Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys

Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys Comprehensive nuclear materials 1 04 effect of radiation on strength and ductility of metals and alloys

1.04 Effect of Radiation on Strength and Ductility of Metals and Alloys M L Grossbeck University of Tennessee, Knoxville, TN, USA ß 2012 Elsevier Ltd All rights reserved 1.04.1 1.04.2 1.04.3 1.04.4 1.04.5 1.04.6 1.04.7 1.04.7.1 1.04.7.2 1.04.7.3 1.04.8 1.04.8.1 1.04.8.2 1.04.9 1.04.10 References Introduction Mechanisms of Irradiation Hardening Tensile Behavior Effects of Neutron Spectrum Tensile Ductility Effect of Test Temperature Ferritic–Martensitic Alloys Introduction Tensile Behavior Helium Effects Refractory Metals Tensile Behavior Helium Effects Amorphous Metals Conclusions Abbreviations A1 ANN appm ASTM ATR bcc BR2 CW DBTT dpa EBR-II fcc FFTF HFBR HFIR HFR JPCA Lowest equilibrium temperature at which the austenite phase exists in steel Annealed Atomic parts per million ASTM International Advanced Test Reactor, Idaho Falls, ID, USA Body-centered cubic Belgian Reactor-2, Mol, Belgium Cold worked Ductile-brittle transition temperature Displacements per atom Experimental Breeder Reactor-II, Idaho Falls, ID, USA Face-centered cubic Fast Flux Test Facility, Richland, WA, USA High Flux Beam Reactor, Brookhaven, Upton, NY, USA High Flux Isotope Reactor, Oak Ridge, TN, USA High Flux Reactor, Petten, The Netherlands Japanese Prime Candidate Alloy 99 100 101 102 103 107 108 108 108 113 116 116 118 119 120 121 LMFBR Liquid Metal Fast Breeder Reactor LWR Light Water Reactor ORR Oak Ridge Research Reactor, Oak Ridge, TN, USA PCA Prime candidate alloy, adopted by the US Fusion Program in mid-1970s ppm Parts per million Unirr Unirradiated 1.04.1 Introduction The most commonly considered mechanical properties of metals and alloys include strength, ductility, fatigue, fatigue crack growth, thermal and irradiation creep, and fracture toughness All these properties are important in the design of a structure that is to experience an irradiation environment While determining the mechanical properties of irradiated materials, tensile properties, typically yield strength, ultimate tensile strength, uniform elongation, total elongation, and reduction of area are the most commonly considered because they are usually the simplest and the least costly to measure In addition, the tensile properties can be used as an indicator of the other mechanical 99 100 Effect of Radiation on Strength and Ductility of Metals and Alloys properties Space in a reactor or in an accelerator target is often so limited that the larger specimens required for fatigue and fracture toughness testing are not practical; consequently, the number of specimens that can be irradiated is so small that a meaningful test matrix is not possible Shear punch testing of 3-mm diameter disks, typically used as transmission microscopy specimens, was developed to address the problem of irradiation space Although much information can be obtained from shear punch testing, the tensile test remains the most reliable indicator of strength and ductility For these reasons, the tensile test is usually the first mechanical test used in determining the irradiated properties of new materials This chapter addresses the tensile strength and ductility of alloys 1.04.2 Mechanisms of Irradiation Hardening Irradiation introduces obstacles to dislocation motion, which results in plastic deformation, in the form of defects resulting from atomic displacement and from transmutation products Small Frank loops and defect clusters, known as black dots, large Frank loops (about an order of magnitude larger), precipitates, and cavities (either voids or bubbles) contribute to hardening in an irradiated alloy Frank loops unfault and eventually contribute to the network dislocation density Precipitates are certainly present in the unirradiated alloy, but additional precipitation results from the segregation of elements during irradiation and from the irradiation-induced changes that shift the thermodynamic stability of phases Transmutation production of new elements in the alloy can also result in the formation of new precipitates The production of insoluble species, most importantly helium, also results in precipitation, especially in the form of bubbles Defects are divided into two classes: long range and short range Short-range obstacles are defined as those that influence moving dislocations only on the same slip plane as opposed to long-range obstacles, which impede dislocation motion on slip planes not containing the obstacle.1 Coherent precipitates and large loops are long-range obstacles, but for this analysis, only network dislocations will be considered as long-range obstacles, a reasonable simplification from observations As recommended by Bement,2 the contributions from short-range obstacles are added directly, DFTS ẳ DFLR ỵ DFSR ½1Š where the quantities in eqn [1] are total stress, longrange contribution to stress, and short-range contribution to stress The contributions from the short-range obstacles are added in quadrature as follows3: DFSR ị2 ẳ DFSMloop ị2 ỵ DFLGLoop ị2 ỵ DFPRECIP ị2 ỵ DFCAVITY ị2 ẵ2 where the term on the left represents the contribution from all short-range obstacles, and the terms on the right represent the stress contributions from small loops, large loops, precipitates, and cavities, either voids or bubbles The contribution to hardening by network dislocations may be expressed by p ẵ3 tnet ẳ aGb rd where tnet is the increment in shear stress, G is the shear modulus, b is the Burgers vector, and rd is the dislocation density The constant a is dependent upon the geometry of the dislocation configuration and is usually determined experimentally However, Taylor has calculated a to be between 0.15 and 0.3,4 and Seeger has determined the value to be 0.2, incorporating the assumption of a random distribution of dislocation directions.5 Short-range defects such as small and large Frank loops and precipitates are treated as hard impenetrable obstacles where dislocations bow around them by the Orowan mechanism The stress increment is expressed by p Dt ẳ Gb Nd =b ẵ4 where N is the defect density and d is the diameter The constant b ranges between and as suggested by Bement2 or as suggested by Olander.6 Voids and bubbles are also treated as hard obstacles using the same expression Precipitates and bubbles have been observed in austenitic stainless steels to nucleate and grow together.7 In this case, the bubbles and precipitates are considered as one obstacle where the hardening increment is calculated assuming rod geometry using a treatment by Kelly expressed by8: pffiffiffiffiffiffi pffiffiffi  6d 0:16Gb Nd pffiffipffiffiffiffiffiffi ln Bubble-precip ẳ ẵ5 3b Nd where the parameters are the same as for eqn [4] From the previous discussion, it can be inferred that because the nature of the irradiation-induced defects determines the degree of hardening, and because the nature, size, and density of defects is a strong function of temperature, radiation strengthening will be a strong function of irradiation temperature Figure illustrates Effect of Radiation on Strength and Ductility of Metals and Alloys 101 Relative contribution to strength Black dots 0.8 Frank loops 0.6 Bubbles-precipitates Network dislocation 0.4 0.2 0 100 200 Temperature (ЊC) 300 400 Figure Relative contribution to strengthening from irradiation-induced defects in the austenitic stainless steel, PCA, irradiated to dpa in the Oak Ridge Research Reactor Reproduced from Grossbeck, M L.; Maziasz, P J.; Rowcliffe, A F J Nucl Mater 1992, 191–194, 808 160 Irrad temp » Test Temp Strain rate ~ ´ 10-5 S-1 1000 140 900 371 ЊC Yield strength (MPa) 800 Test Symbol Temp (ЊC) 427 ЊC 700 371 427 483 538 593 649 704 760 816 483 ЊC 600 500 538 ЊC 593 ЊC 400 300 760 ЊC 100 816 ЊC 100 80 60 649 ЊC 200 120 Yield strength (ksi) 1100 40 704 ЊC 20 10 12 Neutron fluence (n cm−2 ) (E > 0.1 MeV) Figure Yield strength of 20% cold-worked type 316 stainless steel irradiated in the EBR-II Reproduced from Fish, R L.; Cannon, N S.; Wire, G L In Effects of Radiation on Structural Materials; Sprague, J A., Dramer, K., Eds.; ASTM: Philadelphia, PA, 1979; ASTM STP 683, p 450 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA strengthening from individual types of defects as a function of irradiation temperature for the austenitic stainless steel PCA.7 As can be seen from Figure 1, the black dot damage characteristic of low temperatures vanishes at temperatures over 300  C as Frank loops emerge Bubbles and precipitates also become major contributors to hardening above 200  C 1.04.3 Tensile Behavior Tensile behavior is determined by the irradiationinduced defect structure previously discussed Austenitic stainless steels will again be used for the example since they are typical of fcc alloys and in many respects to other alloys (see Chapter 2.09, Properties of Austenitic Steels for Nuclear Reactor Applications and Chapter 4.02, Radiation Damage in Austenitic Steels) The behavior of other example classes of alloys will be discussed in later sections of this chapter The tensile behavior characteristic of austenitic stainless steels is shown in Figure 2, where yield strength is plotted as a function of fluence and displacement level.9 Saturation in strength is clear with the saturation time becoming shorter as irradiation temperature is increased At temperatures above about 500  C, saturation is evident, but in this case, strength decreases This decrease is a result of recovery of the coldworked microstructure of the 20% cold-worked type 316 stainless steel presented in Figure Figure 102 Effect of Radiation on Strength and Ductility of Metals and Alloys 850 600 Ultimate tensile strength (UTS) 650 ЊC 20% cold-worked 400 800 750 200 Strength (MPa) Annealed 538 ЊC Yield strength (MPa) 600 20% cold-worked 400 Yield strength (YS) 650 600 550 200 450 427 ЊC 800 600 Annealed 200 10 20 30 Dose (dpa) 40 50 Figure Strength properties of 20% cold-worked type 316 stainless steel irradiated in EBR-II Reproduced from Allen, T R.; Tsai, H.; Cole, J I.; Ohta, J.; Dohi, K.; Kusanagi, H Effects of Radiation on Materials; ASTM: Philadelphia, PA, 2004; ASTM STP 1447, p Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA 20% cold-worked 400 20% CW 316 stainless steel irradiation temp 375–385 ЊC Test temperature 370 ЊC 500 Annealed 0 700 10 Neutron fluence (1022 n cm−2 ) Figure Yield strength of type 316 stainless steel irradiated in the EBR-II Reproduced from Garner, F A.; Hamilton, M L.; Panayotou, N F.; Johnson, G D J Nucl Mater 1981, 103 & 104, 803 shows yield strength resulting from the recovery of a cold-worked dislocation structure and the generation of a radiation-induced microstructure, resulting in a saturation strength independent of the initial condition of the alloy.10 Again, it is seen that the approach to saturation is faster with increasing temperature, with saturation achieved between and 10 dpa at 538 and 650  C, but 15–20 dpa is necessary to achieve saturation at 427  C Saturation is observed in yield strength curves for fluences as high as  1022 n cmÀ2 in a fast reactor (45 dpa), but more recent data show a hint of softening above 50 dpa,11,12 and other fast reactor data have shown a reduction in strength even for displacements below 50 dpa, as shown in Figure 4.13 This could result from coarsening of the microstructure or depletion of interstitial elements from the matrix due to precipitation This effect is also observed in martensitic steels irradiated to high dpa levels in the FFTF, but in this class of alloys, recovery of the martensitic lath structure is also a factor.12 However, even in austenitic steels, it is difficult to attribute such softening with certainty to an irradiation effect because of the strong influence of irradiation temperature on strength.14 Indeed, uncertainties in irradiation temperature are an inherent difficulty in neutron irradiation experiments 1.04.4 Effects of Neutron Spectrum This discussion has used neutron irradiations for illustration purposes Reactors provide an effective instrument for achieving high neutron exposures under conditions relevant for most nuclear applications However, reactor irradiations suffer from many difficult-to-control and, sometimes, uncontrolled variables The neutron energy spectrum is responsible for large differences in irradiation effects between different reactors The mechanism of atomic displacement is well understood.15 With a known neutron energy spectrum, neutron atomic displacements can be calculated as a function of fluence for a given reactor Transmutation of elements in the material under study, which is a strong function of neutron spectrum, results in wide variation in some mechanical properties This is of particular importance in applying fission reactor results to fusion In a fusion device, helium and hydrogen will be generated through (n,a) and (n,p) reactions in nearly all common structural materials Hydrogen has a very high diffusivity in metals so that an equilibrium concentration will be Effect of Radiation on Strength and Ductility of Metals and Alloys established at a level that is believed to be benign.16 By contrast, helium is insoluble in metals, segregating at grain boundaries and other internal surfaces and discontinuities Although helium is produced in all nuclear reactors, the thermal spectrum is responsible for the highest concentrations The largest contributors to helium in a thermal reactor are boron and nickel by the following reactions: 10 58 Bðn; aÞ7 Li Niðn; gÞ59 Ni 59 Niðn; aÞ56 Fe Boron is present as a trace element in most alloying elements but only at ppm levels Nickel is a major constituent of many alloys and a minor constituent of still others The two nickel reactions constitute a two-step generation process for helium, which starts slowly and accelerates as 59 Ni builds up in the alloy, limited only by the supply of 58Ni, which for practical purposes is often unlimited In austenitic alloys, the high flux isotope reactor (HFIR) has generated over 4000 appm He in austenitic stainless steels The generation rate is so high that multistep absorber experiments have been conducted to reduce the helium generation rate to that characteristic of fusion reactors, 12 appm He per dpa in austenitic stainless steels.17 (see Chapter 1.06, The Effects of Helium in Irradiated Structural Alloys) Other transmutation products may also complicate reactor irradiation studies Examples are the transmutation of manganese to iron by the following reaction: 55Mn (n,g) 56Mn ! 56Fe and the transmutation of chromium to vanadium by 50Cr (n,g) 51 Cr ! 51V The first reaction leads to loss of an alloy constituent, and the second leads to doping with an extraneous element However, neither of these reactions has been shown to significantly affect mechanical properties of steels.18 Helium remains the most studied transmutation product, and it can have profound effects on tensile properties, especially at high temperatures Experiments have been conducted in various reactors throughout the world to assess the effects of helium on mechanical properties of alloys.19 An interesting result is that helium has little effect on strength This is illustrated in Figure where a comparison has been made between austenitic steels irradiated in Rapsodie, a fast spectrum reactor, and steels irradiated in HFIR, a mixed-spectrum reactor with 103 a very high thermal flux The saturation yield strength of all alloys remains within a single scatter band.20,21 The tramp impurity elements sulfur and phosphorus have significantly high (n,a) cross sections at high energies, as shown in Figure Although the cross section for phosphorus is large only at energies characteristic of fusion, a boiling water reactor produces 500 appm He from sulfur and 40 appm He from phosphorus in eight years of operation An Liquid Metal Fast Breeder Reactor (LMFBR) can produce 100 times these concentrations All these elements are expected to enhance embrittlement when segregated to grain boundaries, but it remains to be determined which is more detrimental, helium, sulfur, or phosphorus 1.04.5 Tensile Ductility Tensile ductility is a more vulnerable parameter than strength to radiation effects since it tends to be very high in unirradiated austenitic stainless steels and is often reduced to quite low levels by irradiation It is also of more concern since strengthening, although not reliable due to its slow initiation, is usually a beneficial change In contrast, embrittlement is always detrimental Like strength, ductility exhibits saturation with increasing fluence, although the behavior is significantly more complex than that of strength The general trends in type 316 stainless steel are shown in Figure for material irradiated in the EBR-II These data are for the same specimens for which the yield strength was shown in Figure 2.9 Fast reactor data are used here to avoid the complication of helium effects Once stabilization of the dislocation microstructure is achieved, a smooth curve approaching an apparent saturation is observed More information can be gleaned from ductility data if they are viewed in terms of irradiation and test temperature Figure 822 shows total tensile elongation for a series of irradiated austenitic alloys at a displacement level of 30 dpa in both annealed and cold-worked conditions The room temperature ductility exceeds 10%, but it decreases rapidly with increasing temperature up to approximately 300  C and then exhibits the expected increase with temperature observed for unirradiated alloys Beyond 500  C, ductility again decreases with an onset of intergranular embrittlement resulting from helium introduced through transmutations in the thermal flux of the HFIR 104 Effect of Radiation on Strength and Ductility of Metals and Alloys 900 1.4988 sa 1.4970 sa + cw + a AISI 316 cw AISI 304 sa 800 Saturation yield strength (MPa) 700 600 500 400 US 316 20% cw HFIR JPCA SA HFIR JPCA 15% cw HFIR 300 316 20% cw EBR-II US PCA SA + 800 ЊC, h HFIR 200 Typical yield strength values of unirradiated solution annealed austenitic stainless steel 100 350 400 450 500 550 600 650 Irradiation and test temperature (ЊC) Figure Saturation yield strength as a function of temperature for austenitic alloys irradiated in Rapsodie, EBR-II, and high flux isotope reactor showing similar saturation strength Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C J Nucl Mater 1990, 174, 264 1.00E + 03 (n, a) cross-section (barns) 1.00E + 02 LMFBR flux (arb units) LWR flux (arb units) Sulfur (n, a) Phosphorus (n, a) 1.00E + 01 1.00E + 00 1.00E - 01 1.00E - 02 1.00E - 03 1.E - 11 1.E - 08 1.E - 05 1.E - 02 Energy (MeV) Figure Cross-section for (n,a) reactions as a function of neutron energy 1.E + 01 1.E + 04 Effect of Radiation on Strength and Ductility of Metals and Alloys 105 17 Irrad temp » Test temp Strain rate ~ ´ 10-5 S−1 16 15 14 Test Temp (ЊC) 371 427 483 538 593 649 704 760 816 Symbol 13 Total elongation (%) 12 11 10 593 ЊC 538 ЊC 816 ЊC 760 ЊC 649 ЊC 371 ЊC 704 ЊC Neutron fluence (n cm−2) (E > 0.1 MeV) 10 ´ 1022 Figure Total elongation of 20% cold-worked type 316 stainless steel irradiated in EBR-II Reproduced from Fish, R L.; Cannon, N S.; Wire, G L In Effects of Radiation on Structural Materials; Sprague, J A., Dramer, K., Eds.; ASTM: Philadelphia, PA, 1979; ASTM STP 683, p 450 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA ORNL-DWG 89-13395 20 30 DPA + + J + Total elongation (%) 15 * C US PCA 25% CW HFIR JPCA ANN HFIR JPCA 15% CW HFIR US 316 20% CW HFIR US PCA B3 HFIR 316 20% CW EBR-II J J 10 + J J* 0 + J J C 0 100 200 J C 300 400 Temperature (ЊC) 500 C.J C 00 600 700 Figure Total elongation as a function of irradiation and test temperature for fast (EBR-II) and mixed-spectrum (high flux isotope reactor) reactor irradiation Uniform elongation, the elongation at the onset of plastic instability, or necking, appears to be most sensitive to the effects of irradiation and, in general, is less dependent on specimen geometry than other parameters such as total tensile elongation The low values of uniform elongation are often cause for great concern, which is usually justified However, it should be borne in mind that if stresses remain below the yield stress of a metal, elongation becomes a secondary concern As long as limited plastic deformation relieves the stress that produced it, a structure remains intact The high level of irradiation strengthening observed at temperatures below 300  C, which is due to black dot defect clusters and small loops, also results in low ductility throughout this temperature range Small helium bubbles and helium-defect 106 Effect of Radiation on Strength and Ductility of Metals and Alloys 20 Uniform elongation (%) 10 dpa 250 ЊC Sym (%) 15 0.20 0.20 0.33 0.29 0.31 0.28 0.23 0.18 0.30 0.20 0.10 10 316 SS PCA 0 100 200 300 400 Temperature (ЊC) Symbols 500 600 316 20% CW EBR-II 316 ANN EBR-II US 316 20% CW ORR US PCA 25% CW HFIR US PCA 25% CW ORR US 316 20% CW HFIR US PCA 25% CW HFR US 316 20% CW HFR 316 20% CW DO HFIR US PCA 25% CW BR2 US 316 20% CW BR2 EC 316 ANN HFIR JPCA ANN HFIR EC 316 ANN BR2 JPCA ANN HFR EC 316 ANN HFR JPCA 15% CW HFIR US PCA B3 HFIR 700 Figure Uniform elongation as a function of irradiation and test temperature at a displacement level of 10 dpa The trend curves are for type 316 stainless steel and PCA Reproduced from Grossbeck, M L.; Ehrlich, K.; Wassilew, C J Nucl Mater 1990, 174, 264 clusters also contribute to hardening and reduction in ductility, but this form of helium embrittlement is not related to the severe intergranular embrittlement that is observed above 500  C Both these effects are apparent in Figure where uniform elongation for an extensive set of austenitic alloys irradiated in thermal and fast spectrum reactors is shown.11 The specimens irradiated in the fast spectrum ( 0.1 MeV) Total elongation (%) 28 24 20 16 12 f t = 12 ´ 1022 n cm–2 - 700 - 600 - 500 - 400 - 300 - 200 - 100 100 200 300 400 500 Test Irradiation (ЊC) – temperature temperature Figure 12 Total elongation of 20% cold-worked type 316 stainless steel irradiated in FFTF to displacement levels of 48–63 dpa Hamilton, M L.; Cannon, N S.; Johnson, G D In Effects of Radiation on Materials; Brager, H R., Perrin, J S., Eds.; ASTM: Philadelphia, PA, 1982; ASTM STP 782, p 636 Reprinted, with permission, from Effects of Radiation on Structural Materials, copyright ASTM International, West Conshohocken, PA different holding times at various temperatures before tensile testing Migration of interstitial solutes to moving dislocations is a candidate mechanism for this phenomenon 1.04.7 Ferritic–Martensitic Alloys 1.04.7.1 Introduction The class of ferritic–martensitic alloys with chromium concentrations in the range of 9–12% has attracted interest in the fast reactor programs because of its radiation resistance, in particular, very low swelling and low irradiation creep Alloys such as Sandvik HT-9 (12Cr1Mo.6Mn.1Si.5W.3V)) and other alloys of this class were irradiated in the EBR-II,24 in research reactors25 and with heavy ions.26 The quantitative results from the ion irradiations in this class of alloys and the low neutron absorption cross section led to inclusion of ferritic alloys into the fast reactor alloy development programs, in particular in the United States in the mid-1970s The radiation resistance has been confirmed to displacement levels of 70 dpa.12,14 Further interest in this class of alloys was initiated by the fusion reactor programs in Europe and the United States when the necessity for low neutron activation structural materials was realized Further research on martensitic alloys by fusion programs in Europe, the United States, and Japan led to the development of low-activation alloys by replacing elements that result in long-term activation products Molybdenum and niobium, both of which result in longlived activation products, were replaced by tungsten and tantalum This research led to radiation-resistant alloys with a fracture toughness superior to that of the commercial alloys even in the unirradiated condition.27 The compositions of representative members of this class of alloys referred to in this chapter are presented in Table An excellent review of irradiation behavior of this class of alloys has been published by Klueh and Harries.27 Details of the metallurgy of martensitic alloys appears in Chapter 4.03, Ferritic Steels and Advanced Ferritic–Martensitic Steels 1.04.7.2 Tensile Behavior Unlike the tensile behavior of fcc metals, where there is a smooth increase in strength as plastic deformation proceeds and work hardening progresses, bcc metals typically exhibit a load drop almost immediately following the onset of plastic deformation Interstitial solutes such as carbon in steels effectively lock dislocations leading to a longer period of elastic deformation after which generation of new dislocations results in a load drop, or yield point, until Effect of Radiation on Strength and Ductility of Metals and Alloys Table 109 Nominal or typical compositions of ferritic–martensitic alloys cited Steel type 12Cr–MoVW 8Cr–2WVTa 9Cr–1WVTa 9Cr–1GeV 12Cr–1MoVNiNb 10Cr–MoVNiNb 9Cr–2WVTa 9Cr–2WVTa 9Cr–2WVTa 7Cr–2WVTa 2.25Cr–2WVTa 12Cr–2WVTa 12Cr–2WVTa 9Cr–1MoVNb 9Cr–2Mo–1Ni 9Cr–2W 9Cr–2W Designation HT-9 F82H OPTIFER Ia OPTIFER II MANET I MANET II 9Cr–2WVTa JLF-1 JLF-2 JLF-3 JLF-4 JLF-5 JLF-6 T91 JFMS NFL-0 NFL-1 Composition (wt%) C Si Mn Cr Ni Mo V 0.20 0.1 0.1 0.1 0.14 0.1 0.1 0.1 0.1 0.09 0.1 0.09 0.10 0.1 0.05 0.10 0.10 0.38 0.60 11.95 8.0 9.0 0.60 1.0 10.8 10.0 8.7 9.0 9.16 7.03 2.23 11.99 12.00 9.0 9.6 8.65 9.01 0.9 0.7 0.30 0.2 0.25 0.3 0.2 0.2 0.23 0.19 0.20 0.20 0.20 0.19 0.19 0.2 0.12 0.25 0.26 0.2 0.3 0.67 0.056 0.042 0.4 0.46 0.45 0.45 0.50 0.48 0.46 0.4 0.58 0.050 0.53 0.94 0.75 0.6 1.0 2.3 v Nb W B Other 0.52 1.0 0.16 0.15 0.009 0.007 2.2 1.97 1.93 1.97 1.97 1.98 1.94 0.04 Ta 0.07 Ta 1.1Ge 0.06 Zr 0.03 Zr 0.06 Ta 0.07 Ta 0.07 Ta 0.07 Ta 0.07 Ta 0.07 Ta 0.07 Ta 0.08 0.06 1.92 2.06 0.0032 Source: Maloy, S A.; Toloczko, M B.; McClellan, K J.; et al J Nucl Mater 2006, 356, 62; Klueh, R L.; Harries, D R High-Chromium Ferritic and Martensitic Steels for Nuclear Applications; ASTM: Philadelphia, PA, 2001; Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M L.; Narui, M J Nucl Mater 1999, 271 & 272, 145; Kurishita, H.; Kayano, H.; Narui, M.; Kimura, A.; Hamilton, M L.; Gelles, D S J Nucl Mater 1994, 212–215, 730 (´ 103) 120 ´ 1020 nvt 100 ´ 1018 nvt Stress (psi) 80 Unirradiated 1.7 ´ 1019 nvt 60 40 20 0 0.05 0.10 0.15 0.20 0.25 Strain (in in.–1) 0.30 0.35 0.40 Figure 13 Stress–strain curves for low-carbon steel weld material irradiated at 80  C (nvt is fluence in neutrons/cm2) Reproduced from Wilson, J C Effects of irradiation on the structural materials in nuclear power reactors In Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, United Nations, 1958; Vol 5, p 431 terminated by the work hardening mechanism of dislocation interaction.28 Upon irradiation, the load drop is frequently masked by an early termination of work hardening, leading to very low values of uniform elongation This behavior is evident even at displacement levels below 0.01 dpa and is illustrated in Figure 13 from research presented at the Second Atoms for Peace Conference in 1958.29 Extreme irradiation hardening and severe plastic instability are clearly illustrated by this early research More recent alloys with more careful control of impurities and controlled processing 110 Effect of Radiation on Strength and Ductility of Metals and Alloys 1400 JFMS specimens tested at 25 ЊC after irradiation in FFTF Engineering stress (MPa) 1200 35.3 dpa, Tirr = 390 ЊC 9.8 dpa, Tirr = 373 ЊC 22.2 dpa, Tirr = 390 ЊC 44 dpa, Tirr = 427 ЊC 1000 800 dpa 600 400 200 0 12 14 16 10 Engineering strain (%) 18 20 22 Figure 14 Stress–strain curves for JFMS alloy irradiated in FFTF to 44 dpa at temperatures of 373–427  C and tested at 25  C Reproduced from Maloy, S A.; Toloczko, M B.; McClellan, K J.; et al J Nucl Mater 2006, 356, 62 800 MATRON/tensile(S) 700 656–700 K Yield stress (MPa) 600 500 400 300 200 100 0 10 20 30 40 50 60 70 Displacement damage (dpa) Figure 15 Yield stress as a function of displacement level for martensitic alloys irradiated in FFTF or EBR-II Reproduced from Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M L.; Narui, M J Nucl Mater 1999, 271 & 272, 145 have led to ferritic alloys with appreciable work hardening even at high displacement levels Figure 14 shows tensile curves for the 9Cr–2Mo–1Ni steel, JFMS, neutron irradiated and tested at room temperature Uniform elongations of several percent are evident, a reasonable value for irradiated steels.12 A plot of yield stress as a function of displacement damage level is shown in Figure 15 for low-activation ferritic alloys irradiated in fast reactors,30 and plots of yield stress and total elongation are shown in Figure 16 for fast and mixed-spectrum reactors.31 Unlike the austenitic alloys, the martensitic alloys rapidly reach a peak in strength then soften with further irradiation followed by near saturation in strength beginning at about 30 dpa Total elongation follows a corresponding pattern, demonstrating Effect of Radiation on Strength and Ductility of Metals and Alloys 111 800 JLF-5 (12Cr–2WVTa) Total elongation (%) JLF-5 (12Cr–2WVTa) 600 Yield stress (MPa) 30 JLF-4 (2.25Cr–2WVTa) 700 500 F82H 10B 400 300 F82H STD JLF-1 (9Cr–2WVTa) F82H (8Cr–2WVTaB) JLF-1 (9Cr–2WVTa) 20 F82H STD F82H (8Cr–2WVTaB) 10 JLF-3 (7Cr–2WVTa) 200 F82H 10B : Irradiated in HFIR (spec SS3) Others in FFTF (spec TS(s)) 100 JLF-4 (2.25Cr–2WVTa) 0 10 20 30 40 50 60 Displacement damage (dpa) 0 10 20 30 40 JLF-3 (7Cr–2WVTa) 50 60 Displacement damage (dpa) Figure 16 Yield strength of martensitic alloys following irradiation at 400  C in FFTF or high flux isotope reactor Reproduced from Kohyama, A.; Hishinuma, A.; Gelles, D S.; Klueh, R L.; Dietz, W.; Ehrlich, K J Nucl Mater 1996, 233–237, 138 inverse behavior This appears at first to be a phenomenon totally different from that which occurs in austenitic alloys However, the operable mechanisms are really the same, just occurring at lower fluences Hardening mechanisms are similar, occurring by point defect clusters at low irradiation temperatures and transitioning to loops and network dislocations and precipitation as temperature is increased The martensitic alloys are more complex in that the initial microstructure is determined by the heat treatment The alloys are used in the normalized and tempered condition produced by austenitizing the alloy and quenching below the A1 temperature to produce martensite The martensite is then tempered below the A1 temperature Precipitates, primarily carbides such as M23C6, form on prior austenite grain boundaries and on the martensite laths In the case of the bcc alloys, more rapid radiation-enhanced diffusion results in irradiation-induced recovery and precipitate growth at lower fluence than in the fcc alloys Recall that in the austenitic alloys, saturation was reached and sustained for a long period until microstructural coarsening resulted in a slight decrease in strength at high displacement levels The pronounced peak in strength results from rapid hardening due to irradiation-produced defects, but the effect of irradiation hardening is offset by irradiation-enhanced recovery, resulting in a decrease in strength and hence a peak in strength.32 The martensitic alloys demonstrate the same phenomena but at a lower fluence Figure 17 shows tensile test results from alloys irradiated in FFTF to approximately 30 dpa Tests at the irradiation temperatures show high and nearly constant values of total tensile elongation at temperatures above 425  C.33 Somewhat similar behavior of elongation was also exhibited by 9Cr–1MoVNb steel irradiated to 12 dpa in EBR-II.34 This alloy also exhibits an increase in total elongation at 550  C (Figure 18) The apparent saturation in strength above about 450  C is also in agreement with hardness measurements In a study of irradiated HT-9 and 9Cr–1MoVNb, Hu and Gelles observed that hardness retained its unirradiated value upon irradiation to 26 dpa when irradiated at temperatures above about 450  C in EBR-II.35 From this behavior, it would normally be concluded that fracture properties would remain unchanged or improve with increasing temperature Although fracture toughness testing is used extensively in the study and certification of irradiated alloys, the Charpy impact test is more commonly used in alloy development and fundamental research because the test requires smaller specimens and can be conducted more easily Ductile to brittle transition temperature (DBTT) is a useful tool for comparison of alloys and assessment of radiation damage Charpy impact testing was conducted by Hu and Gelles who observed that in the case of the 9Cr–1MoVNb, the DBTT did in fact retain essentially its unirradiated value for irradiation 112 Effect of Radiation on Strength and Ductility of Metals and Alloys 12 Total elongation (%) Yield stress, sy (MPa) 800 600 400 Unirr 200 (b) 600 700 800 900 1000 Uniform elongation (%) Ultimate tensile stress, su (MPa) (a) 800 600 Unirr 400 Total elongation, er (%) 9Cr–1 MoVNb steel Unirradiated Aged Irradiated Test temperature @ Irradiation temperature @ Aging temperature 200 0 600 700 800 900 400 500 Test temperature (ЊC) 600 Figure 18 Elongation of 9Cr–1MoVNb irradiated in EBR-II to 0.9 dpa Reproduced from Klueh, R L.; Vitek, J M J Nucl Mater 1985, 132, 27 30 20 Unirr 10 (c) 12 600 700 800 Irradiation temperature (K) 900 NFL-0 NFL-1 Figure 17 Tensile properties of two Fe–9Cr–2W steels with and without small additions of boron, irradiated in FFTF to approximately 30 dpa and tested at room temperature Reproduced from Kurishita, H.; Kayano, H.; Narui, M.; Kimura, A.; Hamilton, M L.; Gelles, D S J Nucl Mater 1994, 212–215, 730 temperatures above 450  C at 26 dpa but not for lower irradiation temperatures, as shown in Figure 19.35 However, the DBTT of HT-9 failed to retain its unirradiated value despite the absence of an increase in strength and hardness In both cases the upper shelf energy was reduced by the irradiation, indicating some changes in the irradiation microstructure The shift in the DBTT for both alloys at 13 and 26 dpa is shown as a function of irradiation temperature in Figure 20 The retention of the increase in the DBTT at high temperatures illustrates the caution that must be used in assessing ferritic alloys In impact testing, fracture is generally initiated at carbide particles Even though coarsening of the carbides results in less impediment to dislocation motion, and thus less hardening, the stress intensity factor increases for a nucleating crack at a hard particle so that the effective crack length is the crack nucleus plus the diameter of the carbide particle.36 The result is that fracture toughness can increase with irradiation temperature Despite the caution that must be taken in making generalizations based upon tensile behavior, development of low-activation martensitic alloys has led to alloys with very favorable fracture properties The DBTT is shown in Figure 21 for two low-activation alloys, 9Cr–2WVTa and 9Cr–2WVTa–2Ni, with the conventional Ht-9 and 9Cr–1MoVNb for comparison The irradiation was done in EBR-II at irradiation temperatures from 376 to 405 and to displacement levels of 23–33 dpa After irradiation, the two Effect of Radiation on Strength and Ductility of Metals and Alloys 113 9Cr–1Mo base metal (TV series), 26 dpa Ti = 390 ЊC Ti = 500 ЊC Normalized fracture energy (J cm−2) 600 Control 480 TV11 360 240 TV22 120 -150 -100 -50 50 Test temperature (ЊC) 100 150 HT-9 base metal (TT series), 26 dpa Normalized fracture energy (J cm−2) 400 320 Ti = 390 ЊC Ti = 450 ЊC Ti = 500 ЊC Ti = 550 ЊC Control 240 160 TT23 TT15 TT20 80 -100 -50 TT28 50 100 150 Test temperature (ЊC) 200 250 Figure 19 Charpy impact test results for 9Cr–1Mo and HT-9 irradiated in EBR-II to 26 dpa Reproduced from Hu, W L.; Gelles, D S Influence of Radiation on Material Properties; ASTM: Philadelphia, PA, 1987; ASTM STP 956, p 83 Reprinted, with permission, from Influence of Radiation on Material Properties, copyright ASTM International, West Conshohocken, PA conventional alloys had DBTT values above room temperature, whereas the values for the two reduced-activation steels were below À75  C.32 The figure clearly shows that HT-9 is not the alloy of choice for nuclear applications Despite the Ni content of 9Cr–2WVTa–2Ni, this class of alloys can prove useful in nuclear applications such as fast reactors where the thermal flux is small and the very high-energy neutrons characteristic of a fusion reaction are absent 1.04.7.3 Helium Effects Helium effects are important for systems that generate high-energy neutrons such as fusion reactors and spallation targets that encounter high-energy protons The 14 MeV neutrons produced by D–T fusion will produce (n,a) reactions in nearly all common structural elements such that ferritic steels are not exempt from helium generation However, in the absence of nickel, helium generation rates are lower 114 Effect of Radiation on Strength and Ductility of Metals and Alloys 140 Shift in transition temperature (ЊC) 13 dpa HT9 120 26 dpa HT9 100 26 dpa Mod 9Cr–1Mo 13 dpa Mod 9Cr–1Mo 80 60 40 20 500 450 Irradiation temperature (ЊC) 400 550 Figure 20 Shift in ductile to brittle transition temperature as a function of irradiation temperature for 13 and 26 dpa following EBR-II irradiation Reproduced from Hu, W L.; Gelles, D S Influence of Radiation on Material Properties; ASTM: Philadelphia, PA, 1987; ASTM STP 956, p 83 Reprinted, with permission, from Influence of Radiation on Material Properties, copyright ASTM International, West Conshohocken, PA 75 Transition temperature (ЊC) 50 Unirradiated Irradiated 25 -25 9Cr-1MoVNb 12Cr-1MoVW -50 -75 -100 -125 9Cr-2WVTa 9Cr-2WVTa-2Ni Figure 21 Ductile to brittle transition temperature before and after irradiation in EBR-II to 23–33 dpa at 376–405  C comparing conventional alloys with irradiation-resistant martensitic alloys Reproduced from Klueh, R L.; Sokolov, M A.; Hashimoto, N J Nucl Mater 2008, 374, 220 Reprinted, with permission, from Influence of Radiation on Material Properties, copyright ASTM International, 100 Barr Harbor Drive, West Conshohocken, PA 19428 In austenitic stainless steels, the high nickel content is used to introduce helium to simulate the fusion environment However, the nickel content is so high that unrealistically high concentrations of helium are produced in a thermal or mixed-spectrum reactor As a result, spectral tailoring is necessary to achieve the correct He per dpa ratio.37 In the case of martensitic alloys, several techniques have been used to generate helium in fission reactors Doping with natural nickel has been used in 9Cr and 12Cr alloys27,38,39 and isotopically separated nickel has been used to discriminate against the effect of nickel as opposed to the effect of He The isotope 59Ni has been used as it will generate He but 60Ni, used as a control, will not.40 Doping with 10B has also been used to introduce He into this class of alloys, leading to concentrations of several hundred parts per million.41–43 Nickel doping experiments with 12Cr–1MoVW were conducted in the HFIR using and 2% Ni to generate up to about 300 appm He This experiment demonstrated what appeared to be a strengthening effect of helium.44 However, it was determined that these results were clouded by the fact that doping with Ni lowers the A1 temperature of the alloy, leading to untempered martensite upon cooling from the tempering temperature Adjustments were made in the tempering temperature, but the additional variables cast doubt on the results.44 The isotope separation experiments and the boron doping experiments found no clear indication of a helium effect.40,41 Several mechanisms for hardening by helium have been identified,45 but Klueh and Harries27 have, after a careful review, come to the conclusion that the effect of helium on strength and ductility below 500  C is inconclusive and that if there is an effect, it is probably small and of minor significance for the conditions examined As with austenitic stainless steels, the effect of hardening, if it exists for ferritic alloys, is of minor importance compared with high-temperature intergranular embrittlement High He concentrations and higher temperatures have been investigated in martensitic alloys by means of helium implantation with accelerators Hasegawa implanted He at 600  C to levels of 500 appm and performed tensile tests at this temperature.46 However, all fractures were transgranular Bae et al.47 implanted similar levels of helium, 500 appm, in the 12Cr-steel, MANET, at temperatures as high as 500  C They observed hardening, shown in Figure 22, but again they were only transgranular fractures Bae et al.47 also observed no synergistic effect of 500 appm H and 500 appm He and no effect of hydrogen alone at this level Jung et al.48 found that He implanted in the 9Cr alloy, EUROFER97, to concentrations as high as 1250 appm produced both hardening and reduction in ductility when implanted at 250  C and tested at Effect of Radiation on Strength and Ductility of Metals and Alloys 115 20 900 15 Total elongation (%) MANET Yield stress (MPa) 700 500 MANET 10 300 DSA 100 0 600 200 400 Irradiation and test temperature (ЊC) 800 200 400 600 Irradiation and test temperature (ЊC) 800 0.32 dpa, 500 appm H, 500 appm He 0.30 dpa, 500 appm He 0.02 dpa, 500 appm H Unirradiated Figure 22 Yield strength and total elongation for the 12Cr steel, MANET cyclotron implanted with He and H Reproduced from Bae, K K.; Ehrlich, K.; Mosalang, A J Nucl Mater 1992, 191–194, 905 1200 0.125 EUROFER97 0.25 Timp = 250 ЊC s (MPa) 1000 0.25 0.06 0.125 0.06 800 600 cHe(at.%) Ttest = 25 ЊC 400 Ttest = 250 ЊC 200 0 e (%) 10 Figure 23 Stress–strain curves for EUROFER97 for cyclotron implantation of helium Jung, P.; Henry, J.; Chen, J J Nucl Mater 2005, 343, 275 either 25 or 250  C Their data demonstrate both hardening and loss of work hardening with helium, as clearly shown in Figure 23 Severe intergranular embrittlement was, in fact, demonstrated by Jung et al in the 9Cr steels T91 and EM10 when levels of 5000 appm were attained by cyclotron implantation.49 In cases where He was implanted at 250  C and tested at 25  C and at 250  C, clear intergranular fracture was experienced For implantation temperatures of 550  C, where intergranular embrittlement would be expected in austenitic stainless steels, the fracture surfaces indicated a return of some ductility with little or no intergranular fracture, particularly when tested at 550  C, Figure 24 The latter case can perhaps be explained by the loss of strength at 550  C where plastic deformation blunts any nucleating crack before the local stress at a grain boundary is sufficiently high for grain separation In the case of implantation at 550  C and testing at 25  C, the fracture mechanism is less clear Here, the temperature is sufficiently high for diffusion of He to the grain boundaries, but perhaps capture by other sinks prevents high levels of He at the boundaries More investigation of the high-temperature He embrittlement is necessary to determine the underlying mechanism of fracture It is clear, however, that helium embrittlement is more effective in austenitic stainless steels than in ferritic–martensitic steels Since nickel is the largest source of helium in a fast neutron environment, ferritic alloys clearly have a lower He generation rate than austenitic steels.45 The combination of higher resistance to helium embrittlement and the lower generation rate of He in ferritic alloys makes this class of alloys more favorable with respect to He embrittlement 116 Effect of Radiation on Strength and Ductility of Metals and Alloys 250/25 550/25 250/250 550/550 Figure 24 Fracture surfaces of T91 following helium implantation at 250 and 550  C and tested at the implantation temperatures indicated Reproduced from Jung, P.; Henry, J.; Chen, J.; Brachet, J C J Nucl Mater 2003, 318, 241 1.04.8 Refractory Metals 1.04.8.1 Tensile Behavior The refractory metals are the metals in groups V and VI of the periodic table: vanadium, niobium, and tantalum in group V and chromium, molybdenum, and tungsten in group VI All have the characteristic of a high melting point, hence the term refractory The group VI metals are typically brittle, even without irradiation For example, chromium is almost never used pure or as a major alloy element, although it is invaluable as a minor alloying element Molybdenum and tungsten are both brittle in nature but can be made into useful structural alloys by controlling interstitial impurities and by the addition of minor elements In contrast to the brittle behavior of the group VI metals, the group V metals are inherently ductile Structural alloys based upon this group have been developed, primarily for very high temperature and space applications The primary disadvantage of the refractory metals is their formation of volatile oxides as opposed to protective oxide layers Vanadium and molybdenum oxides have melting points below metal working temperatures so that the metals become wet and can have liquid oxide drip off them Unlike the tensile behavior of fcc metals, where there is a smooth increase in strength as plastic deformation proceeds and work hardening progresses, bcc metals typically exhibit a load drop, or yield point, almost immediately following the onset of plastic deformation, as discussed in Section 1.04.7.2 In the case of refractory metals, mechanical properties are largely determined by interstitial solutes High purity refractory metals not exhibit a yield point but behave more like fcc metals Niobium alloys irradiated in Li at 1200 C for over three months in EBR-II had total elongations of about 60% Despite any irradiation hardening, the near absence of oxygen resulted in a very soft material at these high temperatures.50 However, since interstitials are nearly always present, tensile behavior is more typically characteristic of bcc metals Irradiation-produced defects interact with interstitial elements, resulting, in some cases, in severe embrittlement The tantalum alloy, Westinghouse T111 (Ta–8W–2Hf) is used in Figure 25 to illustrate a commonly observed phenomenon of plastic instability.51 Plastic deformation becomes local, with high levels of slip on closely spaced planes where dislocations sweep out the irradiation-generated defects giving rise to local channels of very high deformation This phenomenon, called channel deformation, is very common in irradiated metals The result is a sudden and severe load drop with the fracture surface showing what appears to be a completely ductile chisel point fracture.52 Addition of 405 wt ppm oxygen to T-111 results in a cleavage fracture with no measurable plastic deformation, as shown in Figure 26.51 In both Figures 25 and 26, corresponding unirradiated alloys are shown demonstrating ductile behavior In the unirradiated condition, the addition of 400 wt ppm oxygen has only minor effects on strength and ductility, as can be concluded by a comparison of Figures 25 and 26 However, irradiation hardening superimposed upon the oxygen interstitial hardening appears to raise the yield stress above the cleavage stress for the alloy It is suggested that interstitial solutes such as oxygen diffuse to irradiation-produced defect clusters, enhancing their hardening effect.53,54 All three behaviors are observed in irradiated refractory metals: ductile with hardening, plastic instability, and cleavage fracture in the elastic range.55 The synergism between interstitial hardening and irradiation hardening does not necessarily lead to immediate catastrophic embrittlement This behavior is shown in Figure 27 for vanadium containing a very high level of oxygen, 2100 wt ppm Irradiation to a fluence level of 1.5  1019 (E > MeV) leads to the familiar plastic instability but with several per cent plastic strain.53 Effect of Radiation on Strength and Ductility of Metals and Alloys 117 1600 111T 1400 Unirradiated EBR-II irradiated ft (E > 0.1 MeV) = 1.6 × 1026 n m–2 Tirrad = Ttest = 873 K Stress (MPa) 1200 1000 800 * 600 400 200 * 10 12 14 16 Strain (%) 18 20 22 24 26 28 Figure 25 Stress–strain curves for the Ta alloy, T-111 showing characteristic tensile behavior following irradiation in EBR-II to dpa at 600  C Reproduced from Grossbeck, M L.; Wiffen, F W In Space Nuclear Power Systems; El-Genk, M S., Hoover, M D., Eds.; Orbit Book Co.: Malabar, FL, 1986; Vol III, p 85 1400 111 T + 405 wt ppm oxygen Unirradiated EBR-II irradiated ft (E > 0.1 MeV) = 1.4 × 1026 n m–2 Tirrad = Ttest = 853 K * 1200 Stress (MPa) 1000 800 600 400 * 200 0 10 12 14 16 Strain (%) 18 20 22 24 26 28 Figure 26 Stress–strain curves for oxygen-doped T-111 T unirradiated and irradiated in EBR-II to dpa at 580  C Reproduced from Grossbeck, M L.; Wiffen, F W In Space Nuclear Power Systems; El-Genk, M S., Hoover, M D., Eds.; Orbit Book Co.: Malabar, FL, 1986; Vol III, p 85 Interstitial solutes, especially oxygen, may be controlled by the addition of gettering elements In the vanadium system, titanium has been successful Alloys in the V–Cr–Ti system have been studied for application to fusion reactors In refractory metal alloys, it is the oxygen in solution that is detrimental, so that the oxygen must be combined with the titanium.56 This usually requires a heat treatment of sufficiently long times and high temperatures to precipitate the oxygen In the vanadium–titanium system, 118 Effect of Radiation on Strength and Ductility of Metals and Alloys 90 Vanadium–2100 wt ppm oxygen Tensile tests at 300 ЊK and 1.67 ´ 10-4 s-1 80 Irradiated, 1.5 ´ 1019 n cm−2 (E > MeV) 70 Stress (Kpsi) 60 Unirradiated 50 40 30 20 10 Uniform elongation 12 14 10 Strain (%) 16 18 20 22 24 Figure 27 Stress–strain curves for oxygen-doped vanadium irradiated at 85  C at the Ames Laboratory Research Reactor Reproduced from Wechsler, M S.; Alexander, D G.; Bajaj, R.; Carlson, O N In Defects and Defect Clusters in B.C C Metals and Their Alloys, Nuclear Metallurgy; Arsenault, R J., Ed.; National Bureau of Standards: Gaithersburg, MD, 1973; Vol 18, p 127 a heat treatment of two hours at 950  C has been found sufficient to achieve ductility, whether or not it is to be irradiated.57,58 Confusion over oxygen present in solution and oxygen combined in precipitates is believed to be one reason for the disparity in tensile data for this class of alloys and perhaps accounts for the relatively high level of ductility observed in Figure 27 Upon irradiation of alloys in the range of V–3– 5Cr–3–5Ti in the HFIR, no cleavage fracture without plastic deformation was observed.59,60 However, plastic instability was commonly observed at irradiation and test temperatures below 400  C Irradiations in the range of 4–6 dpa in the HFIR produced uniform elongations from 0.2 to 0.6% and total elongations below 4% Corresponding irradiations at 500  C did not reveal plastic instability and produced uniform elongations in the range of 2–5%.59,60 Irradiations to 3–5 dpa in the advanced test reactor (ATR) demonstrated plastic instability for irradiation and test temperatures of about 200  C, with uniform elongations below 0.5%.61 Irradiations conducted in the high flux beam reactor (HFBR) at exposures of only 0.1 and 0.5 dpa corroborated these results and demonstrated a transition in the fracture mechanism between 300 and 400  C, resulting in a significant increase in ductility at temperatures above 400  C, Figure 28.62 1.04.8.2 Helium Effects Helium is conveniently introduced into nickelbearing alloys through thermal neutron irradiation Although helium is usually detrimental, especially if the material is to be subsequently welded,63 it offers a method to simulate the production of helium expected in the very hard spectrum of a fusion reactor In the case of vanadium and other refractory metal alloys, the effect of helium has been studied using two primary methods of introduction of helium One method is implantation of a-particles with an accelerator; the other is the use of the decay of tritium Tritium rapidly diffuses into group V refractory metals at elevated temperatures The elevated temperature serves more to dissolve the protective oxide layer than to accelerate the kinetics of dissolution The tritium thus introduced is permitted to decay, by b-decay with a residual nucleus of He Helium-doped specimens have subsequently been neutron irradiated to study the synergistic effects of helium and atomic displacement damage A limited number of experiments have used techniques to simultaneously implant He and produce atomic displacements through an irradiation environment of Li The concept of introducing tritium into an irradiation capsule with the specimens in contact with lithium has been investigated to study vanadium Effect of Radiation on Strength and Ductility of Metals and Alloys 12 140 0.0002 0.5 dpa (V1–V3) Ttest = Tirr 100 200 300 Molybdenum 120 0.1 dpa V4 Ultimate tensile stress (1000 psi) Uniform elongation (%) 10 -2 119 400 500 600 100 700 40 60 Elongation (%) alloys with the He/dpa ratio characteristic of a fusion environment The tritium charge, the production of tritium from lithium, and the production of tritium from 3He are some of the important considerations in the design of the experiment.64 Although conceptually valid, the desired results have not yet been obtained with experiments of this type Cyclotron-implanted helium has been used, also to study the effects of the fusion irradiation environment Tanaka showed severe embrittlement with the introduction of 90 and 200 appm He at 700  C in V–20Ti.65 Grossbeck and Horak showed that a level of 80 at ppm He implanted as part of the same experiment had no significant effect on elongation in V–15Cr–5Ti at 700  C.52 Braski also observed no significant effect on ductility in V–15Cr–5Ti at 600  C with similar levels of He introduced from decay of tritium.66 The alloys, Vanstar-7 and V–3Ti–1Si, were also investigated, in some cases with an improvement in ductility upon introduction of helium.66 Following irradiation, severe embrittlement was observed in V–15Cr–5Ti at 600  C in tritium trick samples by Braski66 and at 625  C in cyclotron-implanted samples by Grossbeck and Horak.52 Irradiation experiments with refractory metals, unless using a Li environment, frequently subject the specimens to contamination by interstitial impurities, also leading to embrittlement.52,67 Unalloyed molybdenum, Mo–0.5Ti, and Mo–50Re were irradiated in EBR-II by Wiffen at exposures of 0.0002 60 20 Test temperature (ЊC) Figure 28 Uniform elongation of V–4Cr–4Ti irradiated in the high flux beam reactor Reproduced from Snead, L L.; Zinkle, S J.; Alexander, D J.; Rowcliffe, A F.; Robertson, J P.; Eatherly, W S Fusion Reactor Materials Semiannual Progress Report for Period Ending, Dec 31, 1997; DOE/ER-0313/23, p 81 80 Controls 3.5 or 4.0 ´ 1022 n cm−2 at 455 ЊC 6.1 ´ 1022 n cm−2 at 857–1136 ЊC 4.0 ´ 1022 n cm−2 at 455 ЊC + 1050 ЊC anneal 40 20 0.0002 -200 200 400 Test temperature (ЊC) 0.0002 600 Figure 29 Ultimate tensile strength and total elongation for molybdenum in the unirradiated condition and irradiated in EBR-II to 20–30 dpa Dashed lines connect results where irradiation conditions or strain rates were not constant Reproduced from Wiffen, F W In Defects and Defect Clusters in B.C.C Metals and Their Alloys, Nuclear Metallurgy; Arsenault, R J., Ed.; National Bureau of Standards: Gaithersburg, MD, 1973; Vol 18, p 176 3.5–6.1 n cmÀ2 (E > 0.1 MeV) (18–32 dpa).68 Although Mo alloys are known to exhibit increased ductility with increasing temperature in the unirradiated condition, at temperatures above 400–550  C, all three materials suffered plastic instability with uniform elongations below about 0.5% This effect is shown in Mo in Figure 2968 where irradiation temperature is shown to be the critical parameter and where specimens irradiated at 455 and 1136  C were embrittled even in room temperature tests This class of alloys is discussed further in Chapter 4.06, Radiation Effects in Refractory Metals and Alloys 1.04.9 Amorphous Metals Stable metallic glasses may be produced, commonly in intermetallic compounds Interest in the Irradiation dose required for amorphization (dpa) 120 Effect of Radiation on Strength and Ductility of Metals and Alloys 16.0 14.0 12.0 40Ar irradiation 10.0 8.0 6.0 4.0 2.0 0 100 200 300 400 500 600 Irradiation temperature (K) 700 Figure 30 Irradiation displacement level as a function of temperature for 0.9 MeV electron and 0.5–1.5 MeV Ar ion irradiation The family of curves is for several dpa rates of 1.04–1.83 mdpa sÀ1 Reproduced from Howe, L M.; Phillips, D.; Motta, A T.; Okamoto, P R Surface Coatings Tech 1994, 66, 411 irradiation properties of this class of materials resulted from preliminary tests that showed that these materials actually became more ductile upon irradiation.69 Other intermetallic compounds have been shown to become amorphous upon irradiation Although semiconductors such as Si and Ge are susceptible to amorphization under irradiation, the phenomenon is almost exclusively restricted to intermetallic compounds.70 To mention only a few, Zr3Al, Mo3Si, Nb3Ge, and Fe2Mo are compounds that have been studied in the amorphous state Results and a detailed review of mechanisms and theories of amorphization have been published by Motta.70 In simple terms, the lattice disruption and defect generation from irradiation disrupts long-range order in the system Thermal annealing competes with the disordering so that there is a critical temperature above which amorphization is not possible Figure 30 shows a plot of the irradiation exposure necessary for amorphization as a function of temperature for Zr3Fe.71 The critical temperatures and the necessary exposures are both functions of the material as well as the impinging particle Once formed, the amorphous phases are stable under irradiation, but the critical temperatures are typically lower than would be experienced for structural materials in nuclear systems They are of interest, however, because some intermetallic phases, such as Fe2Mo and Fe3B found in commercial alloys, become amorphous under irradiation.70,72 In the example of Zr3Fe, the critical temperature under argon ion irradiation is approximately 250  C, a temperature too low for most, but not all, reactors The intermetallic alloys that can be produced in the amorphous state before irradiation are of more interest as potential structural materials, although they remain in the category of research interest at the present time In addition to the increase in ductility upon irradiation, the absence of a crystalline structure with interacting dislocations was further incentive to investigate the irradiation properties of this class of materials Metallic glasses containing boron, such as Fe40Ni40B20 and (Mo.6Ru.4)82B18 are a few examples, with the former receiving the most attention in terms of mechanical properties.69,73–75 Amorphous alloys are complex systems where changes in free volume and segregation into clusters of differing composition result in changes in behavior as irradiation proceeds Investigation of the Fe–Ni–B alloy has shown that ductility first decreases and then increases with increasing fluence due to the competing effects of free volume and formation of regions of boron-depleted and boron-rich clusters.73 For sufficiently high fluences, the result is severe embrittlement In the case of alloys based on the intermetallic Zr3Al, very severe embrittlement upon irradiation is attributed to the formation of new amorphous phases.76 Even though a crystal structure is absent, the atoms may be dislodged from their locations, creating additional free volume Without the bonds present from a crystal lattice, the low binding energy results in high displacement levels for fluence levels that what would be considered low in crystalline alloys Fluence levels in the range of 1016–1021 n cmÀ2 have been investigated resulting in displacement levels exceeding 100 dpa However, simply having similar displacement levels does not permit a true comparison with crystalline materials Much research is necessary before this class of materials becomes of commercial importance 1.04.10 Conclusions The relationship between the irradiation-induced microstructure and tensile properties has been briefly presented using representative classes of alloys The austenitic stainless steels are an important class of alloys, and they are less complex than the martensitic steels In the unirradiated condition, the austenitic Effect of Radiation on Strength and Ductility of Metals and Alloys alloys are primarily hardened by dislocation reactions leading to conventional work hardening, and the martensitic steels are hardened by phase transformations requiring careful heat treatments The primary irradiation effects are similar, but they influence microstructure and, therefore, behavior in different ways Both types of alloys have important applications in the nuclear field Helium embrittlement might be the most important, considering the use of alloys in a neutron environment at high temperatures For the proper conditions, helium can nearly always cause catastrophic failure Repair welding of alloys with as little as 1–10 appm helium can lead to severe intergranular cracking The refractory metals are useful for space reactor application because of their liquid metal compatibility and their high-temperature strength Space reactors can lose heat only by thermal radiation, necessitating high temperatures However, this class of alloys is most susceptible to embrittlement by interstitial impurities, and synergism of impurities with irradiation-induced defects is an area that must be addressed further Amorphous alloys are a research curiosity in that they have interesting properties with respect to irradiation but little application at the present time Any new class of alloys must be understood before it can be engineered, so research is the essential beginning References 10 11 12 Simons, R L.; Hulbert, K A In Effects of Radiation on Materials; Garner, F A., Perrin, J S., Eds.; ASTM: Philadelphia, PA, 1985; Vol II, p 820 Bement, A L In Proceedings on the Strength of 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function of irradiation temperature Figure illustrates Effect of Radiation on Strength and Ductility of Metals and Alloys 10 1 Relative... 1. E + 04 Effect of Radiation on Strength and Ductility of Metals and Alloys 10 5 17 Irrad temp » Test temp Strain rate ~ ´ 10 -5 S 1 16 15 14 Test Temp (ЊC) 3 71 427 483 538 593 649 704 760 816 Symbol... in the Irradiation dose required for amorphization (dpa) 12 0 Effect of Radiation on Strength and Ductility of Metals and Alloys 16 .0 14 .0 12 .0 40Ar irradiation 10 .0 8.0 6.0 4.0 2.0 0 10 0 200 300