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.     

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.     

钼铼合金在空间核电源中的应用性能研究进展

钼（Mo）中加入铼（Re）可显著改善钼的低温脆性进而提高其加工性能及焊接性能,提高强度的同时仍保持良好的塑性。Re元素含量为14%左右时,Mo-Re合金延伸率接近40%,加工性能最好,而同时存在一定的Re元素固溶强化作用。在1550 K以下温度,Mo-Re合金与UO₂的相容性较好。在1 300 K以下时,Mo-Re合金与UN的相容性较好。在1800 K以下时,Mo-Re合金与碱金属Li、Na、K的相容性均较好。钼铼合金与核燃料及碱金属冷却剂均具有良好的相容性,且Re元素是一种较好的谱移吸收体材料,可有效降低反应堆临界事故风险。钼铼合金是空间核电源中最佳反应堆芯结构材料。本文对钼铼合金的研究状况进行总结,为国内相关空间核反应堆电源系统设计选材和研究提供参考。     

Mechanical properties and microstructure in neutron-irradiated nickel-based alloys and stainless steels for supercritical water-cooled-reactor fuel cladding

ABSTRACT

To investigate the irradiation behavior of mechanical properties and microstructural changes of commercial Ni-based alloys and improved stainless steels, a neutron-irradiation experiment was performed at the Joyo reactor, and post-irradiation examinations with tensile tests and TEM observations were carried out. The room-temperature tensile tests showed that all specimens that were irradiated at 485°C exhibited significant hardening and ductile behavior, especially in alloy 625. The irradiation hardening of all specimens irradiated at 668°C was less than that of specimens irradiated at 485°C. The fine-grained stainless steel, T3 and the Zr-added stainless steels, H1 and H2 showed good mechanical-property performance with keeping ductility after neutron irradiation. Most alloys and steels showed ductile behavior on the fracture surface except for alloy 625 specimen. The TEM observations showed that a high density of tangled dislocations and irradiation-induced defect clusters formed in the stainless steels and Ni-based alloys irradiated at 485°C. At 668°C, the material microstructures coarsened and their dislocation density decreased significantly. Long rod-like precipitates of Zr(Cr, Fe) compounds formed in the H1 and H2 steels that were modified with Zr. The yield stress drop of T3 steel in tensile stress was observed and is caused by grain-size coarsening at an irradiation of 668°C.     

Post irradiation plastic properties of F82H derived from the instrumented tensile tests

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.     

Microstructural and mechanical property changes with aging of Mo-41Re and Mo-47.5Re alloys

The changes in microstructure and mechanical properties of Mo-41Re and Mo-47.5Re alloys were investigated following 1100 h thermal aging at 1098, 1248 and 1398 K. The electrical resistivity, hardness and tensile properties of the alloys were measured both before and after aging, along with the alloy microstructures though investigation by optical and electron microscopy techniques. The Mo-41Re alloy retained a single-phase solid solution microstructure following 1100 h aging at all temperatures, exhibiting no signs of precipitation, despite measurable changes in resistivity and hardness in the 1098 K aged material. Annealing Mo-47.5Re for 1 h at 1773 K resulted in a two-phase αMo + σ structure, with subsequent aging at 1398 K producing a further precipitation of the σ phase along the grain boundaries. This resulted in increases in resistivity, hardness and tensile strength with a corresponding reduction in ductility. Aging Mo-47.5Re at 1098 and 1248 K led to the development of the χ phase along grain boundaries, resulting in decreased resistivity and increased hardness and tensile strength while showing no loss in ductility relative to the as-annealed material.     

The effect of irradiation and post-irradiation annealing on the yield stress and microstructure of vanadium-carbon alloys

Vanadium and vanadium-carbon alloys containing 0.15, 0.3 and 1 at% carbon were irradiated in JMTR with fast neutrons (E_n > 1 MeV) at 773 K to a dose of 5 × 10²⁴n/m². Tensile test and microstructural observations were carried out after irradiation and post-irradiation annealing. All of the. specimens showed radiation hardening. The irradiation produced voids, dislocations and radiation-induced quasi-carbides, which were formed by the agglomeration of vacancies and carbon atoms. The radiation anneal hardening in the alloys occurred at 873 K. The void number densities in the alloys had a peak at 873 K while the quasi-carbides decomposed at the same temperature. Therefore, invisible voids existing in the as-irradiated condition would grow by absorbing the vacancies, which were released in the process of decomposition of the quasicarbides during annealing, and the increase of the visible voids would effectively contribute to the radiation anneal hardening of these alloys.     

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.     

Changes in hardness and elasticity of a Ti6Al4V alloy under helium irradiation

Ti₆Al₄V specimens have been irradiated at different temperatures with 200 keV He ions. Microhardness and elastic modulus of the unirradiated and irradiated specimens were measured by means of the nano-indentation technique and analyzed using the Oliver-Pharr method. The indentation depth of all samples is 700 nm, which is comparable in magnitude to the ion range. The subsurface structure of the Ti₆Al₄V specimens was investigated by the X-ray diffraction technique. The measurements indicate that the microhardness increased with the irradiation temperature from room temperature to 600 °C while the elastic modulus almost monotonically decreased. The Irradiation at 700 °C, however, caused softening and slight increase of the elastic modulus within the surface layer of the specimens. The hardening and reduction of the elastic modulus of the Ti₆Al₄V alloy under irradiation conditions used in this study is tentatively explained by a model based on the presence of point defects and dispersed obstacles of β-precipitates. The softening and slight increase of elastic modulus of helium-irradiated Ti₆Al₄V at 700 °C might be related to the coarsening of β-precipitates and formation of the hybrid γ-TiH phase in α-phase.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

Proton irradiation emulation of PWR neutron damage microstructures in solution annealed 304 and cold-worked 316 stainless steels

Solution annealed (SA) 304 and cold-worked (CW) 316 austenitic stainless steels were pre-implanted with helium and were irradiated with protons in order to study the potential effects of helium, irradiation dose, and irradiation temperature on microstructural evolution, especially void swelling, with relevance to the behavior of austenitic core internals in pressurized water reactors (PWRs). These steels were irradiated with 1 MeV protons to doses between 1 and 10 dpa at 300 °C both with or without 15 appm helium pre-implanted at ∼100 °C. They were also irradiated at 340 °C, but only after 15 appm helium pre-implantation. Small heterogeneously distributed voids were observed in both alloys irradiated at 300 °C, but only after helium pre-implantation. The pre-implanted steels irradiated at 340 °C exhibited homogenous void formation, suggesting effects of both helium and irradiation temperature on void nucleation. Voids developed sooner in the SA304 alloy than CW316 alloy at 300 and 340 °C, consistent with the behavior observed at higher temperatures (>370 °C) for similar steels irradiated in the EBR-II fast reactor. The development of the Frank loop microstructure was similar in both alloys, and was only marginally affected by pre-implanted helium. Loop densities were insensitive to dose and irradiation temperature, and were decreased by helium; loop sizes increased with dose up to about 5.5 dpa and were not affected by the pre-implanted helium. Comparison with microstructures produced by neutron irradiation suggests that this method of helium pre-implantation and proton irradiation emulates neutron irradiation under PWR conditions.     

High temperature irradiation effects in selected Generation IV structural alloys

In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 °C and to irradiation doses in the range 1.2-1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 °C; some hardening occurred at 660 °C as well, but the 800H showed extremely low tensile elongations when tested at 700 °C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 °C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.     

The influence of neutron irradiation in FFTF on the microstructural and microchemical development of Mo-41Re at 470-730 °C

Specimens of Mo-41 wt% Re irradiated in the fast flux test facility (FFTF) experience significant and non-monotonic changes in density that arise first from radiation-induced segregation, leading to non-equilibrium phase separation, and second by progressive transmutation of Re to Os. As a consequence the density of Mo-41Re initially decreases and then increases thereafter. Beginning as a single-phase solid solution of Re and Mo, irradiation of Mo-41 wt% Re over a range of temperatures (470-730 °C) to 28-96 dpa produces a high density of thin platelets of a hexagonal close-packed (hcp) phase identified as a solid solution of Re, Os and possibly a small amount of Mo. These hcp precipitates are thought to form in the alloy matrix as a consequence of strong radiation-induced segregation to Frank loops. Grain boundaries also segregate Re to form the hcp phase, but the precipitates are much bigger and more equiaxed in shape. Although not formed at lower dose, continued irradiation at 730 °C leads to the co-formation of late-forming chi-phase, an equilibrium phase that then competes with the preexisting hcp phase for rhenium.     

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.     

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel

Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.     

Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.     

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C.During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.     

Defect structure of irradiated PH13-8Mo steel

PH13-8Mo bolts, which are considered for use in the ITER reactor, were irradiated up to doses of 0.5, 1 and 2 dpa. The microstructure was investigated with transmission electron microscopy and its evolution is discussed with reference to the mechanical properties. PH13-8Mo is a precipitation hardened martensitic steel, but a large amount of austenite has been observed as well. The precipitation hardening results from the formation of small coherent NiAl precipitates in the martensite phase. Their size, size distribution and density are found to be unaffected by neutron irradiation. The dislocations in the martensite phase are mainly a/2〈1 1 1〉 type screw dislocations, whereas in the austenite phase mainly a/2〈1 1 0〉 type screw dislocations are present. The line dislocation structure did not change during irradiation, but small irradiation induced defects were observed. Using the Orowan model, it is argued that the latter are responsible for the irradiation hardening.     

Influence of thermal and radiation effects on microstructural and mechanical properties of Nb-1Zr

The microstructural changes and corresponding effects on mechanical properties, electrical resistivity and density of Nb-1Zr were examined following neutron irradiation up to 1.8 dpa at temperatures of 1073, 1223 and 1373 K and compared with material thermally aged for similar exposure times of ∼1100 h. Thermally driven changes in the development of intragranular and grain boundary precipitate phases showed a greater influence on mechanical and physical properties compared to irradiation-induced defects for the examined conditions. Initial formation of the zirconium oxide precipitates was identified as cubic structured plates following a Baker-Nutting orientation relationship to the β-Nb matrix, with particles developing a monoclinic structure on further growth. Tensile properties of the Nb-1Zr samples showed increased strength and reduced elongation following aging and irradiation below 1373 K, with the largest tensile and hardness increases following aging at 1098 K. Tensile properties at 1373 K for the aged and irradiated samples were similar to that of the as-annealed material. Total elongation was lower in the aged material due to a strain hardening response, rather than a weak strain softening observed in the irradiated materials due in part to an irregular distribution of the precipitates in the irradiated materials. Though intergranular fracture surfaces were observed on the 1248 K aged tensile specimens, the aged and irradiated material showed uniform elongations >3% and total elongation >12% for all conditions tested. Cavity formation was observed in material irradiated to 0.9 dpa at 1073 and 1223 K. However, since void densities were estimated to be below 3 × 10¹⁷ m⁻³ these voids contributed little to either mechanical strengthening of the material or measured density changes.