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.     

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.     

Mechanical properties of Eurofer 97 in Pb-16Li and irradiation effect

To be used in a fusion reactor, structural materials, and in particular steels, has to be selected and optimised in their composition to achieve a reduction in the long-term radioactive waste. A reduction in the long-term radioactive inventory could be reached substituting elements like molybdenum, niobium and nickel with other ones like tantalum and tungsten which have the same functions as alloying elements and, if irradiated, do not produce long lived radioisotopes. The martensitic steel belonging to the family of 8-9% Cr Eurofer 97 is considered the reference structural steel for fusion application. However, only few information are available about its mechanical properties in the liquid eutectic alloy Pb-16%Li. Particularly, the problem of liquid metal embrittlement (LME) has not been studied in detail and the effect of neutron irradiation on LME has not been investigated at all so far. This work presents the results obtained irradiating tensile specimens of Eurofer 97 up to 5.9 dpa in lead lithium. Tensile tests of samples have been performed out of pile in the same alloy at the same temperature at which irradiation was carried out.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

Microstructural and mechanical property changes in the Ta-base T-111 alloy following thermal aging

The microstructural changes occurring in the Ta-base T-111 (Ta-8W-2Hf) alloy during 1100 h thermal aging at 1098, 1248 and 1398 K under inert atmosphere and the influence on mechanical properties are reported. Electrical resistivity, hardness and tensile properties are compared between the as-annealed and aged conditions. Microstructural evaluations were performed by optical, scanning electron microscopy and transmission electron microscopy. An increase in the amount of grain boundary precipitation with increasing aging temperature was found to decrease the electrical resistivity and material strength. Precipitation at the grain boundaries was found to be a mixture of monoclinic and cubic structures, suggesting the development of mixed Hf oxides, carbides and nitrides. Precipitate development caused pronounced embrittlement of the alloy following aging at 1398 K.     

Microstructural development and radiation hardening of neutron irradiated Mo-Re alloys

Stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys with Re contents of 2, 4, 5, 10, 13 and 41 wt% were neutron irradiated up to 20 dpa at temperatures from 681 to 1072 K. On microstructural observation, sigma phase and chi phase precipitates were found in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimens, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874 K or below. From these results, the authors discuss about the relation between microstructure development and radiation hardening and embrittlement, and propose the optimum Re content and thermal treatment for Mo-Re alloys to be used under irradiation conditions.     

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.     

Basic investigation for life assessment technology of modified 9Cr–1Mo steel

In order to develop life assessment techniques for aged components made of modified 9Cr–1Mo steel, specimens were artificially deteriorated by aging, creep and fatigue tests at elevated temperatures, and associated changes in the microstructure and mechanical properties were examined. It was observed that aging resulted in formation of Laves phase causing a decrease in toughness. The creep damage in base metal could be correlated with decrease in hardness, while creep damage in weldments could be correlated with the area fraction and density of creep voids. Creep rupture in weldments occurred in the fine-grained heat affected zone by the formation and growth of creep voids. The fatigue damage in base metal correlated to the maximum length of a crack among micro-cracks initiated during fatigue cycles.     

Potential of direct metal deposition technology for manufacturing thick functionally graded coatings and parts for reactors components

Direct metal deposition (DMD) is an automated 3D deposition process arising from laser cladding technology with co-axial powder injection to refine or refurbish parts. Recently DMD has been extended to manufacture large-size near-net-shape components. When applied for manufacturing new parts (or their refinement), DMD can provide tailored thermal properties, high corrosion resistance, tailored tribology, multifunctional performance and cost savings due to smart material combinations. In repair (refurbishment) operations, DMD can be applied for parts with a wide variety of geometries and sizes. In contrast to the current tool repair techniques such as tungsten inert gas (TIG), metal inert gas (MIG) and plasma welding, laser cladding technology by DMD offers a well-controlled heat-treated zone due to the high energy density of the laser beam. In addition, this technology may be used for preventative maintenance and design changes/up-grading. One of the advantages of DMD is the possibility to build functionally graded coatings (from 1 mm thickness and higher) and 3D multi-material objects (for example, 100 mm-sized monolithic rectangular) in a single-step manufacturing cycle by using up to 4-channel powder feeder. Approved materials are: Fe (including stainless steel), Ni and Co alloys, (Cu,Ni 10%), WC compounds, TiC compounds. The developed coatings/parts are characterized by low porosity (<1%), fine microstructure, and their microhardness is close to the benchmark value of wrought alloys after thermal treatment (Co-based alloy Stellite, Inox 316L, stainless steel 17-4PH). The intended applications concern cooling elements with complex geometry, friction joints under high temperature and load, light-weight mechanical support structures, hermetic joints, tubes with complex geometry, and tailored inside and outside surface properties, etc.     

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.     

Creep rupture of a 9Cr1MoNbV steel at 500 °C: Base metal and welded joint

ASME Grade 91 steel base metal and a similar weld were tested under creep at 500 °C for rupture time up to 18,000 h. Creep failure of cross-weld specimens occurs in the weld metal at this temperature. No significant microstructural changes were observed after creep. Analysis of creep deformation of smooth creep bars, welded joints and slightly notched bars indicated an apparent creep stress exponent of 19. For the creep conditions considered, failure of the material can be explained by the viscoplastic instability of the specimens without significant damage development. This allowed to develop a simple analysis for time to failure prediction.     

Toughness behaviour of the steel X 20 CrMoV 12 1 — Results of optimizing the base- and weld metal

The 12% Cr steels are frequently used in German power plants for tubings, pipes, rotors, and blades. The maximum operating temperature is limited by their creep strength properties to about 550°C. There are applications at even higher temperatures. Sufficient materials toughness is required for the base metal and weld metal to withstand sudden load changes. This is of special interest for use in nuclear power plants. Under operating conditions at elevated temperatures microstructural changes occur which greatly influence the toughness properties of both base metal and weld metal. This paper presents the results of ageing treatments at 550°C, carried out with a 12% Cr steel (DIN X 20 CrMoV 12 1) specifically optimized for toughness. The decrease in toughness is already evident at ageing times as low as 1000 h for conventional and optimized material. This drop in toughness is tentatively explained by differences in grain sizes and carbide content (M₂₃C₆ carbides). Detailed investigations indicate that additional carbide precipitation may significantly contribute to the decrease in toughness.     

Characterization of creep-fatigue in ferritic 9Cr-1Mo-V-Nb steel using ultrasonic velocity

The microstructural evolution of ferritic 9Cr-1Mo-V-Nb steel, subjected to creep-fatigue at 550 °C, was evaluated nondestructively by measuring the ultrasonic velocity. The ultrasonic velocity was strongly depended on the microstructural changes during creep-fatigue. The variation in the ultrasonic velocity with the fatigue life fraction exhibited three regions. In the first region (within 0.2 N_f), a significant increase in the velocity was observed, followed by a slight increase between the fatigue life fractions of 0.2 N_f and 0.8 N_f and a decrease in the final region. The change of the ultrasonic velocity during creep-fatigue was interpreted in relation to the microstructural properties. This study proposes an ultrasonic nondestructive evaluation method of quantifying the level of damage and microstructural change during the creep-fatigue of ferritic 9Cr-1Mo-V-Nb steel.     

Core materials development for the fuel cycle R&D program

The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels’ fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI).To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (∼400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions.     

Low temperature age hardening in U-13 at.% Nb: An assessment of chemical redistribution mechanisms

Low temperature aging (<350 °C) of U-13 at.% Nb martensite results in increased strength levels accompanied by significant ductility loss. To determine the decomposition mechanism(s) responsible for these mechanical property changes, atom probe tomography was used to examine the niobium and impurity distributions after aging at 200 or 300 °C for times ranging from 2 h to 70 days. No patterns of niobium or impurity atoms were observed that would indicate segregation to the martensitic twin interfaces, making this hardening mechanism unlikely. Phase separation into roughly equiaxed regions of high and low niobium concentration was clearly observed after aging at 300 °C for 70 days. However, only subtle niobium concentration changes were observed after aging at 200 °C relative to the as-quenched condition, indicating that conventional phase separation is an unlikely explanation for the dramatic mechanical property changes at 200 °C. Therefore, consideration of aging mechanisms other than segregation and phase separation may be warranted.     

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.     

Optimization of resistance spot welding on the assembly of refractory alloy 50Mo-50Re thin sheet

Resistance spot welding (RSW) was employed to pre-join refractory alloy 50Mo-50Re (wt%) sheet with a 0.127 mm gage. Five important welding parameters (hold time, electrode, ramp time, weld current and electrode force) were adjusted in an attempt to optimize the welding quality. It was found that increasing the hold time from 50 ms to 999 ms improved the weld strength. Use of rod-shaped electrodes produced symmetric nugget and enhanced the weld strength. Use of a ramp time of 8 ms minimized electrode sticking and molten metal expulsion. The weld strength continuously increased with increasing the weld current up to 1100 A, but the probabilities of occurrence of electrode sticking and molten metal expulsion were also increased. Electrode force was increased from 4.44 N to 17.8 N, in order to reduce the inconsistency of the welding quality. Welding defects including porosities, columnar grains and composition segregation were also studied.     

Alloys based on niobium and their properties

Conclusions A number of refractory alloys based on niobium were developed, and their properties were investigated; the complex of these properties permits the recommendation of these alloys for use in various fields of technology. The alloys are characterized by good technological properties, ensuring the production of various objects: sheets, bars, tubes, and wires, under industrial conditions. The most heat-resistant alloys of the RN brand, possessing substantial short-term and long-term strength, are the alloys of tungsten, molybdenum, and zirconium (RN-6, RN-5). The alloys can be subjected to dispersion hardening.It was found that the alloy RN-6, produced by the method of centrifugal casting, possesses better mechanical properties than the alloys produced by arc and electron beam methods, as a result of its fine-grained structure. The alloying of niobium alloys with titanium, instead of zirconium, causes a smaller increase in the strength, especially at increased temperatures; however, their plasticity increases somewhat. Alloys of the RN type possess high corrosion resistance in lithium at 1000°C, as well as in solutions of hydrochloric and sulfuric acids.Translated from Atomnaya Énergiya, Vol. 23, No. 1, pp. 32–37, July, 1967.     

Fracture toughness assessment of ferritic-martensitic steel in liquid lead-bismuth eutectic

The presence of micro-cracks at the surface of a ferritic-martensitic steel is known to favour its embrittlement by liquid metals and thus decrease the mechanical properties of the structural materials. Unfortunately, conventional fracture mechanics methods cannot be applied to tests in liquid metal environment due to the opaque and conducting nature of the LBE. Therefore new methods based on the normalization technique for assessment of plain strain fracture toughness in LBE were examined. This paper discusses the assessment of the plain strain fracture toughness of T91 steel in liquid lead bismuth environment at 473 K, tested at a displacement rate of 0.25 mm min⁻¹ and makes the comparison with results obtained in air at the same temperature and displacement rate. Although there is a decrease of the fracture toughness by 20-30% when tested in LBE, the toughness of the T91 steel remains sufficient under the tested conditions.