Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.     

Tensile behaviour of 9Cr-1Mo tempered martensitic steels irradiated up to 20 dpa in a spallation environment

Tensile specimens of 9Cr-1Mo (EM10) and mod 9Cr-1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Tensile properties of ferritic/martensitic steels irradiated in STIP-I

Specimens of ferritic/martensitic (FM) steels T91, F82H, Optimax-A and the electron beam weld (EBW) of F82H were irradiated in the Swiss spallation neutron source (SINQ) Target-3 in a temperature range of 90-370 °C to displacement doses between 3 and 12 dpa. Tensile tests were performed at room temperature and the irradiation temperatures. The tensile test results demonstrated that the irradiation hardening increased with dose up to about 10 dpa. Meanwhile, the uniform elongation decreased to less than 1%, while the total elongation remained greater than 5%, except for an F82H specimen of 9.8 dpa tested at room temperature, which failed in elastic deformation regime. At higher doses of 11-12 dpa, the ductility of some specimens recovered, which could be due to the annealing effect of a short period of high temperature excursion. The results do not show significant differences in tensile properties for the different FM steels in the present irradiation conditions.     

Tensile properties of the ferritic martensitic steel F82H after irradiation in a spallation target

Ferritic/Martensitic (FM) steel, F82H, was irradiated up to a displacement dose of 20 dpa (displacement per atom) at temperatures ranging from 510 to 1075 K in the third experiment of the SINQ Target Irradiation Program (STIP-III). Tensile testing was performed at 295 and 723 K. The tensile test results demonstrate that not only the specimen irradiated in the low temperature regime (<∼675 K) but also those irradiated at elevated temperatures ?710 K show significant hardening effect. After annealing at 873 K for 2 h the irradiated specimens still persist great hardening, which is usually not observed in FM steels after neutron irradiation at low temperatures and annealing at 873 K. The hardening observed in the specimens is believed to be due to the high-density He-bubbles formed in the specimens.     

Proton irradiation study of GFR candidate ceramics

This work investigated the microstructural response of SiC, ZrC and ZrN irradiated with 2.6 MeV protons at 800 °C to a fluence of 2.75 × 10¹⁹ protons/cm², corresponding to 0.71-1.8 displacement per atom (dpa), depending on the material. The change of lattice constant evaluated using HOLZ patterns is not observed. In comparison to Kr ion irradiation at 800 °C to 10 dpa from the previous studies, the proton irradiated ZrC and ZrN at 1.8 dpa show less irradiation damage to the lattice structure. The proton irradiated ZrC exhibits faulted loops which are not observed in the Kr ion irradiated sample. ZrN shows the least microstructural change from proton irradiation. The microstructure of 6H-SiC irradiated to 0.71 dpa consists of black dot defects at high density.     

Proton irradiation effect on microstructure, strain localization and iodine-induced stress corrosion cracking in Zircaloy-4

The radiation-induced microstructure, strain localization, and iodine-induced stress corrosion cracking (I-SCC) behaviour of recrystallized Zircaloy-4 proton-irradiated to 2 dpa at 305 °C was examined. <a> type dislocation loops having 1/3〈1 1  0〉 Burgers vector and a mean diameter and density of, respectively, 10 nm and 17 × 10²¹ m⁻³ were observed while no Zr(Fe,Cr)₂ precipitates amorphization or Fe redistribution were detected after irradiation. After transverse tensile testing to 0.5% macroscopic plastic strain at room temperature, almost exclusively basal channels were imaged. Statistical Schmid factor analysis shows that irradiation leads to a change in slip system activation from prismatic to basal due to a higher increase of critical resolved shear stresses for prismatic slip systems than for basal slip system. Finite element calculations suggest that dislocation channeling occurs in the irradiated proton layer at an equivalent stress close to 70% of the yield stress of the irradiated material, i.e. while the irradiated layer is still in the elastic regime for a 0.5% applied macroscopic plastic strain. Comparative constant elongation rate tensile tests performed at a strain rate of 10⁻⁵ s⁻¹ in iodized methanol solutions at room temperature on specimens both unirradiated and proton-irradiated to 2 dpa demonstrated a detrimental effect of irradiation on I-SCC.     

Mechanical properties of irradiated ODS-EUROFER and nanocluster strengthened 14YWT

Irradiations to 1.5 dpa at 300-750 °C were conducted to investigate the changes in mechanical properties of an advanced nanocluster strengthened ferritic alloy, designated 14YWT, and an oxide dispersion strengthened ferritic alloy ODS-EUROFER. Two non-dispersion strengthened variants, 14WT and EUROFER 97, were also irradiated and tested. Tensile results show 14YWT has very high tensile strengths and experienced some radiation-induced hardening, with an increase in room temperature yield strength of 125 MPa after irradiation, while results for ODS-EUROFER show a 275 MPa increase following irradiation. Master curve fracture toughness analysis show 14YWT has a cryogenic T_o reference temperatures before and after irradiation of about −188 and −176 °C, respectively, and upper-shelf K_JIc values between 175 and 225 MPa√m. The favorable fracture toughness properties and resistance to radiation-induced changes in mechanical properties observed for 14YWT are attributed to a fine grain structure and high number density of Y-Ti-O nanoclusters.     

Tensile tests and metallography of brazed AISI 316L specimens after irradiation

Stainless steel type 316L tensile specimens were vacuum brazed with three kinds of alloys: BNi-5, BNi-6, and BNi-7. The specimens were irradiated up to 0.7 dpa at 353 K in the High Flux Reactor at JRC Petten, the Netherlands. Tensile tests were performed at a constant displacement rate of 10⁻³ s⁻¹ at room temperature in the ECN hot cell facility. BNi-5 brazed specimens showed ductile behaviour. Necking and fractures were localized in the plate material. BNi-6 and BNi-7 brazed specimens failed brittle in the brazed zone. This was preceded by uniform deformation of the plate material. Tensile test results of irradiated specimens showed higher stresses due to radiation hardening and a reduction of the elongation of the plate material compared to the reference. SEM examination of the irradiated BNi-6 and BNi-7 fracture surfaces showed nonmetallic phases. These phases were not found in the reference specimens.     

Neutron-induced swelling and embrittlement of pure iron and pure nickel irradiated in the BN-350 and BOR-60 fast reactors

Pure iron and nickel were irradiated in the range of 2-15 × 10⁻⁷ dpa/s at 345-650 °C to very high neutron exposures in two fast reactors, BOR-60 and BN-350, to study void swelling and changes in mechanical properties of these two metals. Both nickel and iron swell in this temperature range with the maximum swelling rate at ∼500 °C in nickel, but possibly at ?350 °C for iron. It also appears that the swelling rate in nickel and possibly in iron may be dependent on the dpa rate, increasing with decreasing dpa rate. The evolution of mechanical properties of the two metals is quite different. The differences reflect the fact that b.c.c. iron is subject to a low-temperature embrittlement arising from a shift in ductile-brittle transition temperature, while f.c.c. nickel is not. Nickel, however, exhibits high temperature embrittlement, thought to arise from the collection of transmutant helium gas at the grain boundaries. Iron is not strongly affected by transmutation since it generates much less helium during equivalent irradiation.     

Deformation and fracture properties in neutron irradiated pure Mo and Mo alloys

The effect of neutron irradiation on the mechanical properties of select molybdenum materials, unalloyed low carbon arc-cast (LCAC) Mo, Mo-0.5% Ti-0.1% Zr (TZM) alloy, and oxide dispersion-strengthened (ODS) Mo alloy, was characterized by analyzing the temperature dependence of mechanical properties. This study assembles the tensile test data obtained through multiple irradiation and post-irradiation experiments, in which tensile specimens were irradiated up to 13.1 dpa at 80-1000 °C and tested at −194 to 1000 °C. Irradiation at 80-609 °C increased yield stress significantly, up to 170%, while the increase of yield stress after irradiation at 784-936 °C was not significant. The plastic instability stress was strongly dependent on test temperature but was nearly independent of irradiation dose and temperature. The true fracture stress showed weak dependences on test temperature, irradiation dose and temperature when ductile failure occurred. Among the test materials the stress-relieved ODS material in the longitudinal direction (ODS-LSR) displayed the highest resistance to irradiation embrittlement due to its relatively high fracture stress. The critical temperature for shear failure (CTSF) was defined and evaluated for the test materials and the CTSF values were compared with the ductile-to-brittle transition temperatures (DBTT) based on ductility data.     

Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

Precipitates induced in austenitic stainless steel by nitrogen injection

Samples of solution-annealed Type 316 stainless steel were irradiated at 803 K with 1.1 MeV nitrogen ions up to 60 dpa at peak and the precipitates induced by the irradiation were examined by electron microscopy. The ratio of injected nitrogen concentration to dpa is estimated to be about 0.3 at%/dpa in an observed area. An ordered γ'-phase of nominally Ni₃Si is identified to form in the sample irradiated to 20 dpa; it shows superlattice spots in the electron diffraction pattern and enrichment of Si and Ni at the precipitate in EDS analysis. In the sample irradiated to 42 dpa, platelet precipitates of 18 nm in average diameter were observed with the number density of 7.4 × 10²¹/m³. The precipitate grows to an average diameter of 28 nm with little change in number density during irradiation from 42 to 60 dpa. The precipitate formed in the sample irradiated to 60 dpa was confirmed to be in the CrN phase with electron diffraction patterns and the spacing of moiré fringes.     

The role of niobium carbide in radiation induced segregation behaviour of type 347 austenitic stainless steel

The effect of niobium carbide precipitates on radiation induced segregation (RIS) behaviour in type 347 stainless steel was investigated. The material in the as-received condition was irradiated using double-loop 4.8 MeV protons at 300 °C for 0.43 dpa (displacement per atom). The RIS in the proton irradiated specimen was characterized using double-loop electrochemical potentiokinetic reactivation (DL-EPR) test followed by atomic force microscopic examination. The nature of variation of DL-EPR values with the depth matched with the variation of the calculated irradiation damage (dpa) with the depth. The attack on grain boundaries during EPR tests was negligible indicating absence of chromium depletion zones. The interface between niobium carbide and the matrix acts as a sink for point defects generated during irradiation and this had reduced point defect flux toward grain boundaries. The attack was noticed at a few large cluster of niobium carbide after the DL-EPR test at the depth of maximum attack for the irradiated specimen. Pit-like features were not observed within the matrix indicating the absence of chromium depletion regions within the matrix.     

Tensile properties of EC316LN irradiated in SINQ to 20 dpa

The wide application of 316-type austenitic stainless steels in existing spallation targets requires a comprehensive understanding of their behavior in spallation irradiation environments. In the present study, EC316LN specimens were irradiated in SINQ targets to doses between 3 and 17.3 dpa at temperatures between about 80 °C and 390 °C. Tensile tests were conducted at room and irradiation temperatures. The results demonstrate that the irradiation induced significant hardening and embrittlement in the specimens. The irradiation hardening and embrittlement effects show a trend of saturation at doses above about 10 dpa. Although the ductility was greatly reduced, all specimens broke with strong necking, which indicates a ductile fracture mode.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Radiation stability of ZrN under 2.6 MeV proton irradiation

Zirconium nitride is a promising alternative material for the use as an inert matrix for transuranic fuel, but the knowledge of the radiation tolerance of ZrN is very limited. We have studied the radiation stability of ZrN using a 2.6 MeV proton beam at 800 °C. The irradiated microstructure and hardening were investigated and compared with annealed samples. A high density of nano-sized defects was observed in samples irradiated to doses of 0.35 and 0.75 dpa. Some defects were identified as vacancy-type pyramidal dislocation loops using lattice resolution imaging and Fourier-filter image processing. A very slight lattice expansion was noted for the sample with a dose of 0.75 dpa. Hardening effects were found for samples irradiated to both 0.35 and 0.75 dpa using Knoop indentation.     

Influence of post irradiation annealing on the mechanical properties and defect structure of AISI 304 steel

The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24 dpa in the decommissioned Chooz A reactor. The material was investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as γ′-precipitates (Ni₃Si). Annealing at 400 °C had almost no effect on the radiation induced defects, but annealing at 500 °C resulted in the immediate unfaulting of the Frank loops. As to the mechanical properties, annealing at 400 °C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests under pressurised water reactor conditions, was significantly reduced. Annealing at 500 °C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties.