ER t DAK RIDGE NATIONAL LABORATI ?l LIBRARIES ; ’I‘NinPL nffl\r H|||||||I||||||I!||||‘|| I \DOCUNENT coutgery 3 445k 0549970 b Nt wran mawws vsATIONAL LABORATORY — i operated by ;! UNION CARBIDE CORPORATION % NUCLEAR DIVISION B n for the b U.S. ATOMIC ENERGY COMMISSION H-I ORNL- TM- 1544 CHA g g " ‘# q,l K S ” GRS i SOLUTIONS TO THE PROBLEMS OF HIGH-TEMPERATURE P IRRADIATION EMBRITTLEMENT S R i W.R. Martin J.R. Weir il P I.- F k£ 4 5 b EHERE i CENTRAL RESEARCH LIBRARY §oe DOCUMENT COLLECTION e LIBRARY LOAN COPY 1} et DO NOT TRANSFER TO ANOTHER PERSON T HEIE X If you wish someeone else to see this SRR : i 3 document, send in name with document | s and the library will arrange a loan. P NOTICE This document contains information of o preliminary nature i ond wos prepared primarily for internal use at the Oak Ridge National e Laberatory. It is subject to revision or correction and therefore does ‘Q' i nat represent o finol report. —— LEGAL WOTIKE—7M7m7F ——FF This report wos prepared os an account of Government sponsored work. MNeither the United States, nor the Cammission, ner ony persen acting on behalf of the Commission: A. Mokes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the informotion contoined in this reporl, or thot the use of any information, apparatus, method, or process disclosed in this repert mey not infringe privately owned rights; or B. Assumes any liabilities with respsct to the use of, or for damages resulting from the use of any informotion, opporatus, method, or process disclosed in this report, As used in the above, ''person acting on behelf of the Commission' includes any employee or contracter of the Commission, or employee of such contractor, to the extent that such employee of controctor of the Commission, or employee of such controcter prepores, disseminates, or provides access to, any information pursuont te his employment or controct with the Commission, ot his employment with such contrector. oo ORNL-TM-1544 Contract No. W-7405-eng-26 METALS AND CERAMICS DIVISION SOLUTIONS TCO THE PROBLEMS OF HIGH-TEMPERATURE IRRADIATION EMBRITTLEMENT W. R. Martin J. R. Weir Paper presented at the Sixty-Ninth Annual Meeting of the American Soclety for Testing and Materials, Atlantic City, N. J., June 27=July 1, 1966. JUNE 1966 OAK RIDGE NATTONATL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMLISSION QAK RIDGE NATIONAL LABORATORY LIBRARIES 3 4456 0549970 b Lé Abstract The effect of irradiation on the high-temperature mechanical prop- erties of structural materials is described using type 304 stainless steel as an example. The general effect is one in which the grain-boundary fracture process, but not the deformation process, 1s affected. The data suggest the primary cause to be helium generated from (n,a) reactions. - oeveral metallurgical techniques for improving the ductilities of irradi- ated alloys are suggested and experimental data on type 304 stainless steel are given for which the degree of improvement is demonstrated. Introduction Irradiation embrittlement of iron- and nickel-base alloys at tempera- tures above 500°C is a problem of immense importance to the success of nuclear reactors. The integrity of structural components and fuel element cladding can depend upon properties that are affected or related to the alloy ductility. Because of the paucity of material data at conditions ’ appropriate to the high-temperature reactor environment, a large effort has been centered in recent years in the area of irradiation damage. High-temperature embrittlement of iron- and nickel-base alloys is characteristically different than that observed at temperatures below 500°C. Within the lower temperature range, the damage produced by the displacement of atoms hardens the lattice, lowers the work hardening rate, but does not significantly affect the fracture stress. Material may exhibit low ductility in terms of uniform strains and total elongations without a large change in the true fracture strain. i B S R g 6 A B E > t; now Oy = Ga(L/Zt)l/2 . (3) Since the length of the boundary is proportional to the grain size, 1t follows that grain size can greatly influence intergranular fracture. When the grain size is decreased, a higher stress 1s required to nucleate the wedge-type fracture and also the rate at which cracks propagate should be reduced. Grain-boundary carbides could also act in the same manner. The length of the sliding interface would be the interparticle spacing if the strength of the carbide-matrix interface is large compared to the cohesive strength of the boundary. Weaver (20,21) and Garofalo (22) have shown that the heat treatment of nickel- and iron-base alloys to produce coherent grain-boundary precipitate can improve the ductility. Garofalo (22) cnumerated the following conditions for grain-boundary precipitate to be beneficial: 1. high cohesion between particle and matrix, 2. interparticle spacing of 1 to 2 p to allow grain-boundary migration, and 3. rounded particles that have high shear strength. Examples of the beneficial effect of grain size on tensile properties are given in Table 1 and the creep data at 650°C in Table 2. A 100-hr heat treatment at 800°C following a l-hr sclution anneal at 1036°C was Table 1 -- Grain Size Dependence of Postirradiation Short-Time Tensile Strength and Ductility of Type 304 Stainless Steel Deformation Yield Strength (psi) Ductility (%) Temperature True Uniform Total Elongation ASTM 9 ASTM 5 (°c) ASTM 9 ASTM 5 ASTM 2 ASTM 5 500 23.5 x 10° 18.8 x 10°® 23.6 24,7 32.2 31.8 600 19.3 13.1 25.5 24,1 34,7 32.0 700 16.7 12.4 19.5 19.1 36.3 25.2 800 15.6 11.6 15.2 14.9 29.6 16.2 900 8.2 8.7 10.1 6.2 24,7 10.2 Table 2 -- Grain Size Dependence of Postirradiation Stress Rupture of Type 304 Stainless Steel at 650°C Strength in Terms Ductility (%) Stress of Time to Rupture (psi) (hr) Elongation at Rupture ASTM © ASTM 5 ASTM 9 ASTM 5 30 x 10° 11.3 1.5 4ti .0 11.4 25 79.0 5.5 29.5 9.0 20 191.0 109.5 17.2 3.8 15 514.4 194.4 9.2 3.5 given type 304 stainless steel in order to produce the desired grain- boundary carbide distribution. These microstructures are shown in Fig. 1. The effect of aging is illustrated in Tables 3 and 4 for tensile and creep conditions, respectively. In terms of ductility, the fine-grain size 1is superior to the coarse grain in the aged and unaged conditions. However, because the coarse-grain material creeps at a rate lower than the fine- grain material, the time to rupture for the aged coarse-grain material offers the best properties for those conditions investigated to date. Another approach to improve the ductility is to assume that the radiation- induced grain-boundary embrittlement is due to the generation of helium from the (n,a) reactions. Helium bubbles at the grain boundary would be expected to be deleterious. However, helium generated within the grains would not be harmful. The primary way that these helium atoms could be swept into the boundary would be by a dislocation mechanism as illustrated v el o e et R el TR TR g T R AT T T TR S R T A e S e S IR T T ¢ Y R-29127 ; 7. o " - ; ,’ . d : .. B \‘,» "‘. o’ { ? % . \ 5 ‘;‘I - " ;" e ,3~M'. e t" p— < - R S m— - 4 1 % ot $ - \_.... 4 % Y (b) Fig. 1 -- Microstructures of Irradiated Stainless Steel Having Been Creep Tested at 630°C. 20,000 psi stress. (a) Preirradiation heat treatment of 1 hr at 1036°C, fractured at 3.8% strain. (b) Preirradiation heat treatment of 1 hr at 1036°C followed by a 100-hr aging at 800°C, fractured at 14.3% strain. 1000x. 11 Table 3 -- Effect of Preirradiation Aging on the Short-Time Tensile Properties of Type 304 Stainless Steel Ductility (%) Deformation Strain Temperature Rate Yield Strength True Total (°c) (min-1) (psi) Uniform Elongation Unaged Aged Unaged Aged Unaged Aged 704 20 12.7 x 10° 12.2 x 10° 24.3 25.8 34.7 37.2 0.2 13.1 15.6 15.6 18.1 20.5 30.5 842 20 .11.8 10.6 10.4 12.8 13.7 17.5 0.2 10.9 10.2 5.1 9.4 7.6 13.9 Table 4 ~- Effect of Preirradiation Aging on the Postirradiation Stress Rupture of Type 304 Stainless Steel at 650°C Strength in Terms Ductility (%) Stress of Time to Rupture Elongation at Rupture (psi) (nr) Unaged Aged Unaged Aged 30 x 107 1.5 14.8 11.4 24,2 25 5.5 50.8 9.0 25.1 20 109.5 664.1 3.8 14.3 15 194.4 3638.0 3.5 7.8 12 by Barnes (13) for copper. Therefore, to improve the ductility of irradi- ated materials, one must devise ways to reduce the helium concentration at the grain boundaries. For many reactor applications, the preponderance of helium generated is due to the transmutation of 19B. Boron, a horophillic element, normally segregates to the grain boundaries in the solid state and therefore a large quantity of helium is generated near these boundaries. If one could form a stable boron compound, insoluble either in the melt or at a very high temperature after solidification, it would be possible to get a homogeneous distributicn of this compound. Therefore, helium generated would tend to stay at the precipitate-matrix interface and hence the quantity at the grain boundaries would be greatly reduced. These precipi- tates having an incoherent interface, would also serve as traps for helium generated from other elements and fast neutrons. Thus in principie, this system should result in material with a lower susceptibility to high- temperature embrittlement in thermal and fast-neutron envircnments. We have chosen to first investigate the iron-base systems, In particu- lar 18-8 stainless steel. Among the most stable borides are those of titanium. We have now accumulated data from two different irradiations, and typical data are given in Fig. 2 for a steel containing 0.02 wt % C. Data for the 0.06 wt % C alloy are given in Table 5. It is clear that small additions of titanium greatly improve the ductility of type 304 stainless steels. Titanium additions at the level required to meet the chemical specificatlons for type 321 stainless steel are above the range for which one observes the maximum ductility. 13 ORNL-DWG 66— 239 STRAIN RATE OF 2%/ min 70 A 1 * UNIRRAm/.E\TED 60 IRRADIATED 4X102° neutrons/cm? (THERMAL) | \ 1.5X 1019 neutrons/cm? (FAST) \ o) | ol \ A ~_ DUCTILITY, TOTAL ELONGATION (%) — ¢ 30 20 ¢ )\C’\fi 10 | REGULAR 304 S5 ) . 0 0.2 0.4 0.6 08 1.0 1.2 14 PERCENT Ti Fig. 2 -- Ductility at 842°C of Irradiated Austenitic Stainless Steel as a Function of Titanium Content. 14 Table 5 -- Influence of Titanium on the High-~ Temperature Irradiation Embrittlement of 18-8 Stainless Steels Having 0.06 wt % C Titanium Total Elongation (%)a,b at (wt %) 650°C 700°C 842°C 0.0 31.0 31.5 20.0 0.2 40.0 38.0 45.1 0.3 34.7 28,5 35.2 0.4 31.5 26.5 25.0 0.5 28.8 21.5 19.0 0.6 24,1 19.5 17.9 0.8 22.9 20.5 19.1 1.0 30.6 23.5 23.5 1.2 29.8 23.0 28.5 aMeasured in 1-in.-gage length for tests at a strain rate of 0.2%/min. bSpecimenS irradiated to a fluence level of 1 x 10?0O neutrons/cm? (thermal) and 1.5 x 10'? neutrons/cm?® (E > 1 Mev). 15 Titanium additions in the range up to 0.2 wt % greatly reduce the magnitude of irradiation embrittlement in type 304 stainless steel. The lower ductilities of the higher titanium-bearing alloys are not understood. Although the alloys containing titanium have a grain size smaller than the unstabilized grade, we believe the effect of titanium to be as hypothesized earlier. The helium bubbles in the as-irradiated stainless steels are not always of a size resolvable in the electron microscope. A 1-hr postirradiation anneal at 1200°C will produce bubbles in the grain boundaries of the reguiar 304 stainless steel but not in the 0.2 wt % Ti- bearing steel. These photomicrographs are compared in Fig. 3. On the other hand, one can find evidence suggesting bubble attachment to intra- granular precipitate in the titanium-bearing steels. Figure 4 illustrates the possible bubble attachment. It is possible that these void areas may be a result of sample preparation for electron microscopy examination. We believe the concept of intragranular precipitate serving as sinks for bubbles to be valid. We have in our own studies experienced difficulty in getting the correct precipitate distribution and size. Additionally producing the proper distribution of boron in these precipitates may prove too difficult in many alloy systems. Thus the improvement of alloys for use in thermal reactors may prove to be more difficult than the use of this concept for fast reactor irradiations. In the latter case the pre- cipitate need not contain the boron. The precipitate may also be formed in situ, such as the titanium precipitate, or an inert oxide could be added during fabrication. This latter approach may be desirable if the thermal stability of precipitates during long exposures at high temperatures is small in the base alloy system. 16 Standard Type 304 Stainless Steel ORNL Type 304 Stainless Steel Modified with Titanium " Grain boundary -~ Fig. 3 -- Comparison of Grain Boundaries in Irradiated Stainless Steels & After Postirradiation Annealing Treatments. 17 - YE-9171 Fig. 4 -- Electron Transmission Micrograph of 0.2 wt % Ti-Bearing Stainless Steel after Postirradiation Anneal of 1 hr at 1200°C. 50,000x. 18 The final proposal for reducing the embrittlement concerns the con- centration of helium produced. In fast reactor irradiations, there appears little hope, as there is no single (n,d) reaction that produces the bulk of the helium as is the case in thermal reactors with the 1%B(n,x) reaction. Nitrogen could be a major contributor if the concentrations become larger than those presently in our commercial alloys. In the thermal reactors, one can reduce the boron content to concentrations of 10'8, but this appears impractical for commercial application. There is . another lower limit for boron because even in thermal reactors, the fast (n,a) reaction produces helium at a rate comparable with the thermal neutrons and boron at a concentration of about 0.2 ppm for neutron fluences less than 10°° neutrons/cm?. The boron level in type 304 stainless steel given an electron-beam remelt treatment was reduced from 2.9 to 0.015 ppm. Data at a neutron fluence of 4.5 x 1020 neutrons/cm? have been published earlier (23) for these alloys and others. We have investigated the embrittlement of these steels at lower neutron fluences in order to evaluate the relative con- tribution of helium generated from fast and thermal (n,a) reactions. These data are given in Table 6 and Fig. 5. The atom fraction of helium plotted in Fig. 5 is the total helium content. It is apparent that with the lower boron levels, fast (n,a) reactions are of significant importance. A correlation of this type 1s surprising since the boron, and hence the helium produced therefrom, is believed to be segregated to grain boundaries whereas the helium from fast (n,a) reactions will be produced throughout the material. If this were true, one would not expect a correlation from a simple addition of helium produced by both reactions. We believe the - Table 6 -- Ductility of Stainless Steel as a Function of Boron Concentration and Irradiation Fluence At 8.6 x 1017 neutrons/cn? At 7.8 x 1018 neutrons/cm? At 2.7 x 10'? neutrons/cm® (thermal) and (thermal) and (thermal) and Natural Boron o o . 1016 neutrons/cn® (E > 1 Mev) 7.8 x 1017 neutrons/em? (E > 1 Mev) 2.7 x 109 neutrons/cm® (E > 1 Mev) Concentration Helium Content Helium Content Helium Content (ppm) Elongation Elongation Elongation (atom fraction) (atom fraction) (atom fraction) (%) (%) (%) Thermal Total Thermal Total Thermal Total 0.015 4. 6 x 10711 1.4 x 10710 32 6 x 10719 1.4 x 10°° 27 2 x107? 5 x 10°? 0.110 43 4 x 10710 5 x 10710 31 3 x 10-2 4 x 1079 26 1 x 10"8 1 x 1078 0.150 ... 5 x 10710 6 x 10710 31 5 x 1077 6 x 107° 24 2 x 1078 2 x 1078 3.900 23 2 x 10"8 2 x 1078 18 2 x 1077 2 x 1077 18 6 x 1077 6 x 1077 6T BORON {ppb) A 15 C 110 g A 150 ® 33500 > = _ E TESTED [N VACUU =] 0O STRAIN RATE OF 0.2% PER min 10”10 1072 10~8 1077 10—® LOG OF ATCM FRACTION OF HELIUM Fig. 5 -- Irradiation Embrittlement of Boron-Stainless Steels 20 ORNL-DWG 66-405R 700°C ag a Function of Total Helium Content. e st o e b EBAe b b MR b Es e B 3 e e e R at 21 most plausible explanation is that most of the helium generated from the threshold reactions is swept into the grain boundaries during deformation; thus the majority of helium within the samples is at grain boundaries and one can then get a reasonable correlation by simple addition of the helium from both sources. This means that the actual concentration at grain boundaries for a given ductility is many orders of magnitude greater than that given in Fig. 5. A reasonable figure based on segregation may be a factor of 1000. Although replication techniques using Faxfilm are prone to exhibit artifacts we attempted fractographic evaluation of these low boron bearing alloys in order to evaluate the bubble density along the fractured boundary. Typical photographs are given in Figs. 6 and 7. We observed protrusions on the replica that cast shadows, and this i1s what we would expect for cavities on the metal fracture surface. The density of these protrusions appears to be related to the total helium content in the sample and not to neutron fluence or boron content. Assuming, therefore, that these are cavities along the grain boundary and not artifacts, it is not clear as to their role in the fracture process. Summary The irradiation of stainless steels at temperatures in the range of 600°C and above results in an embrittlement of the alloy that is signi- ficantly different than the neutron displacement damage which is of principal importance at temperatures below 600°C. This embrittlement at high temperatures does not necessarily affect the strength of an alloy, in terms of the stress necessary to produce a given value of strain. The 22 Photo J-349 Photo J-352 (b) Fig. 6. Fractographs of Irradiated Stainless Steels Containing 0.015 ppm B after Fracture at 842°C. (a) Fracture at 18% strain with calculated helium atom fraction of 5 x 10711, (b) Fracture at 12% strain with calculated helium atom fraction of 4 x 1010, 6, 500x. 23 Photo J-318 Fig. 7. Fractographs of Irradiated Stainless Steels Containing 3.9 ppm B after Fracture at 842°C. (a) Fracture at ll% strain with a calculated helium atom fraction of 2 x 10-?. (b) Fracture at 10% strain with a calculated helium atom fraction of 2 x 1078, 6, 500x. 24 embrittlement can be severe, and ductilities less than 1% have been observed for creep conditions. Irradiation affects the ability of the alloy to resist intergranular fracture and most experiments point to the principal cause as one related to the production of helium from two sources. The first one is the 10B(n,o) reaction with neutrons having thermal energies and the second source 1s from reactions between the major alloy constituents and fast neutrons having energies in the 3-Mev range. Solutions to the problem of embrittlement are then related to (1) alterations of the unirradiated alloy that affect the process of inter- granular fracture and (2) modifications of the alloy that reduce the amount of helium that is located in the grain boundaries of the irradiated alloy. Metallurgical variables, such as grain size, annealed vs cold-work structures, are variables that greatly affect the dvctilities of alloys in the temperature range for which the alloys fracture intergranularly. After irradiatlion a fine-grain size alloy can be an order of magnitude more ductile than the coarse grain. Grain sizes in the range of ASTM & to 11 are preferred. To reduce the amount of helium at the grain boundary, one must produce the helium at sites other than the grain boundary and prevent the movement of helium to the grain boundary. A reduction in boron content can reduce the amount of helium produced, but is not a complete solution because of the helium generated from nickel, iron, chromium, nitrogen, and other elements. Thus, a more suitable approach would appear to be a desegregation of boron and the production of helium sinks within the matrix of the grains 25 which will greatly reduce the quantity of helium moving to the grain boundary. Titanium additions to stainless steel are believed to form complex metal borides dispersed homogeneously within the matrix and the precipitate-matrix interfaces serve as a depository for helium. Other approaches for improving high-temperature ductility are available; one of these is proper aging of existing grades of stainless steels. Acknowledgments The authors thank their colleagues, E. E. Bloom, J. W. Woods, J. 0. Stiegler, T. E. Nolan and R. E. McDonald for their assistance in our irradiation damage program. We also thank V. Bullington, D. Gates, H. Kline, and K. W. Boling for their kind attention to details in conducting the irradiation and deformation experiments. 26 References 1. N. E. Hinkle, "Effect of Neutron Bombardment on Stress-Rupture Properties of Some Structural Alloys," Radiation Effects on Metals and Neutron Dosimetry, ASTM STP 341, Am. Soc. Testing Mater., 1963, p. 344. 2. F. C. Robertshaw, J. Moteff, F. D. Kingsbury, and M. A. Pugacz, "Neutron Trradiation Effects in A286, Hastelloy X and René 41 Alloys," Radiation Effects on Metals and Neutron Dosimetry, ASTM STP 341, Am. Soc. Testing Mater., 1963, p. 372, 3. N. A. Hughes and J. Coley, "The Effects of Neutron Irradiation at Elevated Temperatures on the Tensile Properties of Some Austenitic Stainless Steels,” Journal of Nuclear Materials, Vol. 10, 1963, p. 60. 4. P.C.L. Pfeil and D. R. Harries, "Effect of Trradiation in Austenitic Steels and Other High-Temperature Alloys," Flow and Fracture of Metals and Alloys in Nuclear Environments, ASTM STP 380, Am. Soc. Testing Mater., 1965, p. 202. 5. W. R. Martin and J. R. Weir, "Irradiation Effects in Stainless Steels at High Temperatures,”" Proceedings of Sodium Components Development Program, USAEC CONF-650620, U.S. Atomic Energy Commission, 1966, pp. 36—46. 6. P.R.B. Higgins and A. C. Roberts, "Reduction in Ductility of Austenitic Stainless Steel after Irradiation,” Nature, Vol. 206, No. 4990, 1965, pp. 12491250, 7. W. R. Martin and J. R. Weir, "Effect of Post-irradiation Heat Treatment on the Elevated Temperature Embrittlement of Irradiated Stainless Steel," Nature, Vol. 202, No. 4936, June 1964, p. 997. 27 8. J. B. Rich, G. P. Walters, and R. S. Barnes, "The Mechanical Properties of Some Highly Irradiated Beryllium," Journal of Nuclear Materials, Vol. 4, 1961, p. 287. 9. J. R. Weir, "The Effect of High-Temperature Reactor Irradiation on Some Physical Properties of Beryllium," International Conference on the Metallurgy of Beryllium, London, 1961, Chapman and Hall, London, 1963, pp. 395—409. 10. D. Kramer, W. V. Johnston, and C. G. Rhodes, "Reduction of Fission-Product Swelling in Uranium Alloys by Means of Finely Dispersed Phases," Journal of the Institute of Metals, Vol. 93, No. 5, 1964—65, Pp. 145152, 11. W. H. Chatwin and E. D. 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Stroh, "The Formation of Cracks as a Result of Plastic Flow," Proceedings of the Royal Society (London) Series A: Vol. 223, 1954, p. 404. 18. C. Zener, Elasticity and Anelasticity of Metals, University of Chicago Press, Chicago, 1948, p. 158. 19, C. E. Inglis, Institute of Naval Architects (London) Transactions, Vol. 55, 1944, p. 1Z6. 20. C. W. Weaver, "The Influence of Annealing Twins on Intergranular Creep Cracking," Journal of the Institute of Metals, Vol. 88, 1959-60, p. 296. 21. C. W. Weaver, "Application of Stroh's theory to intercrystalline creep cracking," Acta Metallurgia, Vol. 8§, 1960, p. 343. 22. TF. Garofalo, Fundamentals of Creep and Creep-Rupture in Metals, Macmillan Company, New York, 1965. 23. W. R. Martin, J. R. Weir, R. E. McDonald, and J. C. Franklin, “ "Irradiation Embrittlement of Low-Boron Type 304 Stainless Steel,” Nature, Vol. 208, No. 5005, 1965, pp. 7374. 615, 16. 17. 18. 19. 20. 21. 22. 23. 24, 25, 71. 2. 73. 7475, 76. 77, 78. 79-80. 81. 82. 83. 84. 85-87, 88. 89. 20. 91-105. ot — 29 ORNL-TM-1544 INTERNAL DISTRIBUTION Central Research Library 26-28. M. R. Hill Reactor Division Library 29. C. F. Leitten, Jr. ORNL — Y-12 Technical Library 30. A. P. Litman Document Reference Section 31. H. G. MacPherson Laboratory Records Department. 32. H. E. McCoy, Jr. Laboratory Records, ORNL R.C. 3342. W. R. Martin ORNL Patent Office 43. E. C. Miller R. G. Berggren 44, P. Patriarca G. E. Boyd 45. G. M. Slaughter R. B. Briggs 46, D. B. Trauger J. E. Cunningham 47. J. T. Venard W. W. Davis 48. J. R. Weir D. A. Douglas, Jr. 49. M. S. Weschler J. H Frye, Jr. 50-69. G. D. Whitman W. O. Harms 70. J. W. Woods NG e gqgHOg g0 EXTERNAL DISTRIBUTION M. Adams, Massachusetts Institute of Technology Brunhouse, Aerojet, General Nucleonics B. 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