CENTRAL RESCARCH LIBRARY i s ol UNION CARBIDE CORPORATION % for the U.S. ATOMIC ENERGY COMMISSION ORNL- TM- 1005 ! EFFECT OF ELEVATED TEMPERATURE IRRADIATION ON THE STRENGTH AND DUCTILITY OF THE NICKEL-BASE ALLOY, HASTELLOY N W. R. Martin J. R. Weir CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION LIBRARY LOAN COPY DO NOT TRANSFER TO ANOTHER PERSON If you wish someone else to see this document, send in name with document and the library will arrange a loan. NOTICE This document contains informotion of a preliminary noture ond was prepared primarily for internal use ot the Ook Ridge MNational Laboratory. It is subject to revision or correction and therefore does not represent a final report. 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Weir * FEBRUARY 1965 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMIGSION £ 1 et gl e ek L iit CONTENTS Abstract . & ¢« v ¢ ¢ 4 vt e e e e e e e e e . Introduction .« ¢ ¢ ¢ ¢ o o & « ¢ o o o o « o Experimental Procedures and Test Conditions Influence of Deformation Temperature . . . . . Strain Rate Sensitivity at Elevated Temperature Postirradiation Annealing . . . . . . . . . . Discussion « « + o« o ¢ o ¢« ¢« o o 4« o o o . Acknowledgments . .« . ¢ ¢ v 4 4 4 e e 4 e e Page Q0 o O I N 16 EFFECT OF EIEVATED TEMPERATURE IRRADIATION ON THE STRENGTH AND DUCTILITY OF THE NICKEL-BASE ALLOY, HASTELLOY N W. R. Martin and J. R. Weilr ABSTRACT The tensile properties of Hastelloy N have been deter- mined after irradiation at 700°C to a dose level of 7 x 1020 nvt (E > 1 Mév) and 9 x 1029 nvt (thermal). The strength and ductility of the material were determined as functions of deformation temperature for the range of 20 to 900°C. These properties were also examined as functions of strain rate within the limits of 0.002 and 0.2 in./min for deformation temperatures of 500, 600, 700, and 800°C. , The stress-strain relationship is not affected by irra- diation at 700°C. Ductility, as measured by the true uniform and fracture strains, is reduced for deformation temperatures of 500°C and above. The loss in ductility results in a reduc- tion in the true tensile strength. This loss is more signif- icant at test conditions resulting in intergranular failure, such as low strain rates at elevated temperature. Postirradi- E)Ij . . . . et ation annealing of the irradiated alloy does not result 1n ) ;fig&wa ; improved ductility. These data are compatible with the GLMNWQC experiments suggesting helium generation from the (n,a) reaction of boron as the cause of low ductility. The low ductility of irradiated alloys in general 1is described in terms of the present knowledge of intergranular fracture. Means of improving the ductility are discussed. INTRODUCTION Nickel-base alloys and stainless steels are used extensively in nuclear reactors because of their resistance to corrosion, suitable mechanical properties, and fabricability. Because of the lack of con- sistent data on the strength and ductility of material irradiated at well-defined conditions, the design of reactor components is usually based on the mechanical properties of unirradiated material with appropriate safety factors. Investigations of the mechanical properties of irradiated material are needed to establish confidence in that element of safety. o N A T R e T M e e et e s 0% e vt ™em e Postirradiation tensile tests are considered useful tools for evalua-~ 1 reduction in ductility for tion of irradiation damage. We have shown irradiated stainless steel by using the tensile test technique. Hastelloy N has been irradiated at 700°C and the pre- and postirradiation tensile data compared to determine the irradiation effect. Tensile test temperature and strain rate are two variables markedly affecting the apparent strength and ductility of an alloy and both are considered in this investigation. EXPERIMENTAL PROCEDURES AND TEST CONDITIONS To evaluate the type and extent of irradiation damage at elevated temperature, Hastelloy N was irradiated at 700°C in the B-8 lattice posi- tion of the Oak Ridge Research Reactor (ORR) to an exposure of about 7 X 10%° nvt (E > 1 Mev) and 9 x 1029 nvt (thermal). The composition of Hastelloy N is as follows: Weight Element Percent Mo 16.87 Cr 7.43 Fe 3.35 C 0.03 W 0.03 Si 0.60 Mn 0.55 vV 0.26 P 0.001 S 0.006 Al 0.010 Ti 0.01 B 0.004 Co 0.07 Cu 0.02 Ni Balance Figure 1 shows a photograph of the irradiation rig. After irradia- tion the tensile samples (Fig. 2) are removed from the chamber and tested in an Instron tensile machine. The mechanical property data generated from these tests are then compared to data generated from unirradiated W. R. Martin and J. R. Weir, "The Effect of Irradiation Temperature on the Postirradiation Stress-Strain Behavior of Stainless Steel," paper presented at the ASTM Symposium on Flow and Fracture of Metals and Alloys in Nuclear Environments, Chicago, T11., June 2126, 1964, et L g el gL material given a similar thermal history and deformed at identical condi- tions. The uniform elongation is the strain at maximum load. The true fracture strain is calculated using the formula, D - 0 g =248 =— , (1) f D T where E; = true fracture strain, DD = initial diameter, and Df = Tinal diameter. The true tensile strength is defined as the true stress at the ultimate engineering stress and is calculated as given by Dieter.? 2G. E. Dieter, Jr., Mechanical Metallurgy, p. 245, McGraw-Hill, New York, 1961. UNCL ASSIFIED PHOTO 62795 - ¥ E'T H“ i |“|J"E|n| i i Fig. 1. Photograph of In-Reactor Irradiation Rig Showing Tensile Specimens in Furnaces. Fig. 2. Testing of Specimens in Instron Tensile Machine. INFLUENCE OF DEFORMATION TEMPERATURE The data for Hastelloy N strained at a rate of 2% per min are given in Table 1 for irradiated and unirradiated material as a function of postirradiation deformation temperature. The 0.2% offset yield stress was not significantly affected by irradiation. The deviations in the true tensile strength are believed to be insignificant except for defor- mation temperatures of 500°C and above. The true tensile stress was reduced about 50% at 700°C. Ductility of the alloy, as measured by the true uniform and fracture strains, e, was affected at deformation tempera- tures of 500°C and above. The true fracture strain of the unirradiated Table 1. Tensile Strength and Ductility of Irradiated and Unirradiated Hastelloy N Deformation Stress, psi Ductility, % Temperature, °C Yield Strength True Tensile Strength True Uniform Strain True Fracture Strain Irrad. Unirrad. Irrad. Unirrad, Irrad. Unirrad. Irrad. Unirrad. x 10° X 10° x 10° x 10° Room temperature 46.3 45,5 168.6 166.5 42,3 40.6 42.5 39.0 100 43,9 43.9 159.5 161.0 40,1 40,3 b, 6 37.2 200 38.4 40,7 150.6 157.5 40,3 41,9 42.5 50.7 300 36.0 40,7 154.3 147.0 42 .2 37.9 bty 6 4l.4 400 35.0 40.7 146.9 153.0 40.2 39.3 42,5 46.9 500 35.8 35.8 129.5 144.0 35.3 42 4 600 32.5 36.2 82.4 109.0 11.8 26."7 21.9 31.6 700 31.0 34.1 53.4 102.8 8.0 30.8 11.6 42.1 800 28.5 30.9 38.4 59,9 3.7 12.2 6.9 86.6 material exhibits a minimum at elevated temperature. This minimum is typical for the alloy and is reflected in the reduction of area and total elongation measurements normally reported. However, no ductility minimum 1s observed for the irradiated alloy, and the ductility as measured by either uniform or fracture strain decreases with increasing deformation temperature. STRAIN RATE SENSITIVITY AT ELEVATED TEMPERATURE The strength and ductility of the irradiated and unirradiated alloy are given in Table 2 as functions of strain rate for deformation tempera- tures of 500°C and above. The ductility decreases with decreasing strain rate. The ductility of the irradiated material is particularly low at 800°C and 0.2% per min strain rate. The effects of irradiation on the uniform and fracture strains, shown in Fig. 3., are approximately equiva- lent in magnitude. However, the effect on uniform strains appears to saturate, whereas the magnitude of the irradiation effect on the fracture strains continues to increase with increasing deformation temperature. UNCLASSIFIED ORNL-DWG 64-4742 oL LN\ JAN TN V X 0 O Wit 0.8 515 2 & 20%/min ol iz 0.6 ; \ % Z%/min \ S 04 N 29 /mi - . o/ MIN ™ & OZ?@/%HT’H—% o _ . F//f 0.2 % /min— q 0.2 \ ] UNIFORM STRAINS FRACTURE STRAINS ] ~. O O 200 400 600 800 O 200 400 600 800 DEFORMATION TEMPERATURE (°C) Fig. 3. Effect of Irradiation on Ductility of Hastelloy N as a Function of Strain Rate and Deformation Temperature. Table 2. Strain Rate Sensitivity of Irradiated and Unirradiated Hastelloy N Deformatlion otraln Rate, otress, psi Ductility, % Temperature, °C per min Yield Strength True Tensile Strength True Uniform Strain True Fracture Strain Irrad. Unirrad. Irrad. Unirrad, Trrad. Unirrad. Trrad. Unirrad. X 10° x 10° x 10° x 103 500 0.2 32.7 34.9 136.1 145.6 41.2 42.3 46.6 53.4 500 0.02 35.8 35.8 129.5 144.0 35.3 42,4 51.4 500 0,002 34,4 37.4 122.5 131.5 32.3 33.3 36.5 34.2 €00 0.2 32.9 34,1 112.7 134.4 31.2 38.9 36.5 48,7 600 0.02 32.5 36.2 82.4 109.0 17.7 26.7 21.9 31.6 €00 0.002 34.2 34.6 63.1 106.0 10.3 29.9 13.2 29,7 700 0.2 30.5 30.9 66.8 106.5 13.5 32.2 19.2 39.2 700 0.02 31.0 34,1 53.4 102.8 8.0 30.8 11.6 42.1 700 0.002 32.1 33.7 47.0 80.5 5.6 20,0 7.8 29.0 800 0.2 29.3 29.3 45,7 79.8 6.7 11,6 800 0.02 28.5 30.9 38,4 59,9 3.7 12.2 6.9 86.6 800 0.002 29.3 32.5 32.2 42,9 1.8 6.5 4,9 93.7 POSTIRRADTATION ANNEALING The effect of postirradiation heat treatment of the alloy at the recommended solid solution temperature of 1175°C for 1 hr is shown in Table 3. The elevated temperature ductility of the irradiated alloy is not improved by the heat treatment, thereby indicating the thermal sta- bility of the configuration causing the reduced ductility. Since the heat treatment results in resolution of carbide precipitates, any influence of irradiation on the precipitation of these carbides is not responsible for the observed ductility reduction. Table 3, Effect of Postirradiation Heat Treatment of Irradiated Hastelloy N at 1175°C for 0.5 hr (Tested at a strain rate of 0.002% per min) Deformation Stress, psi Ductility, % Temperature, °C Condition 0.2% True True True Offset Yield Tensile Uniform Fracture x 103 x 102 700 Unirradiated 33.7 80.5 20.0 29.0 700 Irradiated 32.1 47,0 5.6 7.8 —~ 700 Irradiated plus 30.5 bbe 2 6.3, 6.9 postirradiation heat treatment 200 Unirradiated 21.8 2.2 1.2 70.0 900 Trradiated 23.2 23.2 < 0.4 < 0.4 900 Irradiated plus 23.5 23.5 < 0.6 < 0.7 postirradiation heat treatment DISCUSSION The results indicate that the effect of irradistion at elevated tem- perature on the strength and ductility of Hastelloy N is qualitatively the ‘same as that reported for stainless steel.?’% The stress-strain relation- ship at elevated temperature is not affected by irradiation at 600°C and above. This is in contrast to irradiation below 600°C, where the stress- 3 strain relationship 1s affected by irradiation,” as shown in Fig. 4. UNCLASSIFIED ORNL-DWG 64-1326R IRRADIATED * / UNIRRADIATED ; 3 ! x i STRESS —» ! ; | . | MATERIAL DEFORMED AND IRRADIATED — b AT LOW TEMPERATURE (7< Y47.) i ; i 1 MATERIAL DEFORMED AND IRRADIATED AT ELEVATED TEMPERATURE (7 >1/5 7)) ; UNIRRADIATED - //1§;AMATED STRESS ——#= STRAIN —— Fig. 4. Effect of Irradiation on the Stress-Strain Curves. The defects introduced by fast neutrons annealed during irradiation at the elevated temperatures. The reduction in tensile strength at elevated temperatures is a result of the inability of the irradiated alloy to strain plastically. The fracture at a reduced strain therefore decreases the true tensile strength, and the magnitude of the reduction increases as the test conditions are altered to increase the strain-hardening coefficient (i.e., increased strain rates). If the alloy is irradiated at elevated temperature, the reduction in ductility occurs only for deformation at elevated temperatures. Metallographic examination by the 3W. R. Martin and J. R. Weir, "The Effect of Irradiation Temperature on the Postirradiastion Stress-Strain Behavior of Stainless Steel,” paper presented at the ASTM Symposium on Flow and Fracture of Metals and Alloys in Nuclear Environments, Chicago, I1l., June 2126, 1964, “W. R. Martin and J. R. Weir, Nature 202, 997 (1964). 10 1light microscope shows that the deformation temperature at which the irradiated alloy becomes embrittled is associated with the transition from transgranular to the intergranular mode of fracture. When the irradiated and unirradiated specimens fracture transgranularly, no loss of ductility is observed (Fig. 5). Figure 6 shows that grain boundary R SR = < Bh 7oA.t D <= UNCLASSIFIED e A - 4 o 5 N . Y.56352 - : UNCLASSIFIED b . : GR R-19723 - o e - % : 8 P . AL et i »jtfi;xfi. _ - & Wt —_— o 3 - o G T o Wil w S i - & . i _i' % 8 : : 1 . ] R = ‘ " :E,c *‘«'}'fjfld’.fl“ - 0 1, £ s s E '.Ff :j;r !”; ;. 'T.'s" = B ':L' . _""“‘-—..-w ' ] : < Ly & o > 2 : — £ 2 . " 1 ot -fi — — =eiyp 2 4 w -, fi : = Somt w ol : Tt a7 4 (D) B3 ML T g " T % ey B = ' Sy . S Cme W Vel Fig. 5. Comparison of the Fracture of Irradiated and Unirradiated Hastelloy N at Approximately 35% Strain. Tested at a strain rate of 0.29 per min and 500°C. Etchant: aqua regia. 100x. (a) Unirradiated Hastelloy N shows no grain boundary failure at 46.6% strain and specimen rupture by a transgranular mode. (b) Irradiated alloy shows intergranular surface cracks but specimen failure by a transgranular mode. JLl UMCL ASSIFIED Y.57842 UNCLASSIFIED R-19725 Fig. 6. Comparison of Irradiated and Unirradiated Hastelloy N at 29% Strain. No fractures are observed in unirradiated sample tested at 700°C and 0.2% per min strain rate. Etchant: aqua regia. 100x. (a) Unirradiated. (b) Irradiated. cracks are found in the irradiasted material at smaller strains than in the unirradiated alloy when observed at 100X using the light microscope. These cracks propagate along the boundary rather than widen in the irradi- ated material and cause complete specimen rupture at a considerably reduced strain, as shown in Fig. 7. Therefore, it would appear that both the nucleation and propagation of cracks are affected. UNMCL ASSIFIED Y«56357 N UNCLASSIFIED . Fe R-19727 e B k {3 (b) Fig. 7. Comparison of the Grain Boundary Cracks at Fracture for Unirradiated and Irradiated Alloy Strain at 900°C and at a Strain Rate of 0.2% per min. Etchant: aqua regia. 100x. (a) Unirradiated alloy fracture at 70% strain. (h) Irradiated alloy fracture at approximately 0.4% strain. Thus, the influence of irradiation at elevated temperature is one affecting only the fracture of grain boundaries. Intergranular fracture is classified into two types: 1. wedge type that originate at triple points, 13 2. cavity type 1n which small cavities are nucleated along grain boundaries. The intergranular tensile test fractures observed for the irradiated and unirradiated alloys have been of the wedge type. These type cracks are formed on boundaries transverse, and on occasion obligue, to the direction of the stress applied to the bulk specimen. Although many questions need to be answered about intergranular wedge-type fracture, it is generally believed’ that a prerequisite is grain boundary sliding. Localized deformation along the boundaries results in stress concentrations that nucleate fracture if not dissipated by boundary migration or recrystalliza- tion. Any mechanism proposed for the reduced ductility of irradiated alloys must be one that does not increase the effectiveness of the boundaries as dislocation barriers. Grain boundary sliding could be affected by irradia- tion either reducing the magnitude of boundary deformation that an aggregate can accommodate before fracture or by increasing the rate of boundary sliding. Measurement of grain boundary deformation, perhaps by the Rachinger® method, needs to be determined as a function of stress for irradiated and unirradiated alloys. Certainly, the reduced ductility 7 governing the rate of extension of the cracks. could be related to factors Electron fractography is required to elucidate this area. The elevated temperature embrittlement has been reported by Roberts and Harries® to be related to thermal rather than fast neutrons. Slow neutrons, apart from scattering, show four types of capture reactions: 1. emission of gamma radiation (n,y), 2. eJjection of an alpha particle (n,a), 3. eJjection of a proton (n,p), and 4, Tfission (n,f). Of the four reactions, the (n,a) reactions appears the most likely to cause the embrittlement. This reaction, producing helium localized at the grain boundary, could affect the nucleation of cracks at the grain °J. R. Low, The Fracture of Metals, p. 61, MacMillan and Company, New York, 1963. ®W. A. Rachinger, J. Inst. Metals 81, 33 (1952). 7D, McLean, Grain Boundaries in Metals, p. 335, Clarendon Press, - Oxford, 1957. 84, C. Roberts and D. R. Harries, Nature 200, 773 (1963). 14 boundary and, conceivably, crack propagation. The thermal stability of the defect, as indicated by the postirradiation heat treatment data, is compatible with the hypothesis of helium generation. The thermal stabil- ity of this defect ig in sharp contrast with data by Ba.iley9 and others!® that show that the low-temperature (< 500°C) damage caused by fast neutrons can be annealed at 400 to 980°C. However, the concentration of the species necessary for (n,a) reaction is in the parts per million range for stainless steel and nickel-base alloys. Boron-10 and lithium-6 are two such species, given by Barnes,11 that generate large specific volumes of gas. If the boron is concentrated at the grain boundaries, the atomic fraction of the helium produced can be large even with the average concen- tration of 1°B in the 1- to 10-ppm range. Studies by other investigatorsl2 indicate boron segregation in the grain boundaries of austenitic alloys. et The grain boundary concentration of helium necessary to cause embrittlement 13 is unknown although it can be estimated from the beryllium irradiations to be < 0.7 X 10=%, Studies'*»1° to examine the influence of boron content on the properties of irradiated alloys show complex relationships between boron content and ductility for the alloy deformed at elevated tempera- ture. However, these studies are complicated by the fact that boron for the concentrations investigated influences the strength and ductility of R. E. Bailey and M. A, Silliman, Symposium on Radiation Effects of Materials, vol. 3, ASTM Special Technical Publication 223, 1958, 1%, E. Murr and F. R. Shober, Annealing Studies on Irradiated Type 347 A o g o T s e Stainless Steel, BMI-1621 (March 1963). 11R. s. Barnes and G. W. Greenwood, Proc. U, N, Intern, Conf. Peaceful Uses At. Energy, 2nd, Geneva, 1958 5, 481 (1958). 12¢, Crussard, J. Plateau, and G. Henry, Proceedings of the Joint International Conference on Creep, pp. 1-91, vol. 1, Institution of Mechanical Engineers, London, 1963. 13J. R. Weir, Proceedings of the International Conference on the Metallurgy of Beryllium, London, 1961, pp. 395409, Chapman & Hall, London, 1963, 14y, Hinkle, W. E. Brundage, and J. C. Zukas, Solid State Div. Ann. Progr. Rept. Aug. 31, 1960, ORNL-3017, p. 120. 15High-Temperature Materials Program Progr. Rept., Jan. 24, 1964, GEMP No. 31, Part A, p. 30. B =y 15 the unirradiated alloy. The influence of boron on the solubility of carbon, alteration of precipitate distribution, and grain size is well documented.1%,16,17 Of course, the foregoing discussion is dependent on the embrittle- ment being related to thermal neutrons and helium generation. The data generated, to date, 1s indirect evidence. However, if the ductility is a result of helium generation, the problem of embrittlement of irradiated materials deformed at elevated temperature will have to be solved by removing the species undergoing (n,a) reaction from the grain boundaries. Several approaches are available: 1. The average concentration may be lowered by electron-beam melting. 2. Distribute the boron within the grain as stable precipitates, such as borides and combination precipitation of Fe,3(BC)g¢. Helium atoms and/or bubbles may be trapped at the precipitate, thereby greatly reducing the concentration of helium bubbles in the grain boundary. 3. Lower the concentration at the boundary by increasing the number of boundaries on which the specie precipitates or segregates. The characteristics of the damge caused by irradiation at elevated temperature are as follows: 1. Yield stress and tensile strength are not affected. 2. Ductility of material is reduced. 3. The reduction in ductility is noted by large decreases in the uniform and fracture strain and by small decreases in the true tensile and fracture stresses. 4. The reduction in ductility is more significant at test conditions resulting in intergranular failure, such as low strain rates at elevated temperature. 5. Heat treatments that anneal the damage causing the low-temperature irradiation effect do not improve the ductility at elevated temperature. 16y, V. Levitin, Phys. Metals Metallog. 11, 67 (1961). 17R. F. Decker, J. P. Rowe, and J. W. Freeman, Boron and Zirconium from Crucible Refractories in a Complex Heat Resistant Alloy, NACA-1392 (1958). 16 ACKNOWLEDGMENTS The authors wish to acknowledge the following from the Metals and Ceramics Division of the Oak Ridge National Laboratory for their work on this study: J. W. Woods, V. R. Bullington, and D. G. Gates for the design and installation of the irradiation experiments; J. C. Zukas for the flux determinations; K. W. Boling for the mechanical property tests; and E. H. Lee and Z. R. McNutt for the metallography. o — - ¥ \-‘ -, g g, 17 INTERNAL DISTRIBUTION ORNL-TM~1005 1—-2. Central Research Library 2628. M. R. Hill 3. Reactor Division Library 29. C. F. Leitten, Jr. 4~5, ORNL — Y-~12 Technical Library 30. H. G. MacPherson Document Reference Section 31. H. E. McCoy, Jr. 6—15. Laboratory Records Department 32~%41. W. R. Martin 16. Laboratory Records, ORNL RC 42. E, C. Miller 17. ORNL Patent Office 43, P. Patriarca 18. R. G. Berggren 44, G. M. Slaughter 19. G. 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