C to a thermal fluence of 1.5 X 1022 neutrons/cm?, (2) two heats of modified Hastelloy N, annealed 1 hr at 1177°C and exposed to the MSRE core for 7244 hr at 650°C to & thermal fluence of 5.1 x 1022 neutrons/cm?, and (3) a single heat of modified Hastelloy N, annealed 3 for 1 hr at 1177°C and exposed to the MSRE cell enviromment of N, + 2 to e 5% 0, for 17,033 hr at 650°C to a thermal fluence of 2.5 X 101° neutrons/cm?. The results of tests on these materials will be presented in detail, and some comparisons will be made with the data from the groups removed previously. EXPERIMENTAL DETATLS Surveillance Assemblies The core surveillance assembly’ was designed by W. H. Cook and others, and the details have been reported previously. The specimens 2H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — First Group, ORNL-TM-1997 (1967). 3. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — Second Group, ORNL-TM-2359 (1969). . “H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — Third Group, ORNL-TM-2647 (1970). °W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965, o/ ORNL-3872, p. 87. | - S . ¥ <) ooy &) gl v‘g ™\ I are arranged in three stringers. Each stringer is about 62 in. long and consists of two Hastelloy N rods and a graphite section made up of various pieces that are joined by pinning and tongue-and-groove joints. The Hastelloy N rod has periodic-reduced sections 1 1/8 in. long by 1/8 in. in diameter and can be cut into small tensile specimens after it is removed from the reactor. Three stringers are joined together so that they can be separated in a hot cell and reassembled with one or more new stringers for reinsertion into the reactor. The assembled stringers fit into a perforated Hastelloy N basket that is inserted into an axial position about 3.6 in. from the core center line. A control facility is associated with the surveillance program. It utilizes a "fuel salt" containing depleted uranium in a static pot that is heated electrically. The temperature is controlled by the MSRE computer so that the temperature matches that of the reactor. Thus, these specimens are ekposed to conditions similar to those in the reac- tor except for the static salt and the absence of a neutron flux. There is another surveillance facility for-Hastelloy N located outside the core in a vertical position about 4.5 in. from the vessel. These specimens are exposed to the cell enviromment of N> + 2 to 5% O,. Materials The compositions of the two heats of standard Hastelloy N are given in Table 1. These heats were air melted by the Stellite Division of Union Carbide Corporation. Heat 5085 was used for making the cylindrical portion of the reactor VéSSél'and heat 5065 was used for forming the top ~and bottom heads. These materialS'wefe given a mill anneal of 1 hr at 1177°C and a final anneal of 2 hr at 900°C atxORNL after fabrication. The chemical compositions of the three modified alloys are given in Table 1. The modifications in composition were made principally to imprové the alloy's,resiStanéejtdnradiation demage and to bring about -general improvements in the fabricability, weldability, and ductility.® ®H. E. MeCoy and J. R. Weir, Materials Development for Molten-Salt Breeder Reactors, ORNL-TM-1854 (1967). 4 Table 1. Chemical Analysis of Surveillance Heats Element Content, wt % (» Heat 5065 Heat 5085 Heat 7320 Heat 67-551 Heat 67-504 Cr 7.3 7.3 7.2 7.0 6.9 Fe 3.9 3.5 < 0.05 0.02 0.05 Mo 16.5 16.7 1 12.0 12.2 12.4 C 0.065 0.052 0.059 0.028 0.07 si 0.60 0.58 0.03 0.02 0.010 Co 0.08 0.15 0.01 0.03 0.02 W 0.04 0.07 < 0.05 0.001 0.03 Mn 0.55 0.67 . 0.17 - 0.12 0.12 v 0.22 . 0.20 <0.02 < 0.001 0.01 P 0.004 0.0043 0.002 0.0006 0.002 S 0.007 0.004 0.003 < 0.002 0.003 Al 0.01 0.02 - 0.15 < 0.05 0.03 Ti 0.01 < 0.01 0.65 1.1 < 0.02 Cu 0.01 0.01 0.02 0.01 0.03 0 0.0016 0.0093 0.001 0.0004 < 0.0001 N 0.011 0.013 0.0002 - 0.0003 0.0003 Zr 0.876 ev) 3.8 x 1020 1,2 x 102} 2,5 x 10!® 2.8 x 10! 1,6 x 10! 5.0 x 10! 3.7 x 10! 9,1 x 10?° 3,9 x 101? (> 50 kev) 1.2 x 1020 3,7 x 10°° 2,1 x 1017 8.5 x 1020 4.8 x 1020 4,2 x 10'? 1,1 x 10*! 1.1 x 102°© 3.3 x 101° (>1.22 Mev) 3.1 x10'% 1.0x 10?0 5,5 x 1018 2.3 x10°° 1,3x10%0 1,1 x10'° 3.,1x10%° 0.8x10%° 8,6 x 1018 (> 2.02 Mev) 1.6 x 101° 0.5 x 1029 3,0 x 101% 1,1 x 1020 0.7 x10%°® 6,0 x 10'% 1.5 x 102° 0.4 x 102° 3,5 x 108 Heat 5081 21545 5065 5065 67-502 5065 5065 7320 67-504 Designations 5085 21554 5085 5085 67-504 5085 5085 67-551 ®Information compiled by R. C. Steffy, Resctor Division, ORNL, July 1969. ¢ o Revised for fullepower operation at & Mw. 4)‘( » «) Testing Techniques The laboratory creep-rupture tests of unirradiated control speci- mens were run in conventional creep machines of the dead-load and lever- arm types. The strain was measured by a dial indicator that showed the total movement of the specimen and part of the load train. The zero strain measurement was taken immediately after the load was applied. The temperature accuracy was +0.75%, the guaranteed accuracy of the Chromel-P—-Alumel thermocouples used. The postirradiation creep-rupture tests were run in lever-arm machines that were located in hot cells. The strain was measured by an eXtensometer‘with rods attached to the upper and lower specimen grips. The relative movement of these two rods was measured by a linear differ- ential transformer, and the transformer signal was recorded. The accuracy of the strain measurement is difficult to determine. The exten- someter (mechanical and electrical portions) produced measurements that could be read to about *0.02% strainj however, other factors (tempera- ture changes in the cell, mechanical vibrations, etc.) probably combine to give an overall accuracy of +0.1% strain. This is considerably better than the specimen-to-specimen reproducibility that one would expect for relatively brittle materials. The temperature measuring and control system was the same as that used in the laboratory with only one excep- tion. In the laboratory, the ccntrol system was stabilized et the desired temperature by use of a recorder with an expanded scale. In the tests in the hot cells, the control point was established by settlng the controller Wlthout the ald of the expanded-scale recorder. This error _'and the thermocouple accuracy combine to give a temperature uncertainty of about *1%.- The ten31le tests were run on Instron Universal Testlng Machines. The strain measurements were’ taken from the crosshead travel and gener- . _ally are accurate to *2% straln. The test environment was air in all cases. Metallegrepnic examina- tion showed that the depth of oxldat;on was small, and we feel that the environment did not sppreciasbly influence the test results. EXPERIMENTAL RESULTS Visual and Metallographic Examination J _ e W. H. Cook was in charge of the disassembly of the core surveil- lance fixture. As shown in Fig. 1, the assembly was in excellent mechan- -ical condition when removed. The Hastelloy N samples were more discolored than noted previously; however, surface marking such as numbers were readily visible. The detailed appearance of the stringer has been described previously by Cook.® The Hastelloy N surveillance rods located outside the core were oxidized, but the oxide was tenacious. Metallographic examination of the Hastelloy N straps that held the graphite and metal together revealed intergranular cracks. A typical 8W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 87-92. - R-4B618 Fig. 1. Overall View of MSRE Surveillance Assembly Removed After Run 18. Parts of this assembly had been exposed to the salt for , . 22,533 hr at 650°C. The center portion is graphite and the long rods gfi; are Hastelloy N. - «} «) “s_Haste1loy N'survelllance rods. Typlcal.photomlcrographs of heat 5065 . after exposure to the core for 22,533 hr are shown in Fig. 3. The crack is shown in Fig. 2 and extends to a depth of about 3 mils. A similar strap on the modified samples which had been in the reactor for 7244 hr had cracks to a defith of about 1.5 mils. These straps are about 0.020 in. thick and they likely encountered some deformation while being removed. However, the cracks were quite uniformly spaced on both sur- faces of the straps, and their genefal gppearance attests to a general corrosion that rendered the grain boundaries extremely brittle. Examina- ‘tion of unirradiated control straps failed to reveal a similar type of cracking. R-48571 Fig. 2. Typlcal Mlcrostructure of a Hastelloy N (Heat 5055) After Exposure to the MSRE Core for 22,533 hr at 650°C. This material was used for straps for the surveillsnée assembly. As polished. 500X. These observatlons led to the examlnatlon of tabs from the surveil- lance stringers. Small sections were cut from the centers of the unetched view in Flg. 3(a) shows the surface layer that led to the dis- colored appearance and a s1ngle graln boundary that is visible. Much of the surface layer looks metallic, but this is difficult to judge on Fig. 3. 10 R-48860 Typical Photomicrographs of Hastelloy N (Heat 5065) Exposed to the MSRE Core for 22,533 hr- a.t 650°c, (a) As polished. (b) Etchant: aqua regia. 500x. «) ) 11 such a thin film. The etched view in Fig. 3(b) reveals some carbide prec1p1tat10n near the surface and grain boundaries that are generally lined with carbides. Heat 5085 was exposed an identical time and typical photomicrographs are shown in Fig. 4. The unetched view in Fig. 4(a) shows a modified grain boundary structure to a depthiof about 2 mils. Etching [Fig. 4(b)] reveals a grain boundary network of car- bides. The grain boundaries near the surface seem to etch differently from the rest of the sample, but little more can be said. We interpreted these observétions as being indicative of some cor- rosion and performed one further crude experiment to reveal the depth of this attack. One tensile sample had been cut too short for testing, and we bent the remaining portion in a vise. The sample was then sectioned and examined metallographically; the resulting photomicrographs are shown in Fig. 5. The tension side cracked to a depth of about 4 mils whereas the compression side did not crack. Both-sides etched abnormally to a depth of about 4 mils. | | ' Samples of heats 5065 and 5085 that were exposed to the static bar- ren salt in the control fac111ty'were examined. Figure 6 shows'typical photomlcrographs of heat 5065 after exposure for 22,533 hr. There is some surface roughening, but no structure modification near the surface such as that shown in Fig. 3 for the sample from the reactor. Likewise, " heat 5085 (Fig. 7) showed some surface effects that were minor compared with its irradiated counterpart in Fig. 4. Thus, there is little doubt that the samples in the core experienced some modifications, apparently to a depth of 3 to 4 mils. This alters up to about 12% of the sample cross section and can be expected to influ- ence the mechanical properties;“ We shall dwell fUrther on thls very 1mportant subaect later in this report._ _ A sample of heat 5085 fram the core was examined by transm1sszon | electron mlcroscopy ThlS sample had received suff1c1ent thermal fluence o transmute about 97% of the 108 4o helium. The helium bubbles are obvious in Fig. 8. Another point of concern‘was the formation of voids in the material due to fast neutrons. No defects other than helium bub- bles and dislocations were present. Fig. 4. to the MSRE Core for 22,533 hr at 650° (a) As polished.- (b) Etchant° glycerla regia. 500x Typical Photomlcrographs of Hastelloy N (Heat 5085) Exposed ( a) b «) " "’) oy ( 8 R-50280 LIRS a4 Fig. 5. Typical Photomlicrographs of a Hastelloy N (Heat 5085) Sample Exposed to the MSRE Core for 22,533 hr at 650°C. The sample was bent in a vise. (a) As polished, tension side. (b) As polished, tension side. (¢) Etched, tension side. (d) Etched, compression side. Etchant: aqua regia. 500X. Reduced 27%. Fig. 6. Static Barren Fuel Salt for 22,533 hr at 650°C. (a) As polished. (b) Etched. 14 fl v-98169 Typical Photomicrographs of Heat 5065 After Exposure to Etchant: glycerie regia. 500X. —~—— i ¢ ] <) () ‘Fig. 7. Typical Photomicrographs of Heat 5085 After Exposure to ~Static Barren Fuel Salt for 22,533 hr at 650°C. (a) As polished. (b) Etched. Etchant: glyceria regia. 500X. “) . ~~ . uM 5mfe , | 0 o~ . O N < n¥e8 ..mtm_b o 08 3 | $acs L - o f -ou . Au@.‘uw.u t "y 88 ” 5850 _ lauul | . ~ ol QO | o™ o o) fi o PE.88 | Qd &4 O HI.&P.W 4 ou o G4 M | Oflma.mw . | Ao T . . o g X SR EES T y/oz En.nw SRES o o ~ N 288 . -rl.m_ | O~ = _ n3rvm 03._me S0 3= 42 w0 ol c2n2m v N NN : r~ - e = g g~ A d 0 S o] o O X ._.w 0 M N | a3 8 a m l/n g8« 48 amoo &~ - & Q4D 0O OS3P A g od &tmn% Oro wt Y o nBEH ) 8 EX S | B s - gm P <) § ) a) 17 MEchanical-Pfoperty Data — Standard Hastelloy N Two heats of standard Hastelloy N were exposed to the MSRE core environment for 22,533 hr at 650°C and received a thermal neutron fluence of 1.5 x 10?1 neutrons/em®. Similar control samples were exposed to static barren fuel salt for a corresponding length of time. The results of tensile tests on heat 5085 are summarized in Tables 3 and 4 for | unirradiated and irradiated samples, respectively. The fracture strains are shdwn as a function of test temperature in Fig. 9. With increasing temperature the unirradiated samples exhibit first an increase in frac- ture strain, then a sharp decrease, and then an increase. The irra- diated samples followrthe same general pattern except for the absence Table 3. Results of Tensile Tests on Control Samples of Heat 5085> Strain ___§E£E§§;_B§i_ Elongation, % Reduction True Test Sfii;;gin z:gg:r- (gngl) Yield 312;??2? Uniform Total in(2§ea Fg::;gie (°c) (%) 10410 25 - 0.05 ' 48,100 108,500 40.5 40.6 32.8 40 10409 200 0.05 39,300 101,400 44.7 45.1 33.4 41 10408 400 0.0 35,400 94,900 48.2 48.9 39.3 50 10407 400 0.002 35,200 98,500 48.6 49.0 36.3 45 10406 500 0.05 34,000 92,800 - 48.0 48.8 40.0 51 10405 500 0.002 35,800 90,100 37.5 38.2 28.8 34 10416 550 0.05 32,700 88,900 51.6 52.2 37.1 46 10417 550 0.002 36,500 79,500 29.4 30.1 24.1 28 10418 - 600 0.05 33,400 82,200 37.3 7.9 28.6 34 10420 600 0.002 35,200 69,000 25.7 26.9 19.1 21 10421 650 0.05 32,600 75,200 31.0 32.0 28.9 3% 10422 650 0.002 34,300 65,600 23.3 23.9 23.9 27 10404 650 0.0002 35,700 64,500 . 24.9 25.5 . " 19.7 - 22 10412 650 0.000027 30,000 59,700 15.2 22.2 2L.2 24 10423 760 0.05 33,900 64,900 25.6 27.5 19.7 - 22 10424 760 ~ 0.002 31,600 52,000 12.2 27.0 31.6 38 10426 < 850 0.05 33,600 50,600 10.0 33.0 37.1 - 46 10425 850 0.002 24,100 . 26,300 4.2 3607 25.3 29 .+ Banmealed 2 hr -at 900°C;-expééed'fo a static vessel of'baireh'fuél galt for 22,533 hr at 650°C. , | | 18 Table 4. Postirradiation Tensile Properties of Hastelloy N (Heat 5085)% Test i True Stress, psi : Specimen Temper- Strain __—L-E_—Ultimat Elongation 4 Reduction Fracture Namber ature oO0%,) Yield goqy® Uniform Total 1RQY€® gtrain (°c) 7 | (%) 8806 25 0.05 53,900 89,000 22.0 22.1 19.6 22 8805 200 0.05 44,800 85,600 26.4 27.0 26.4 31 8804 400 0.05 41,600 80,900 30.2 30.5 24.3 28 8803 400 0.002 41,300 80,100 28.4 29.2 26.8 31 8802 500 0.05 40,800 73,200 24.8 25.0 23.2 26 8801 500 0.002 39,400 61,100 1n.7 12.4 13.9 15 8812 550 0.05 37,900 62,600 144 14.9 4.3 15 8813 550 0.002 36,200 49,100 5.8 6.2 - 2.1 10 8814 600 0.05 36,700 56,200 9.8 10.5 13.3 14 8816 600 0.002 39,000 48,500 4.5 4.9 6.7 7 8817 650 0.05 36,400 50,800 8.8 9.3 9.1 10 8818 650 0.002 37,200 42,800 4.7 5.0 5.8 6 8800 650 0.002 35,200 37,700 2.1 2.4 2.8 3 8799 650 0.000027 32,900 34,500 . 1.2 1.4 6.1 6 8819 760 0.05 30,700 36,300 6.1 6.4 6.4 7 8820 760 0.002 35,100 - 35,100 1.7 2.0 2.5 3 8821 850 0.05 32,100 32,700 2.3 2.3 2.1 2 8822 850 0.002 22,200 22,200 1.6 1.7 2.6 3 ®annealed 2 hr at 900°C prior to insertion in reactor. Irradiated to a thermal fluence of 1.5 X 1021 neutrons/cm? over a period of 22,533 hr at 650°C. ORNL- DWG 70-7290 80 t | { 1 STRAN RATE UNRRADIATED RRADIATED 70 bt 0.05 o . 0.002 a a 0.0002 a . 60 I 0.000027 - - TR NN FRACTURE STRAN (%) p 30 "E-ix ] L //”/ \\0\ \w-__ . e 20 \ 0 ‘\\\ - '--..-,_.- A ha =TT ° 0 100 200 300 400 500 600 700 800 800 TEST TEMPERATURE (°C) - Fig. 9. Fracture Strains of Hastelloy N (Heat 5085) After Removel from the MSRE and from the Control Facility. All samples annealed 1 hr at 900°C before irradiation for 22,533 hr at 650°C to a thermal fluence of 1.5 X 10?2 neutrons/cm?. | i» «} +) ) " a strain rate of C.05 min 19 of a ductility increase at high temperatures. However, the levels of the fracture strains are lower for the irradiated material over the entire temperature range. For both materials the fracture strain decreases with decreasing strain rate. Another difference in behavior between the irradiasted and unirradiated samples is that at the highest strain rate (0.05 min~!) the fracture strain begins its precipitious drop at a lower temperature for the irradiated material. Further characteristics of the effects of irradiation on the ten- sile properties of heat 5085 are apparent when the ratios of the irra- diated and unirradisted properties are compared (Fig. 10). The yield stress is from 10 to 20% higher for the irradiated material at test tem- peratures up to 760°C. The ultimate tensile stress is about 20% lower for the irradiated material and drops even further as the test tempera- ture is increased above 500°C. The fracture strains of the irradiated samples are about 50% of those of the unirradiated samples up to about 500°C, above which the reduction is even greater. ORNL- DWG 70 - 720¢ 1.4 1.2 O e _ © <& a N E 1.0 ~ o 9 o S~ Q . % 0.8 Ky & 2 0 b : a < I : F—l | [=] : a g 0.6 — !_._\ \._____ E Coo 0.4 — O YIELD STRESS - o | ~~g 0.2 = A ULTIMATE STRESS " — ' ~ 0 FRACTURE STRAN = | 1. . ‘ \\ O 100 200 300 ' 400 500 600 700 _ 80O 900 : © TEST TEMPERATURE (°C) , .Fig; 10. Comparison 6fr£helTehsile'Properties of Control and Irra- diated Hastelloy N (Heat 5085). Samples irradiated to a fluence of 1.5 x 1021 neutrons/cm? over a period of 22,533 hr at 650°C. Tested at e 7 20 The results of tensile tests on the samples of heat 5065 are sum- marized in Tables 5 and 6. The fracture strains of these samples are shown in Fig. 11 as a function of test temperature. The results are quite similar to those shown in Fig. 9 for heat 5085. A notable excep- tion is the much higher fracture strains at low temperatures of heat 5065 after irradiation. The ratios of the irradiated and unirradiated ten- sile properties are shown in Fig. 12. The yield stress was not altered appreciably by irradiation. The ultimate tensile stress was decreased about 10% by irradiation at low test temperatures and up to 35% at high temperatures. Similerly, the fracture strain was reduced about 10% at low temperatures, and this reduction progressed to about 85% at high test temperatures. The progressive change in the fracture strain'with increasing fluence is illustrated in Fig. 13 for heat 5085. The fracture strain has been reduced over the entire range of test temperatures investigated. A similar trend was noted at a slower strain rate of 0.002 min~1 (Fig. 14), although the absolute values of the fracture strains were lower than noted in Fig. 13 at the higher strain rate of 0.05 min~!. These samples have experienced various holding times at 650°C and some of the property changes can be attributed to thermal aging. The frac- ture strains for several sets of control semples of heat 5085 are com- pared in Fig. 15. At low test temperatures and at 760°C and above the fracture strain seems to show a progressive decrease with increasing holding time at 650°C. At intermediate temperstures the behavior is more complex. At a test temperature of 650°C, the results indicate a progressive decrease in fracture strain up to an exposure time of A15,289 hr and then an increase with further aging time. | Heat 5065 was not included in one set of surveillance samples, but there are sufficient daté to follow the property changes with fluence and aging time at 650°C. The fracture strain is shown in Fig. 16 as a function of temperature for various fluences. At low test temperatures, excluding the one'apparently anomalous point, the fracture strain was | actually higher for irradiation to fluences of 1.3 and 2.6 X 101° neutrons /cm?. Higher fluences reduced the fracture strain at low test temperatures, but not to values as low &s noted in Fig. 13 for heat 5085. e m 3 a) " 1 ay 21 Table 5. Results of'Tensile Tests on Control Samples of Heat 506Sa Test . ' : True . —oLress, pst Specimen Temper- S;:::n StressUltiiate Elongation, % Rzfi?:::Zn Fracture Number ature . “oqy Yield Uniform Total Strain u (min~1) Tensile (%) (°c) (%) 10384 25 0.05 61,200 126,500 48.5 49.8 36.77 46 10388 200 0.05 42,000 107,900 475 49.4 40.16 51 10392 400 0.05 42,600 102,000 48.9 50.8 45.31 60 10383 400 0.002 43,900 106,000 47.5 4707 39.30 50 10391 500 0.05 42,200 99,400 48.1 49.6 37.54 47 10378 500 0.002 45,800 98,700 45.5 46.2 34.08 42 10395 550 0.05 - 42,500 . 98,800 48.2 49.9. 38.14 48 10375 550 0.002 43,800 84,600 25.5 26.1 27.51 32 10377 600 0.05 41,800 93,800 42.0 43.0 34.08 42 10393 600 0.002 39,400 76,700 25.6 26.1 34.29 42 10382 650 0.05 41,500 81,900 26.4 26.8 23.84 27 10394 650 0.002 39,200 71,700 23.1 23.6 20.89 23 10397 650 0.0002 41,800 71,200 18.8 19.6 17.88 20 10400 650 0.000027 42,900 60,800 7.5 - 19.7 24.57 28 10385 760 0.05 32,900 70,500 24.3 27.3 22.85 26 10398 760 0.002 41,100 42,100 5.7 39.6 45,48 61 10390 850 0.05 35,500 49,500 8.7 38.1 45.31 60 10379 850 0.002 24,800 25,100 2.9 56.5 47.15 64 Bpnnealed 2 hr at 900°C prior to insertion in the reactor. Exposed to a static vessel of barren fuel salt for 22,533 hr at 650°C. Table 6. Postirradiation Tensile Properties of Hastelloy N (Heat 5065)% Test . . ' . True Specimen = Temper- St;::n : Stressuizizat Elongation, % R:guggzzn Fracture Number atgre (min=1) Yield Tensilee Uniform Total (%) Strain (°c) (%) 8833 25 0.05 56,700 109,500 42.5 42.8 3.32 3 8832 200 0.05 46,000 99,900 42.8 4.6 4.44 5 8831 400 0.05 43,700 - 94,300 | 39.0 39.3 31.22 37 8830 400 0.002 43,200 91,300 35.7 37.2 27.31 32 8829 500 0.05 43,600 82,100 29.1 29.5 22.98 26 8828 500 0.002 39,800 63,200 10.9 11.3 10.70 11 8839 550 0.05 41,200 68,700 15.5 15.7 16.83 18 8840 - 550 0.002 - 37,800 - 57,700 7.1 7.2 8.93 9 8841 600 0.05 - 40,500 60,100 8.6 9.0 9.30 10 8842 600 0.002 - 41,000 - 50,400 4.3 4.5 7.40 8 8843 650 - 0.05 39,700 52,100 - 6.0 6.2 ~ 7.37 8 8844 650 0.002. 40,000 42,800 3.0 3.1 4.89 5 8827 650 0.0002 34,300 36,100 2.1 2.2 3.30 3 . 8826 - . 650 0.000027 28,400 -36,900 1.1 1.2 1.1 1 - 8845 760 0.05 37,800 41,400 2.5 2.7 3.33 3 - 8846 - 760 0.002 - 35,200 . 35,200 1.3 1.3 2.86 3 8847 - 850 0.05 35,400 . 35,400 1.5 1.6 2.22 2 8848 g50 - 0.002 - 20,300 20,300 1.0 1.0 0.16 0 ®annealed 2 hr at 900°C prior:to-insertion in the reactor. Irradiated to a thermal fluence of 1.5 X 1020 neutrons/em? over a period of 22,533 hr at 650°C. 22 ORNL-~ DWG 70-7292 60 : o T T N\ / ; ~ | fi / = 40 \ K A 2 8 AY, & = . <{ £ \\ w 30 / - - Q & oo L STRAN L RATE UNIRRADIATED IRRADIATED \ ~=>2 (mifl_i) -~ @ 1 0.05 o ® 'Y | 01 0002 - a a AN 0.0002 o s ~L T 0.000027 ° - —p o | _ | ! - o 100 200 300 400 500 600 700 800 900 TEST TEMPERATURE {°C) Fig. 11. Fracture Strains of Hastelloy N (Heat 5065) After Removal from the MSRE and from the Control Facility. All samples annealed 2 hr at 900°C before irradiation for 22,533 hr at 650°C to a fluence of 1.5 X 10?1 neutrons/cm?. ORNL—DWG TO- 7293 1.2 0 o 1.0 /—"" ~— o b Q . et 2 T~ @ o a = 5 \ - 8 os = g o =1 g S 04 B © YIELD STRESS o« a ULTIMATE STRESS o FRACTURE STRAIN o 0.2 \\ = ‘-—______ o 0 o 100 200 300 400 500 600 700 800 500 TEST TEMPERATURE (°C) Fig. 12. Comparison of the Tensile Properties of Control and Irra- diated Hastelloy N (Heat 5065). Samples irradiated to a fluence of 1.5 X 102! neutrons/cm® over a period of 22,533 hr at 650°C. Tested at a strain rate of 0.05 min~1. b1 (n . n 0 23 ORNL-—-OWG 70~-7294 80 © ANNEALED 2 hr AT 900°C a 1,3 x 10" neutrons /em2, 1100 hr 0 2.6 % 10'? neutrons/cm2, 20,789 hr 70 20 2 v.4.3 x10%Y neuirons/cm<, 4800 hr 094 x1029 neutrons/cmé<, 15, 289 hr < {5x 102 neufrons/cma. 22,533 hr 60 o o 7 ] 0 & 50 . L ! a / z ] ) b g / v ° 40 & / \ k_/ / =2 v - :/ Q 5 . & 30 N [ et 20 \ v ”\ T \\\ 10 3 %\#—- 0 0 100 200 300 400 500 600 700 800 900 Fig. 13. Fracture Strains of Hastelloy N (Heat 5085) After Irra-~ diation to Various Thermal Fluences in the MSRE. rate of 0.05 min™?!. TEST TEMPERATURE (°C) Tested at & strain ORNL-DWG €9-4467R * I | NEUTRON FLUENCE TIME AT {neutrons /em®) 650° C {hr} a0 |— 8 13x10% 11,000 UNIRRADIATED /o,/ o 26x0° 20,789 2 v 13x4020 4800 =z o 94x10%° 15289 . A\ - <30 1~ e 15x102t 22,533 — T‘ & _ - ‘ o w \ A & - asx02t\ fi £ 20 5 2 ) \ & . - ! s o $ 0 0 100 200 300 - 'n-:sf TEMPERATURE (°C) Fig. 14 Postlrradiation Ten31le Propertles of Hastelloy N (Heat 5085) After Exposure to. Various. Neutron Fluences. strain rate of 0.002 min~l. Tested at a ORNL-DWG 69-4468R 60 l o O ] o 50 1, T / : —— o /l‘l - \ \ / ——’“—-— \ -t \ \Y o = 40 |g—== ST - | 1\ 2 z \ \ 47 @x \ ) = ®, /( n 30 3 Y Z2 Y SN & wi ./ g % .— g N\ T 20— AGING TIME AT Ny 650°C (hr) oo o 4800 0 — A 15,289 e 22,533 Lo 0 100 200 300 400 500 600 700 800 TEST TEMPERATURE (°C) Fig. 15. Variation of the Tensile Properties of Hastelloy N (Heat 5085) with Aging Time in Barren Fuel Salt at 650°C. strain rate of 0.05 min™1l. Tested at a ORNL-DWG 70—7295 70 - — O - 1 A\\ ° 7 TSIV 50 / el [N i b. . £ ! ¢ \ / = NN : | " \ \ g \ ' 3 a s 30 N \ = \ \<1.3 %1019 [T 20 95x 107, \ | 26x10° o UNIRRADIATED \ \ 4 1.3 x 10'? neutrons /cm? ‘\ ‘]' o 2.6 x40' neutrons/em? ol o zc"neu rons/ mz (5% 102! \ ‘b\ o 9.5 x 10" neutrons /cm X\ \ N . a ¢ 1.5u10""-m:ulrons/.:fn2 \&\ \ 1&-—_5 o | — 0 100 200 300 400 5C0 600 700 800 TEST TEMPERATURE (°C) 900 Fig. 16. Variation of the Postirradiation Tensile Propertles of Hastelloy N (Heat 5065) with Thermal Neutron Fluence. i 4 1) ) 25 At test temperatures above 550°C the fracture strain decreases progres- sively with increasing fluence. As shown in Fig. 17 some of these changes in fracture strain can be attributed to thermal aging at 650°C. Except at test temperatures above 750°C, the fracture strain is lowest for the material aged 15,289 hr and is improved after aging 22,533 hr. The changes in properties at low test temperatures are less than those for heat 5085 (Fig. 15). ORNL—OWG 70-7296 70 60 ' /[ g i | ! | | ! | | | | 4 j FRACTURE STRAIN (%) o H o Qo o / I 4/ R \ - N \\\ e 3 ] U / / 1 \ \ AGING TIME "/ {hr} B0 {0 +— o 15,289 a 22,533 o | _ 0 100 200 300 400 500 600 700 800 900 - TEST TEMPERATURE (°C) Fig. 17. Effects of Thermal Aging at 650°C on the Tensile Proper- ties of Hastelloy N (Heat 5065) at a Strain Rate of O. 05 min~1. As d1scussed prev1ous1y in thls series of reports, the changes in fracture strain at low temperatures were not expected. Although the fracture stralns have not reached values below 20%, we are still 1nter- ested in its progress1on Our experlence to date is summarized in F1g._18. If the noted effects were due simply to thermal aging, then " the results should correlate with the time at 650°C. Obviously such a _ correlation does not exist. ‘Where data are available for_pairs of irra- diated and unirradiated samples, the irradiated sample has the lower 26 ORNL-DWG 70-7297 70 T T I HEAT IRRADIATED UNIRRADIATED 5065 ¢ o 60 - 5085 = ’ o . ® 50 o o o ® ) Pl . 40 - a = & . . - 1 1 | w ! 1 1 w 30 ¥ ! s 1 &« ' : 1 > i 1 ! ! 5 ! i : ! q 1 t ' I Id ' I : I = L 29 S o 8 g' b Q e Q 2 % = * = ® o " " < ° o 10 0 0 q 8 t2 16 20 (xt0%) TIME AT 650°C (hr) Fig. 18. Variation of the Fracture Strain at 25°C with Annealing Time and Thermal Fluence. fracture strain. embrittlement. must be associated with carbide precipitation. This indicates that irradistion has a role in the We previously showed that the ductility could be recovered by an anneal of 8 hr at 871°C and concluded that the changes 9 The higher susceptibil- ity of heat 5085 to this type of embrittlement is not understood, since heat 5065 actually has a higher carbon concentration (Table 1, p. 4). The data in Fig. 18 defy extrapolation, so one cannot conclude whether the room temperature embrittlement is likely to become worse. ‘In these discussions of tensile properties we have emphasized the changes in fracture strain and made little mention of strength changes. Some pertinent data are summarized in Table 7 for heat 5085. The sam- ples were tested at 25 and 650°C and include thrée histbries: (1) as annealed, (2) thermally aged, and (3) irradiated. The changes in yield strength are likely significant, but the main point is that they are °H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — Third Group, ORNI~TM-2647 (1970), p. 7. " 0 3) 27 Table 7. Comparison of the Tensile Properties of Heat 5085 Before and After Irradiation® Ultimate Tensile Heat Treatment Yield Stress, psi Stress, psi Fracture Strain, % | at 25°C at 650°C at 25°C at 650°C at 25°C at 650°C Annealed 2 hr at 900°C 51,500 29,600 120,900 75,800 53.1 33.7 Annealed 2 hr at 900°C and 48,100 32,600 108,500 75,200 40.6 32.0 aged 22,533 hr at 650°C « : | Annealed 2 hr at 900°C, 53,900 36,400 89,000 50,800 22.1 9.3 irradiated for 22,533 hr at 650°C to a thermal fluence of 1.5 X 10?1 neutrons/cm? 8Strain rate of 0.05 min~1. small considering the long thermal history of 22,533 hr. The ultimate tensile strength is reduced due to the lower fracture strain and the resulting fact that the tést is interrupted before it reaches the high " ultimate strength.noted for the as-annealed material. Similar results are shown in Table 8 for heat 5065. The yield strength is increased by aging,‘but the main effects are on the fracture strains and the ultimate tensile strengths. Table 8. Comparison of the Tensile Properties of Heat 5065 Before and After Irradiation® Ultimate Tensile Fracture Strain, % Stress, psi ~ =0° #t 25°C at 650°¢ °¢ °C at 650°C Yield Stress, psi at 25°C at 650°C Heat Treatment Amealed 2 hr at 900°C 56,700 32,100 126,400 81,200 55.3 34.3 Annealed 2 hr at 900°C and 61,200 41,500 126,500 81,900 49.8 26.8 aged 22,533 hr at 650°C o | - _ Annealed 2 hr at 900°C, 56,700 39,700 109,500 52,100 42.8 6.2 irradiated for 22,533 hr at 650°C to a thermal fluence of . _ 1,5 X 102! neutrons/cm? aStrain rate of O.OSVmin'1.~;‘ 28 The results of creep-rupture tests on heats 5085 and 5065 from the &a)" fourth group of surveillance samples are summarized in Tables 9 and 10. - iy The results are most interesting when they are compared with those obtained on previous groups of surveillance samples so that the progres- . sive changes in properties can be noted. The stress-rupture properties of heat 5085 are shown in Fig. 19. of samples (1.5 X 102! thermal neutrons/cm?) are not detectably differ- ent from those irradiasted previously to a thermal fluence of 9.4 X 102° The rupture lives of the last group neutrons/cm?. In the unirradiated condition, the rupture lives for the samples aged 22,533 hr are slightly less than in the as-annealed condi- tion, but greater than those observed for samples aged for lesser times. Table 9. Creep-Rupture Tests on Heat 5085 at 650°C Test Specimen Stress Rupture . Rupture Reduction Mgnimum : . . . reep Number Number (psi) %;fi? S?;sln 1n(25ea Rate (%/hr) Unirradiateda . 7978 1406 55,000 22.1 21.0 23.4 0.288 . 7897 10415 40,000 500.1 24.5 28.5 0.0299 7896° 10414 32,400 2500.7 12.7 9.4 0.0045 7895° 10413 27,000 2187.1 5.40 5.40 0.0026 7894° 10412 21,500 2064.9 2.10 2.44 0.0011 7893° 10411 17,000 2000.0 1.1 0 0.0007 Irradiated® R-969 8811 40,000 0.6 = 1.1 0.78 R-958 8810 32,400 31.6 0.54 0.0067 R-956 8809 27,000 60.2 0.57 '0.0055 R-957 8808 21,500 362.0 0.63 0.0010 R-964 8807 17,000 426.4 0.69 0.0004 ®Annealed 2 hr at 900°C, exposed to static barren fuel salt for 22,533 hr at 650°C. bDiscontinn.ed prior to failure. CAnnealed 2 hr at 900°C, exposed to MSRE core for 22 650°C, received thermal fluence of 1.5 X 10?! neutrons/cm®. 3 533 hr at 'fl 29 Table 10. Creep-Rupture Tests on Heat 5065 at 650°C : . Minimum Test Specimen Stress Rggture _ Ruptu?e R?ductlon Creep . ife Strain in Area Number Number (pSl ) (hr) ( %) ( %) Rate | (%/hr) Unirra.dia.teda 7978 10406 55,000 22.1 21.0 23.4 0.34 7892 10388 40,000 = 651.2 36.3 29.9 - 0.0331 '7891b 10387 32,400 = 2144.3 4 7 37.7 0.0096 7890b 10386 27,000 2187.1 8.8 7.4 0.0040 7889b 10385 21,500 2373.3 3.6 2.2 0.0013 7888 10384 17,000 2000.0 2.8 0 | Irradiatedc R-966 8838 40,000 0 R-962 8836 - 32,400 - 22.7 0.25 0.0075 R-954 8836 27,000 - 40.5 0.44% 0.0061 R-960 8835 21,500 46.2 0.47 0.0051 R-963 8834 17,000 1567.3 0.86 0.0004% @pnnealed 2 hr at 900°C, exposed to static barren fuel salt for 22,533 hr at 650°C. | bDiscontinued prior to failure. CAnnealed 2 hr at 900°C, exposed to MSRE core for 225533 hr at 650°C, received thermal fluence of 1.5 X 1021 neutrons/cm?. The minimum creep rates are shown as & function of stress in Fig. 20 for heat 5085. As noted previously, neither irradiation nor aging has a detectable effect on the creep rate. The data deviate from - the line shown in Fig. 20 atfboth extremes. At high stresses, the irra- diated samples fail at such low Strains that a minimum creep rate is not established over‘long‘éfiough'pefiod for measurement. At low stress f_leVelS-the data deviate due to the semilog plot. The fracture strain is the parameter_most affected by irradiation, " and a variation of this parameter with minimum creep rate is shown in - Fig.'Zl for heat 5085 at_650°c;;_There has been a continual deteriora- tion of the fracture strain with increasing thermal fluence. ' Tensile and creep tests were run at 650°C at several different . strain rates and stresses. The fracture strains from these tests are plotted together in Fig. 22 although different parameters are controlled 30 ORNL~DWG 69-4470R " 70 r\ \\\ 60 UNIRRADIATED T <|0 \"n . 50 L3 NN T N - ISR o] TR & TR~ ™ \\\\\ » Q a0 ~ "% N '\\\2»\ 8 -.‘\ I ™ iy, - ., ‘l:\\‘ \ F‘h\ ~ ~ ~H SN o THERMAL R ;7 SRa i o © 30 - FLUENCE TIME AT ] Sl 3 e {neutrons/cm?) 650°C (hr) \é;c-._._ o| [& « e UNIRRADIATED O / T~ 20 |~ ® UNIRRADIATED 4800 / P T a UNIRRADIATED 15,289 o] \ ¢ UNIRRADIATED 22,533 L ol ° 13 x w0 11,000 3 x10%° neutrons/ecm? (MSRE) AND a26 x 10" 20,789 3-5 x 10%% neutrons/cm? (ORR) o 1.3 x 10%° 4800 oL 094 x 102 15,289 < 1.5 x 103 22,533 L L [t iEL 1o 10° . o' 102 103 104 RUPTURE TIME (hr) Fig. 19. Postirradiation Stress-Rupture Properties of MSRE Surveil- lance Specimens (Heat 5085) at 650°C. 70 / A 60 A A 7 o8 ¢ 4 £ 50 // p . wt a A ; 2 ‘XB&T o 40 IR £ o H S &Il -AA THERMAL = L‘l// FLUENCE TIME AT o 4o ly| o1V 0 (neutrons/cm?) 650°C (hr) fi 30 AT e UNIRRADIATED © il e ‘ A o 4.3 x to% 11,000 20 Ly Ao a 2.6 x 4% 20789 || &4 o/f/ o 1,3 x 10%° © 4800 P4l ¢ 9.4 x 102 . 15,289 10 o 1.5 x 102! 22,533 1} = UNIRRADIATED 4800 a UNIRRADIATED 45,289 o ¢ UNIRRADIATED 22,533 1074 1073 4072 -t 100 10! MINIMUM CREEP RATE (% / hr) Fig. 20. Minimum Creep Rate of Hastelloy N (Heat 5085) Surveil- ORNL—-DWG 69-4471R2 - lance Specimens from the MSRE at 650°C. ”» 4 a FRACTURE STRAIN (%) Fig. 21. N w H FRACTURE STRAIN (%) i 0 104 31 ORNL-DWG 69-4472R RANGE 2-5x1029 neutrons/cm? (ORR) LT T 1.3x10%° neutrons/cm2 2.6x10'° 1.3x10%° neutrons neutrons/cm?2 9.4x1020 1.5x102! neutrons /cm? - neutrons /em?2 10~3 102 10! 100 10! MINIMUM CREEP RATE (%/hr) Variation of ‘Fracture Strain with Strain Rate for Hastelloy N (Heat 5085) Surveillance Specimens at 650°C. ORNL-DWG 70-7298 40 IR UNIRRADIATED Lk /0’ 30 = * ----'._’-_____—"- 20 hll I 10 [ 1T HTTHI 1. IRRADIATED |/ /] 8 7 A . A 6 o TENSILE TESTS o CREEP TESTS o« d"’ 4 A LA » -"./ 2 . o—""-— . [ Lid ] Ol LLLL ] - 0 [ L _ w04 1073 02 ot - 10° 0o 108 10° STRAIN RATE (%/hr) F':i.g. 22. Fracture Strain at 650°C of Heat 5085 in the Unirradiated ~and Irradiated Conditions. TIrradiated to a thermal fluence of 1.5 X 1021 neutrons/cm® over a period of 22,533 hr at 650°C. in the two types of tests. The most striking feature is.the marked dependence of the fracture strain of the irradiated samples on the strain rate and the rather weak dependerice of the fracture strains of the unirradiated samples on the strain rate. The irradiated samples at this 32 high fluence level show a general decrease in fracture strain»with decreasing strain rate, with a possible slight increase in fracture strain at very low strain rates. The sensitivity of the fracture strain of heat 5085 at 650°C to the helium content is illustrated in Fig. 23 for three strain rates. This material contains 38 ppm B (Table 1, p. 4) that can yield an equiv- alent amount of helium when transmuted. (Several factors contribute to the relationship that 1 ppm of natural boron by weight leads to 1.1 ppm He on an atomic basis.) The points shown in Fig. 23 all come from surveillance samples from the MSRE. The two higher strain rates are obtained by tensile testing, and the fracture strain decreases with increasing helium content. The 0.1% strain rate is the creep rate that corresponds to the lowest fracture strain (Fig. 21). The fracture strain drops abruptly with the presence of 1 ppm He and then decreases gradually with increasing helium content. The creep properties of heat 5065 are illustrated in a series of graphs similar to that just presented for heat 5085. The stress-rupture properties at 650°C are shown in Fig. 24 for heat 5065. The rupture ORNL-DWG €9-7297R2 o THERMAL FLUENCE \g” ‘:\o‘ & (neutrons/cm?) 3 q® » 0" 36 32 28 FRACTURE STRAIN (%) - - N N o n » o D F 0o 02 0.5 { 2 5 10 20 50 ~ HELIUM CONTENT (ppm} Fig. 23. Variation of the Fracture Strain with Calculated Helium Content and Strain Rate for Hastelloy N (Heat 5085) at 650°C. n 3} 3 33 ORNL-DWG 68-4474R 80 : : T T7TTT — T TTT717 THERMAL FLUENCE TIME AT (neutrons/m?) 650°C (hr) 70 . - ¢ UNIRRADIATED o | \\ u UNIRRADIATED 15,289 i "\ ¢ UNIRRADIATED 22,533 \\\ o |3x10 11,000 €0 N a26 HOZO 20,789 1| . . '\\‘ 094 MO21 15,289 . N N ¢ 1.5 x10%" - 22,533 o . "N o Sh¥ E - ~’. LF \ a ~— St AVERAGE ORR ~_ [\ | N A g 40 2-5x1020 peutrons/em®—= [~1 W] Y g i . . . \-.::,:.. ‘ N 2 Tttt e LR e g:_l 30 "--.___-- Q L ‘ k= 9 -""-L.“_ \\:-..\ \ 9 ---\-o.. ™ 20 S q o g \ 10 ) 0 107! 10° T} 102 10° 104 RUPTURE TIME (hr) Fig. 24. Stress-Rupture Propertles of MSRE Surveillance Specimens (Heat 5065) at 650°C. times are equivalent for the samples irradiated to the two highest fluences. The rupture times are also quite comparable with those shown in Fig. 19 for heat 5085. The unirradiated samples that were aged for 22,533 hr at 650°C in static barren fuel salt have longer rupture lives than the as-receivedrmaterial. The minimum creep rates are shown in Fig. 25 and show a lack of sensitivity to any of the variables being studied. The fracture strain is shown as a function of strain rate in Fig. 26. The fracture straln decreases with increasing fluence up to the two highest fluence levels, where the strains are about equivalent. This heat of materlalrshcws a-ductlllty minimum with the fracture strain ~ increasing slightly with’ decreasing creep rate except for the samples ”'ShOWIHg the lowest fluence. The tensile and creep test results for heat 5065 have been combined in Fig. 27. These results agaln ‘show the marked dependence of the frac- ture straln on the straln rate for the irradiated materlal. A comparison with the similar plot for heat 5085 (Fig. 22) reveals some slight dif- ferences in the fracture strains of the two heats, but shows generally ORNL-DWG 69-4475R ] T 1 reinrn T. T T V00178 A . THERMAL FLUENCE TIME AT ,/ (neutrons/em?) €50°C (hr}) /’ 60 | ® UNIRRADIATED ) ® UNIRRADIATED 15,289 Z ¢ UNIRRADIATED 22,533
    - 0 100 200 300 400 500 600 700 800 900 TEST TEMPERATURE (°C) Fig. 31. Fracture Strains of Heat 67-551 After Removal from the MSRE and from the Control Facility. All samples annealed 1 hr at 1177°C before irradiation to a thermal fluence of 5.1 X 102° neutrons/cm® over 7244 hr at 650°C. ORNL—DWG 70-7304 1.0 O.B o . \ 0.6 ~ © YIELD STRESS NG 04 a ULTIMATE STRESS * @ FRACTURE STRAIN \\ RATIO {IRRADIATED/UNIRRADIATED) 0.2 o 0 100 200 300 400 500 600 700 800 900 TEST TEMPERATURE (°C) Fig. 32. Comparison of Unirradisted and Irradiated Tensile Proper=- ties of Heat 67-551 After Irradiation to a Thermal Fluence of 5.1 X 10°° neutrons/cm® over a period of 7244 hr at 650°C. Tested at & strain rate of 0.05 min~1. 43 Table 16. Comparison of the Tensile Properties of Heat 67-551 Before and After Irradiation® Ultimate Tensile Yield Stress, psi Stress, psi Fracture Strain, % Heat Treatment _o__z_l’_;_ —_— s ° at 25°C at 650°C at 25°C at 650°C at 25°C at 650°C Annealed 1 hr at 1177°C = 44,600 26,900 113,600 78,900 79.6 57.3 Annealed 1 hr at 1177°C, 62,400 41,900 120,200 84,600 52.3 42.6 aged 7244 hr at 650°C Annealed 1 hr at 1177°C, 49,600 31,300 107,800 56,600 51.0 R2.6 aged 7244 hr at 650°C, irradiated to & thermal fluence of 5.1 X 1020 neutrons/cm® Srested at & strain rate of 0.05 min~! decreased by irradiation. At 25°C the fracture strain is reduced from 79.6% to 52.3% by aging, and irradiation does not cause any further change. At 650°C the fracture strain is decreased by aging and decreased further by irradiation. The creep-rupture properties of heat 7320 are summarized in Table 17. These results are compared in Figs. 33, 34, and 35 with those for unirradiated samples that were given a pretest anneal of 1 hr at 1177°C. The stress-rupture properties in Fig. 33 show that the aging treatment given the controls, 7244 hr at 750°C, increased the rupture life. The irradiated samples failed after longer times than did the unirradiated sa.mples that were sn.mply annealed 1 hr at 1177°C, but 'ruptured in shorter times than the control semples However, the minimum - ereep rates shown in Fig. 34 seem'to fall within & common scatter ‘band ~ for all conditions._ Thus, neither 1rrad1at10n nor aging seems to have a detectable effect on the minimum creep rate The fracture strain is shown as a function of strain rate in Fig. 35, The fracture stfains of the unlrradlated samples vary from about 30% in a tensile test to about 10% in a long-term creep test. The irradiated samples have lower frac- ture strains that vary from 9 5% for the fastest tensile test to 2 to 3% for creep tests. Table 17. Creep-Rupture Tests on Heat 7320 at 650°C : . Minimum : . Rupture Reduction Test Specimen Str?ss Iife Strein in Ares Creep Number Number (psi) (br) (%) (%) Rate (%/nr) Unirradiated — Annealed 1 hr at 1177°C prior to test 7013 7276 55,000 4.7 28.7 29.6 0.190 7425 10329 50,000 11.5 21.2 28.3 0.125 7014 7277 47,000 49.9 21.8 31.3 0.019% 7016 7279 43,000 103.4 18.1 20.5 0.0069 7015 7278 40,000 384.6 16.3 17.6 0.0059 Th24 10252 35,000 1602.1 27.3 30.6 0.0056 7017 7280 30,000 1949.5 10.9 19.9 0.0017 Unirradiated Controls® 7885 9462 55,000 32.5 1.2 32.9 0.25 7991 9468 47,000 224.1 15.3 26.3 0.045 7886 9463 40,000 653.7 17.9 29.3 0.019 7887 9465 40,000 501.9 7.9 _1.4 0.015 7884 9460 32,400 2420.2 6.9 4.8 0.0006 Irradiated® R-1151 9204 63,000 1.7 12.24 1.2 R-1016 9216 63,000 4.0 2.0 0.44 R-951 9217 55,000 12.3 2.7 0.20 R-955 9219 47,000 95.1 2.3 0.018 R-967 9218 40,000 329.9 3.2 0.0074 R-950 9220 32,400 2083.3 3.1 0.0058 Annealed 1 hr at 1177°C and exposed to & vessel of static barren fuel salt for 7244 hr at 650°C. bDiscontinued prior to failure. Cannealed 1 hr at 1177°C. Irradiated to & thermsl fluence of 5.1 x 10?9 neutrons/cm? over a period of 7244 hr at 650°C dTest loaded so that strain on loadlng was included. tests did not include this strain. All other ¥ » STRESS (1000 psi) 45 ORNL=—DWG 70— 7305 70 nl €0 \~\ . & \ o ™ 0 A ~. X A 1 T ° a\\ 40 - \b- N g ™M a \.o— P 20 o UNIRRADIATED CONTROLS _ 20 _ & UNIRRADIATED, ANNEALED 1hr HT7°C ® IRRADIATED, 5.1 * 1029 neutrons/em? 10 0 W 2 5 . 2 5 102 2 5 w0 2 s 40 RUPTURE TIME (hr} Fig. 33. Stress-Rupture Properties of Heat 7320 at 650°C. 70 ORNL-DWG TO-7306 \ 0 Ll 6 L1 //}4( | { tat % so AT i . o L = s / s & Q g g L ' 2 A . fi o 1 / - g 30 a1 > g ,/ -1 1-1]]© VUNIRRADIATED CONTROLS P ® IRRADIATED, 5.1x10%° neutrons /em® 20 ' - & UNIRRADIATED, ANNEALED fhr AT #177°C | | 1] 10 o - 10 2 5 w02 2 5 0% 2 5 ' 2 5 ~_ MINIMUM CREEP RATE (%/hr) L Fig. 34. Creep Properties at 650°C of Heat 7320. 46 ORNL- DWG 70~ 7307 40 T UNIRRADIATED -t 20 Sen) + — Q o o . o= . ® e o . 10 T | P e IRRADIATED . z i ] % L w e TENSILE TESTS Py % © CREEP TESTS | o 6 & I~ . ’! ¢ . T A 4 - —"-- 4 Pl .-____r" Q o 2 ° -3 -2 -1 O i 102 3 -3 2 5 102 2 5 w' 2 5 «° 2 5 1o 2 5 2 5 10 STRAIN RATE (% /hr) Fig. 35. Fracture Strains at 650°C of Heat 7320 in the Unirradiated and Irradiated Conditions. Irradiated to a thermal fluence of 5.1 X 10%° neutrons/cm® over a period of 7244 hr at 650°C. The results of creep-rupture tests on heat 67-551 with 1.1% Ti are given in Table 18. The stress-rupture properties of the control and irradiated samples are compared in Fig. 36 for 650°C. The irradiated samples generally fail in slightly shorter times than the unirradiated samples. The minimum creep rate seems to be unaffected by irradiation, although there are two data points that seem to be anomalous (Fig. 37). The fracture strains of this alloy (Fig. 38) are higher in both the unirradiated and irradiated conditions then those for heat 7320. The fracture strains of the unirradiated samples varied over the range of 43 to 25% and those of the irradiated samples varied from 22.6 to 5.8% over the range of strain rates studied. One heat of modified Hastelloy N containing 0.49% Hf (heat 67-504) was exposed to the cell environment for 17,033 hr at 650°C. The .thermal - fluence was 2.5 X 10'° neutrons/em?. This same heat of material was included previously in our surveillance program and was exposed to a 47 Table 18. Results of Creep-Rupture Tests on Heat 67-551 at 650°C . Minimum Test Specimen Stress —T_BEEEEEE“-T— R?ductlon Creep 7o . Life Strain in Area Number Number (psi) (br) (%) (%) Rate (%/nr) Unirradiated, Annesled 1 hr at 1177°C before testing 7872 6889 55,000 14.2 29.5 33.2 0.32 7871 6884 47,000 3.8 33.6 26.0 0.071 Unirradiated Controls™ 7882 9409 55,000 44.7 25.5 28.5 0.190 7880 9404 4'7 ,000 271.2 29.1 29.6 0.067 7883 9411 40,000 956.2 30.4 25.5 0.018 7881 9408 32,400 2034.9 25.2 29.9 0.0053 Irradiated’ R-1150 9150 63,000 0.9 20.8° 2.2 R-1036 9151 63,000 0.6 2.2 2.6 R-949 2162 55,000 50.5 5.8 0.079 R-~961 9165 47,000 117.1 8.3 0.058 R-948 9164 40, 000 "746.9 9.1 0.0079 R=-947 2163 32, 400 1485.6 12.5 0.0053 ®pmnealed 1 hr at 1177°C prior to exposure to a vessel of static barren fuel salt for 7244 hr at 650°C. Pannesled 1 hr at 1177°C. Irradiated to a thermal fluence of 5.1 x 10?° neutrons/cm® over a period of 7244 hr at 650°C. ®Test loaded to include straln on loading. All other tests did not include this strain. thermal fluence of 5.3 X 10%° neutrons/cm while being at temperature in the core for 9789 hr (ref. lO) In the group of samples presently being discussed, there were twoquds of heat 67-504. Two heats of material should have been_involved,fbut'postirradiation'chemical analysis revealed that both rods were made ofrthe same material. This was not discovered, until after many of the samples were tested, 80 we have several tests under dupllcate conditions. 10H. E. MecCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — Third Group, ORNI-TM-2647 (1970). ORNL-DWG T0-7308 ettt w Irra- 10 60 T ] N a b\\ ‘\‘ 50 on\‘ N = N e 40 8 T £ \\\ o © UNIRRABIATED CONTROLS . °\ w 30 E ® IRRADIATED \ o & UNIRRADIATED, ANNEALED {hr AT HTT7°C 20 " 10 o 5 10° 5 w0 2 5 w00 2 0° = 5 RUPTURE TIME {hr) Fig. 36. Stress-Rupture Properties at 650°C of Heat 67-551. diated to a thermal fluence of 5.1 X 102?° neutrons/cm® over a period of 7244 hr. 0 ORNL~DWG TO— 7309 | st 60 / = 1 - o s 50 "/ /v|1n _ " % v /./ ;:.’ 30 © UNIRRADIATED CONTROLS 20 © IRRADIATED - A UNIRRADIATED, ANNEALED thr AT HTT*C 0 0 L w3 s w? 2 5 0 2 5 W 2 5 INIMUM CREEP ATE (%/hr} ‘ Fig. 37. Creep-Rupture Properties at 650°C of Heat 67-551. Irra- diated to a thermal fluence of 5.1 X 10?° neutrons/cm? over a period of 7244 hr at 650°C. » 49 ORNL- OWG 70— 7310 60 P UNIRRADIATED ¢ . __,—‘ 40 . ___/’ o . d ® ® et ° e 20 24 T ‘ [RRADIATED : ”’/’,r g 20 §” Z /// = ] - LT w 46 ot g __.// = —-__dl—" B a1 q — E 2 \‘\o . N & TENSILE TESTS j\ o CREEP TESTS y \ | 8 ~ - SN "-...... Ny I’ \ \\ 4q "u-,..\ .“ 0 103 2 5 1072 2 5 10! 2 5 w° 2 5 10 2 5 102 2 5 10° STRAIN RATE (% /tr) Fig. 38. Comparison of the Fracture Strains of Unirradiated and Irradiated Heat 67-551 at 650°C. Irradiated to a thermal fluence of 5.1 X 10?° neutrons/cm?® over a period of 7244 hr at 650°C. The results of the postirradiation tensile tests . on heat 67-504 are given in Table 19. Thé'postirradiation fracture strains are shown in Fig. 39 as & function ofltemperature and strain rate. The fracture strain generally decreases with increasing temperature above about 400°C and with decreasing strain réte;'rThe results from the present tests and those reported préviously10 show_éqmé'Very striking-éhahges'in fracture strain with varying historiesr(Fig.’40). The samples were all initially ~annealed 1 hr at 1I77°C.béfore.being:given the treatment indicated in Fig.VAO. In the as-annealedzébnditiOn,ithe alloy has a fracture strain of 70 to 80% up to about'600°C,_where the fracture strain drops precip- ‘jtiously. Aging for 9789 hr at 650°C reduced the fracture strain at lower temperatures and increased it above about 700°C. However, the - fracture strains are in'the;range of 40 to 50% over the entire range of temperatures studied. Irradiation to a thermal fluence of 2.5 X 101° neutrons/cm?* over a period of 17,033 hr in N» + 2 to 5% O, resulted in 50 Table 19. Postirradiation Tensile Properties of ' Hastelloy N (Heat 67-504)% Test Stress, psi A True Strain Elongation, % Reduction Specimen Temper- Ultimate Fracture Number ature (E:zf 1) Tield Tensile Uniform Total in(:;l)'ea. Strain (°c) (%) 5111 25 0.05 50,300 112,200 45.9 48.1 34.63 43 5147 25 0.05 50,700 112,800 54.6 56.6 36.54 45 5110 200 0.05 41,000 96,400 43.0 44.8 38.34 48 5162 200 0.05 40,900 95,300 45.7 47.3 46.66 63 5109 400 0.05 36,700 90,800 49.8 52.0 36.15 45 5145 400 0.05 37,500 90,800 52.8 54.8 34.84 43 5108 400 0.002 36,700 91,800 52.2 54.7 45.71 61 5144 400 0.002 41,100 97,900 52.7 53.7 33.31 41 5107 500 0.05 40,600 92,700 45.7 47.2 32.23 39 5143 500 0.05 40,900 94,000 49.6 50.6 39.97 51 5106 500 0.002 41,600 90,500 44 .5 46.2 39.40 50 5142 500 0.002 33,800 178,600 41.1 42.6 36.00 45 5117 550 0.05 39,700 74,900 34.5 36.5 31.45 38 5153 550 0.05 34,700 77,800 39.7 43.5 37.94 48 5118 550 0.002 44,600 62,000 11.9 15.1 16.02 17 5154 550 0.002 33,200 59,900 20.5 21.5 20.38 23 5096 600 0.05 31,900 74,500 34.7 36.9 26.02 30 5164 600 0.05 32,900 70,800 32.9 34.0 35.69 44, 5120 600 0.002 36,400 49,600 8.9 10.9 20.38 23 5156 600 0.002 30,900 49,500 12.6 15.7 23.35 27 5128 650 0.05 39,900 55,500 - 24.0 25.8 22.06 25 5121 650 0.05 38,800 59,600 14.6 16.5 19.15 21 5155 650 0.05 30,200 46,200 12.8 15.8 10.17 11 5122 650 0.002 52,200 54,500 2.2 6.2 3.51 4 5158 650 0.002 31,700 44,800 13.2 - 14.3 11.54 2 5105 650 0.0002 32,200 40,800 7.4 9.8 13.52 15 5141 650 0.0002 30,700 38,800 - 7.8 8.8 6.65 7 5140 650 -0.000027 30,200 37,100 v b 6.5 5.1 5 5104 650 -0.000027 30,700 37,200 4.8 6.5 5.4 6 5098 760 0.05 27,000 38,600 9.0 23.7 18.94 21 5123 760 0.05 33,300 40,100 6.7 8.7 10.17 11 5159 760 0.05 28,800 38,000 9.6 13.5 13.19 14 5124 760 0.002 33,300 33,300 1.2 2.7 4.63 5 5160 760 0.002 28,200 34,600 4.2 5.6 6.00 6 5102 850 0.05 27,300 29,700 2.9 4.3 2.40 2 5138 850 0.05 27,500 30,900 3.2 4.5 7.10 7 5103 850 0.002 21,100 - 21,100 1.0 1.9 0.64 1 5139 850 0.002 23,700 23,700 1.0 2.5 0.48 0 %pnnealed 1 hr at 1177°C. Irradiated to & thermsl fluence of 0.25 x 1020 neutrons/cm® over a period of 17,033 hr at 650°C. 0 " » 60 50 40 30 FRACTURE STRAIN (%} 20 10 o Fig. 39. Fracture Strains of Heat 67-504 After Irradiation to a Thermal Fluence of 2.5 X 101° neutrons/em® for 17,033 hr at 650°C. Fig. 40. Comparison of the Fracture Strains of Heat 67-504 in 51 ORNL-DWG TO-T731t o o O o T H § > . Ve o o Q STRAIN RATE {(min"") - o 0.05 ° & 0.002 ° o 0.0002 5 \l }\\o A 0 400 200 300 400 500 600 700 800 TEST TEMPERATURE (°C) ORNL=-DWG 70-73(2 ‘80 o 80 / a1 S \ 70 . \ 60 g | | E °--.- ————— -""-—‘., 5""—‘:-{'0 o \ @ - 2 N W _ >d:——-" —————— * . £ 40 5 , - , , TN g AN \ & fra _ A / . \ \ 50 7/ o ANNEALED 1 hr AT #77°C \ : a4 & ANNEALED { hr AT H77°C, . \ : AGED 9789 hr AT 650°C \ : © ANNEALED { hr AT H77°C, \ \ 20 IRRADIATED TO 5.3 x402% neutrons/em® X OVER 9789 hr > o ANNEALED { hrATH77°C \ - , IRRADIATED TO 2.5 x40 neutrons/cm?2 ~ . OVER 17,033 hr "~ 10 - ‘\ b . N \ 0 o - 100 200 . 300 - 400 - 50O 600 700 800 900 _TEST TEMPERATURE (°C) 900 Various Conditions When Tested at a Strain Rate of 0.05 min~!. 52 fracture strains of about 50% up to 500°C, above which the fracture strain dropped progressively with increasing temperature. Irradiation to a thermal fluence of 5.3 X 102C neutrons/cm® at 650°C in fuel salt over a period of 9789 hr resulted in a low fracture strain at 25°C (which may be anomalous), fracture strains of about 45% up to 500°C, and decreasing fracture strains with increasing test temperature. The most striking feature is that above 550°C the fracture strains are lower for material irradiated in N, + 2 to 5% Oy for 17,033 hr to a fluence of 2.5 X 101° neutrons/em? than for those irradiated in fuel salt for 9789 hr to a fluence of 5.3 X 10%C neutrons/cm®. Some values of the tensile properties at 25 and 650°C are given in Table 20. The yield stress at both test temperatures was increased by aging and by irradiation. The ultimate tefisile stress at 25°C was increased by aging and by irradiation; at 650°C it was increased by aging,ybut decreased. by irradiation. We have already discussed the changes in the fracture strain. Table 20. Comparison of the Tensile Properties of Heat 67-504 After Various Treatments® Ultimate Tensile Stress, psi et 25°C at 650°C Fracture Strain, % at 25°C at 650°C Yield Stress, psi at 25°C at 650°C Heat Treatment Annealed 1 hr st 1177°C 37,400 25,600 105,000 83,000 73.2 65.9 Annealed at 1177°C and 67,500 43,300 133,000 94,300 52.3 50.9 aged 9789 hr at 650°C Annealed at 1177°C, irra- 102,000 32,900 119,300 70,000 27.0 30.7 diated toP a thermal fluence of 5.3 x 102° neutrons/cm? over 9789 hr (salt enviromment) Annealed at 1177°C, 50,300 38,800 112,000 = 59,600 48.1 16.5 irradiated to a ther- mal fluence of 2.5 % 101? neutrons/cm? over 17,033 hr (N, + 2 to 5% 0, enviromment) ®Tested at a strain rate of 0.05 min-l. bData. from: H. E. McCoy, Jr., An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimens — Third Group, ORNL-TM-2647 (1970). L} 53 The results of Creep-rupture tests on heat 67-504 are summarized in Table 21. These results and those obtained previously'! were used to prepare Figs. 41, 42, and 43. The stress-rupture propertiés in Fig. 41 show that the rupture'lifé was not changed appreciably'by aging, at least for 9789 hr at 650°C. The stress-rupture life was reduced at least two orders of megnitude by the lower fluence and only a factor of 2 by the higher fluence. The mi ndmum creep rates in Fig. 42 show that this property was not changed as much as the rupture life. Aging for 9789 hr increased the'minimfim creep, but samples irradiated over the same period had a lower creep rate than the aged samples. The samples irradiated to the lower fluénce had a creep rate an order of magnitude higher than that of the as-annealed material. The fracture strains are 11H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment Hastelloy N Surveillance Specimen — Third Group, ORNL-TM-2647 (1970). Table 21. Creep-Rupture Tests on Heat 67-504 at 650°C _ : Minimum Test Specimen Stress LifeRupturgtrain Creep Number Number (psi) () (%) Rate | (%/hr) Irradiated at 650°C to a thermal fluence of 2.5 X 1019 neutrons/ecm? R-952 ' 5112 : 55,000 0.6 - 8.6 9.88 R-953 5148 55,000 1.0 10.6 6.6 R-959 5113 . 47,000 7.6 12.1 0.83 R-965 5149 47,000 - 3.3 6.2 0.52 R-971 5114 40,000 1.5 10.9 0.42 R-1017 5150 - . 40,000 33.8 4.1 0.050 R-968 5115 - - 32,400 235.5 3.8 0.0071 R-1033 5137 . 32,400 - 268.7 4.2 0.0066 R-1019 - 5116 - 27,000 1175.4 2.7 0.0016 S . Annealed 1 hr at 1177°C 7432 6247 70,000 13 329 3.9 - 7431 6245 63,000 - 17.7 31.2 O0.16 6255 491 55,000 127.9 27.2 0.029 - 6254 . 4936 ¢ 47,000 0 425.5 28.7 0.0068 6253 4933 -~ 40,000 876.9 16.3 0.0030 54 ORNL-DWG 707315 70 o 1 '-._....- 60 S ™ I N ™ \‘lk E\ \\ 50 P M N P . e NN \- u\:\ ‘\~ \ \ = \ \\ \\ a a. . ™la N a0 ® Q / ’/’u ~ V4 R & A //fl /.4 W Z A 7 STANDARD HASTELLOY N = "7 RV UNIRRADIATED, IRRADIATED 4 < | y /‘// | - A & (™ — 10° 10! 102 10° 10° 20 ” // 7320, 20 z RUPTURE TIME (hr} yd 4 d L ' | g 7 ‘ & 2 w 10 |— g G | < i E 0 ?‘ 2 5 32 2 5 w02 2 s ot 2 5 0 2 MINIMUM CREEP RATE (%/hr) | Fig. 67. Postirradiation Creep Properties at 650°C of Several Modified Alloys That Were Included in the and heat numbers are shown by ea.c%n 1 ine (see Table 1, p. 4, for the chemical compofitien). 86 ORNL-DWG 70-3979 6 67-504, 5.3%10%° AND 67-502, 5.3x10%° 7320, 5.1x10%° 1.3x40%%, STANDARD HASTELLOY N 9.4%10%°, STANDARD HASTELLOY N 0 04 1073 1072 10! MINIMUM CREEP RATE (%/ hr) Surveillance Program. The thermal fluence 9 87 larger smounts of titanium or combined amounts of these elements along with niobium and hafnium. SUMMARY AND CONCLUSIONS The heats of Hastelloy N used in fabricating the MSRE have shown & systematic deterioration of mechanical properties with increasing neutron fluence. The material expoeed for the longest period of time in the core has reached a thermal-fluence of 1.5 x 10?1 neutrons/cm? and a fast fluence (> 50 kev) of 1.1 X 102! neutrons/cm®. These values are quite close to those anticipated fer future reactors with a 30-year design | life. The ductility of the material was too low, but the microstructure was free of irradiation-induced voids and defects other than helium bub- bles. Several heats of the modified alloys have been exposed to the MSRE and these have bhetter postirradiation properties. They also seem to have good corrosion resistance. The standard Hastelloy N removed from the core shows some evidence of corrosion. The corrosion seems generally to be due to the selective removal of chromium, as predieted by prenuclear tests.- Some observations that have not been explained adequately are (1) the presence of grain- boundary cracks in the straps that held parts of the surveillance assem- bly together, (2) the modified microstructure near the surface, and (3) the formation of intergranular'cracks originating from the surface when irradiated materials are strained. ' One of the modified alloys, heat 67-504, was ‘exposed to the cell enviromment. This heat had been previousLy exposed to the MSRE. The fiuence was. higher in the core, but the postirradiation properties were _superlor to those of the material exposed to the cell env1ronment We presently have no explanatlon for the observed behevmor. ACKNOWLEDGMENTS The author- is 1ndebted to many people for 3851st1ng in this study - W. H. Cook and A. Taboada for design of the surveillance assembly and insertion of the specimens; W. H. Cook and R. C. Steffy for measurements 88 of flux; J. R. Weir, Jr., R. E. Gehlbach, and C. E. Sessions for review of the manuscript; E. J. Lawrence and J. L. Griffith for assembling the surveillance and control specimens in the fixture; P. Haubenreich and the MSRE Operation Staff for the extreme care with which they inserted and removed the surveillance specimens; E. M. King and the Hot Cell Operation Staff for developing techniques for cutting long rods into individual - ‘specimens, determining specimen straightness, and assistance in running creep and tensile tests; B. C. Willjams, B. McNabb, and H. W. Kline for running tensile and creep tests on surveillance and control specimens; J. Feltner for processing the test data; H. R. Tinch and N.:M. Atchley for metallography of the control and surveillance specimens; Frances Scarboro of The Metals and Ceramics Division Reports Office for -preparing the menuscript; and the Graphic Arts Department for preparing the drawings. | | 1y * 89. ORNL-TM-3063 INTERNAL DISTRIBUTION 1-3. Central Research Library 67. J. R. DiStefano 4. ORNL Y-12 Technical Library 68. S. J. Ditto , Document Reference Section 69. W. P. Eatherly 5-24. Laboratory Records . 70. J. R. Engel 25. Laboratory Records, ORNL RC 71. J. I. Federer 26. ORNL Patent Office 72. D. E. Ferguson 27. G. M. Adamson, Jr. 73. J. H Frye, Jr. 28. J. L. Anderson 74. W. K. Furlong 29. R. F. Apple 75. C. H. Gabbard 30. W. E. Atkinson _ 76. R. B. Gallaher 31. C. F. Baes 77. R. E. Gehlbach 32. S. J. Ball | 78. L. 0. Gilpatrick 33. C. E. Bamberger 79. G. Goldberg 34, C. J. Barton _ 80. W. R. Grimes 35. H. F. Bauman 8l. A. G. Grindell 36. S. E. Beall | ‘ 82. R. H. Guymon 37. M. J. Bell 83. W. O. Harms 38. C. E. Bettis _ 84. P. N. 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Iundin 65. J. M. Dale 113. R. N. Lyon 66. J. H. DeVan 114. R. E. MacPherson 90 115. D. L. Manning 155. J. L. Scott 116. W. R. Martin : 156. C. E. Sessions 117. R. W. McClung 157. J. H. Shaffer 118-123. H. E. McCoy , 158. W. H. Sides 124. D. L. McElroy 159. G. M. Slaughter 125. C. K. McGlothlan 160. A. N. Smith 126. C. J. McHargue 161. F. J. Smith 127. H. A. Mclain , 162. G. P. Smith 128. B. McNabb 163. 0. L. Smith 129. L. E. McNeese ' 164. P. G. Smith 130. J. R. McWherter 165. T. Sp:.ewak 131. A. S. Meyer 166. R. C. Steffy 132. R. L. Moore 167. R. A. Strehlow 133. D. M. Moulton 168. R. W. Swindeman 13%4. T. R. Mueller | 169. J. R. Tallackson 135. H. H. Nichol 170. R. E. Thoma 136. J. P. Nichols 171. D. B. Trauger 137. E. L. Nicholson | 172. W. E. Unger 138. T. S. Noggle - 173. G. M. Watson 139. L. C. Oakes 174. J. S. Watson 140. S. M. Ohr 175. H. L. Watts 141. P. Patriarca 176. C. F. Weaver 142. A. M. Perry 177. B. H. Webster 143. T. W. Pickel 178. A. M. Weinberg 144. H. B. Piper 179. J. R. Weir 145. C. B. Pollock 180. K. W. West 146. B. E. Prince 181. M. E. Whatley 147. G. L. Ragan 182. J. C. White - 148. D. M. Richardson 183. R. P. Wichner 149. R. C. Robertson 184. L. V. Wilson 150. K. A. Romberger 185. Gale Young 151. M. W. Rosenthal ' ) 186. H. C. Young 152. H. C. Savage 187. J. P. Young 153. W. F. Schaffer 188. E. L. Youngblood 154. Dunlap Scott 189, F. C. Zapp EXTERNAL DISTRIBUTION 190. G. G. Allaria, Atomics International 191. J. G. Asquith, Atomics International 192. D. F. Cope, RDT, SSR, AEC, Oak Ridge National Laboratory 193. C. B. Deering, Bla.ck and Veatch Kansas Clty, Missouri 194. A. R. DeGrazia, AEC, Washlngton' 195. H. M. Dieckamp, Atomcs International 196. David Elias, AEC, Washington 197. A. Giambusso, AEC Washington 198, J. E. Fox, AEC, Wa.shington 199. F. D. Haines, AEC, Washington 200, C. E. Johnson, AEC, Washington y o . 201. 202. 203. 204—-205. 206, 207. 208. 209. 210. 211. 212. 213. 214, 215. 216. 217. 218. 219. 220. 221. 222. 223. - 224, 225. 226. 227 . 228-242. 91 W. L. Kitterman, AEC, Washington ' Kermit Laughon, AEC, OSR, Oak Ridge National Laboratory C. L. Matthews, AEC, OSR, Osk Ridge Nationel Laboratory T. W. McIntosh, AEC, Washington A. B. Martin, Atomics International J. M. Martin, The International Nickel Company, Huntington, W. Va. D. G. Mason, Atomics International G. W. Meyers, Atomics International D. E. Reardon, AEC, Canoga Park Area Office 'T. C. Reuther, AEC, Washington R. Riley, AEC, Washington K. Roche, Stellite Division, Cabot Corporation, 1020 W. Park ve., Kokomo, Ind. 46901 M. Roth, AEC, Oak Ridge Operations Shaw, AEC, Washington J. M. Simmons, AEC, Washington T. G. Schleiter, AEC, Washington W. L. Smalley, AEC, Washington Earl O. Smith, Black and Veatch, Post Office Box 8405, Kensas City, Missouri 64114 S. R. Stamp, AEC, Canoga Park Area Office E. E. Stansbury, The University of Tennessee D. K. Stevens, AEC, Washington R. F. Sweek, AEC, Washington A. Taboada, AEC, Washington M. J. Whitman, AEC, Washington R. F. Wilson, Atomics International Laboratory and University Division, AEC, Oak Ridge Operations Division of Technical Information Extension D. T. H. M.