Materials Open Test Assembly Research Articles

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility–Materials Open Test Assembly (FFTF–MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412°C–109dpa, chromium enrichments – consistent with the α′ phase – appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412°C–109dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

SANS and TEM of ferritic–martensitic steel T91 irradiated in FFTF up to 184dpa at 413°C

Ferritic–martensitic steel T91 was previously irradiated in the Materials Open Test Assembly (MOTA) program of the Fast Flux Test Reactor Facility (FFTF) at 413 °C up to 184 dpa. The microstructure was analyzed by small angle neutron scattering (SANS) and transmission electron microscopy (TEM). Both SANS and TEM revealed a large fraction of voids with an average size of 29–32 nm leading to a calculated void swelling of 1.2–1.6% based on the volume fraction of the voids in the sample. SANS gave no indication of second phase particles having formed under irradiation in the material. Using TEM, one zone was found where a few G-phase particles were analyzed. Quantities were however too low to state reliable particle densities. No alpha prime (α′) or Laves phase were observed in any of the investigated zones.

Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation

The in-reactor creep rupture tests of 11Cr–0.5Mo–2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using materials open test assembly in the fast flux test facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The results of post-irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

High-dose neutron irradiation of MgAl 2O 4 spinel: effects of post-irradiation thermal annealing on EPR and optical absorption

Electron paramagnetic resonance (EPR) and optical absorption spectra were measured during thermal annealing of stoichiometric MgAl 2O 4 spinel that was previously irradiated in the Materials Open Test Assembly in the Fast Flux Test Facility (FFTF/MOTA) at ≈680 K to ≈50 dpa. Both F and F + centres are to persist up to very high temperatures (over 1000 K) suggesting the operation of an annealing mechanism controlled by the thermal stability of extended defects. Using X-ray irradiation following the different annealing steps it was shown that an optical absorption band at 37 000 cm −1 is related to a sharp EPR band at g = 2.0005 and that the defect causing these effects is the F + centre.

Impact of transmutation issues on interpretation of data obtained from fast reactor irradiation experiments

The subject of fission–fusion correlation is usually cast in terms of reactor-to-reactor differences, but recently the fusion community has become aware of the impact of differences within a given surrogate facility, especially in constant time experiments when different dose levels are attained in different positions of one reactor. For some materials, it is not safe to assume that in-reactor spectral variations are small and of no consequence. This point is illustrated using calculations for fusion-relevant materials that were irradiated in the Fast Flux Test Facility–Materials Open Test Assembly (FFTF–MOTA) over a wide range of in-core and out-of-core positions spanning more than two orders of magnitude in dpa rate. It is shown that although both the neutron spectrum and flux changes, the spectral effectiveness factor, dpa/10 22 n/cm 2 ( E>0.1 MeV), remains remarkably constant over this range. The transmutation rate per dpa varies strongly with reactor position, however.

Microstructural development of neutron irradiated W–Re alloys

Tungsten (W) alloys are candidate materials to be used as high-heat-flux materials in fusion reactors. In our previous work, W–26 wt% Re showed drastic hardening and embrittlement after the neutron irradiation. In this study, to clarify the irradiation hardening and embrittlement behavior of W–26 wt% Re, from the viewpoint of microstructural development, the microstructure observation of the neutron irradiated W–26 wt% Re was carried out using transmission electron microscope (TEM). The specimens were irradiated at the materials open test assembly of the fast flux test facility (FFTF/MOTA-2A cycle 11) up to ∼1×10 27 n/m 2 , ( E n >0.1 MeV). The irradiation temperatures were 646, 679, 792, 873 and 1073 K. In all neutron irradiated W–26 wt% Re samples, sigma-phase precipitates and chi-phase precipitates were observed, while in the thermally aged specimen, only sigma-phase precipitates were observed. Irradiation effects on microstructural development are discussed.

The dependence of helium generation rate on nickel content of Fe–Cr–Ni alloys irradiated to high dpa levels in EBR-II

Fusion-relevant helium-effects experiments conducted on austenitic steels in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF) fast reactor had to recognize the contributions of both the high neutron energy (n,α) reactions and that of the 58Ni(n,γ) 59Ni(n,α) reaction sequence with low energy neutrons. An experiment conducted in the harder neutron spectra found within the core of Experimental Breeder Reactor-II (EBR-II) has shown that the helium in this reactor was generated almost exclusively from the interaction of high energy neutrons with the natural isotopes of nickel. There was very little contribution from 59Ni. The helium production was found to scale directly with the nickel content over the range 25–75% Ni. Even at very high neutron exposures, the helium production in such reactors can be predicted within 5% accuracy on the basis of high energy reactions, as demonstrated by an experiment conducted on three Fe–15Cr–Ni ternary alloys irradiated to doses of 75–131 dpa in EBR-II.

Radiation hardening of V–C, V–O, V–N alloys neutron-irradiated to high fluences

Vanadium has a large affinity for interstitial impurities such as C, N and O. Mechanical properties and irradiation performance of vanadium alloys are affected by the impurities. Radiation hardening and defect microstructures of vanadium alloys doped with relatively large amounts of these interstitial elements were studied. Neutron irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-1F) to 47.9 dpa at temperatures of 679, 793 and 873 K. Irradiation hardening decreased with increasing irradiation temperature. Increase in hardness for the V–C alloy was relatively greater after irradiation at the low temperatures. Decorated dislocations and voids were observed depending on the alloying elements. The factors for irradiation hardening were different for each interstitial element in the alloys irradiated at 873 K to 47.9 dpa.

Microstructures of deformed VTiCrSi type alloys after neutron irradiation

The alloy of V5Ti5Cr1SiAl,Y (nominal composition, weight percentage) was developed to improve oxidation properties and high temperature strength, and has been studied as one of the candidates for fusion applications. This alloy showed low swelling properties and enough tensile ductility after neutron irradiation to high fluence levels. The dislocation microstructures after tensile deformation and defect microstructures in the neutron-irradiated alloy to high fluences were studied. Irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-2A) at 406°C to 46 dpa and the deformation microstructures were examined by transmission electron microscopy. Slip dislocations were developed inhomogeneously in the specimen deformed at ambient temperature after neutron irradiation. Dislocation loops contributed mainly to hardening of the alloy after irradiation; however, cavities and radiation-induced precipitates did not so much.

Precipitation Behavior of Irradiated Reduced-Activation Ferritic Steels.

In order to study the effects of small additional elements on precipitation behavior in reduced-activation ferritic steels under neutron irradiation, analytical transmission electron microscopy was used to examine the microstructure and the precipitation of the extraction replica of several reduced activation ferritic/martensitic steels which have different contents of small additional elements after 60 displacements per atom (dpa) irradiation at 693, 698 and 733K in the Fast Flux Test Facility (FFTF)/Materials Open Test Assembly (MOTA). All of reduced-activation ferritic/martensitic steels were found to have a good phase stability after irradiation. Micro-voids were observed in both materials of Fe-9Cr-2W with or without boron, the density of micro-voids in the steel with boron is larger than that without boron, and the mean size of micro-voids is smaller than that without boron. However void swelling was less than 1%. Most of the precipitates in the irradiated specimens were found to be M23C6 which consists of mainly Cr. The remainder of the precipitates were found to be Ta-rich M6C. Laves phase was observed only at 733K. Several precipitates which were Ti-rich including Si and W were also observed at grain boundary in Ti addition steels at 733K irradiation. Several Y2O3 particles were observed in an yttrium containing alloy. No precipitation including Al was observed in an Al containing alloy. Ti addition decreased precipitation of Ta-rich M6C in 9Cr and 12Cr steels in this irradiation condition.

Defect formation in austenitic stainless steels during shutdown procedure of FFTF

Abstract Defect formation during a shutdown procedure of a fast reactor has been investigated. A Fe-16Cr-17Ni alloy and two phosphorus-containing alloys, Fe-16Cr-17Ni-0.024P and Fe-16Cr-17Ni-0.1P, were irradiated in the Fast Flux Test Facility (FFTF) using Materials Open Test Assembly (MOTA) during Cycle 10 and Cycle 11. In the case of Cycle 10, in which the reactor was shut down by a controlled procedure, interstitial-type dislocation loops were formed in addition to voids. In the case of Cycle 11, in which the reactor scrammed, stacking fault tetrahedra (SFT) were formed in addition to the voids. Standing on the rate theory of defect processes under irradiation, it was predicted that the interstitial loops were nucleated by the short neutron exposure during a shutdown process, but the SFT were formed as agglomerates of vacancies and their small clusters which were accumulated in the matrix as a result of balance of interstitials and vacancies under irradiation.

Dimensional Changes and Microstructure of In-pile Crept Ferritic-Martensitic Steel

Pressurized tube specimens of Japanese Ferritic-Martensitic Steel (JFMS) were irradiated in Fast Flux Test Facility (FFTF) Materials Open Test Assembly (MOTA) to 37.5 dpa at 680 and 793 K. Diametral creep strain following irradiation at 680 K was 0.11% even at zero hoop stress and slightly increased with increasing the hoop stress. The creep strain following irradiation at 793 K increased from zero to 0.23% with the increase in hoop stress from zero to 86 MPa. As for the void swelling, nothing was recognized. The martensitic phase in JFMS was stable after the irradiation at 680 K, but this was not the case at 793 K where a considerable recovery of dislocation structures was found

Irradiation Creep Behavior of Low Activation Steels in FFTF/MOTA

Irradiation creep behavior of low activation steels, developed as structural materials for fusion reactors, was investigated. The objective of this study is to provide a fundamental understanding of the irradiation creep mechanism based on microstructural evolution under fast neutron irradiation. Pressurized tube creep specimens fabricated from tube segments were irradiated in the Fast Flux Test Facility (FFTF), Materials Open Test Assembly (MOTA) during FFTF Cycles 11 and 12. This paper provides the first creep results obtained after FFTF cycle-11 irradiation. (2.25-3)Cr-(1-2)W bainitic steels and 12Cr-2W ferritic/martensitic steels showed equivalent or superior creep resistance to a modified 316 stainless steel, known as Japanese Prime Candidate Alloy (JPCA), under fast neutron irradiation up to 600 o C

Microstructural changes in Fe-10Cr-2Mo steel by neutron or charged particle irradiation

An Fe-10Cr-2Mo type ferritic/martensitic dual phase steel. JFMS (Japanese ferrite/martensite steel), has been extensively studied in the Japanese fusion reactor materials research programs as a reference ferritic alloy. This work is intended to characterize the effects of neutron or charged particle irradiation to high doses on microstructural evolution and swelling of JFMS. Fast neutron irradiation was performed in an open packet in MOTA (materials open test assembly) during FFTF (fast flux test facility) cycle 10, at 643, 679, 793 and 873 K up to the maximum dose of about 40 dpa. For charged particle irradiation, 4 MeV nickel ions were injected up to 200 dpa, with or without helium ion implantation, at 723 K, at the high-fluence irradiation test facility (HIT facility), University of Tokyo. No void swelling was found at any temperature in the neutron irradiated specimens. X-ray energy dispersive spectroscopy revealed that precipitate structure is quite stable below 679 K, while it changed significantly at 793 K. Cavities were formed by charged particle irradiation only after exposure of more than 100 dpa. Helium implantation resulted in very small volume changes in the martensitic phase even at 200 dpa and demonstrated the excellent swelling resistance of JFMS.

Helium generation rates in isotopically tailored FeCrNi alloys irradiated in FFTF/MOTA

Three FeCrNi alloys have been doped with 0.4% 59Ni for side-by-side irradiations of doped and undoped materials in order to determine the effects of fusion-relevant levels of helium production on microstructural development and mechanical properties. The alloys were irradiated in three successive cycles of the Materials Open Test Assembly (MOTA) located in the Fast Flux Test Facility (FFTF). Following irradiation, helium levels were measured by isotope dilution mass spectrometry. The highest level of helium achieved in doped alloys was 172 appm at 9.1 dpa for a helium(appm)-to-dpa ratio of 18.9. The overall pattern of predicted helium generation rates in doped and undoped alloys is in good agreement with the helium measurements.

The influence of isotopically controlled boron addition on void swelling of nickel irradiated in FFTF

Pure nickel specimens doped with several levels of boron-10 ( 10B) and natural boron ( nB) were irradiated in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA) for the purpose of examining the applicability of boron addition to the study of helium production effects. It was shown that the boron addition enhances void nucleation as its transmulation effect and suppresses as its chemical effect. The former effect is dominant at higher temperatures.

FFTF as an irradiation test bed for fusion materials and components

The relatively large irradiation volume, instrumentation capabilities, and fast neutron flux associated with the Fast Flux Test Facility (FFTF) make this reactor a useful test bed for fusion materials and components irradiations. Significant fusion materials irradiations are presently being performed in the Materials Open Test Assembly (MOTA) in FFTF. The MOTA is providing a controlled temperature and high neutron flux environment for materials such as low activation alloys, copper alloys, ceramic insulators, and high heat-flux components. Conceptual designs utilizing the versatile MOTA irradiation vehicle have been developed to investigate irradiation effects on the mechanical and tritium breeding behaviors of solid breeder materials. More aggressive conceptual designs have also been developed to irradiate solid breeder blanket submodules in the FFTF. These specific component test designs are presented and their potential roles in the development of fusion technology discussed.

Design of a single variable helium effects experiment for irradiation in FFTF using alloys enriched in nickel-59

Nickel enriched in Ni-59 was extracted from the fragments of a fracture toughness specimen of Inconel ∗ ∗ Inconel is a registered trade mark of the International Nickel Company. 600 irradiated in the Engineering Test Reactor (ETR). The nickel contained 2.0% Ni-59. Three heats of austenitic steel doped with Ni-59 were prepared and inserted in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The experiment is single variable in helium effects because chemically identical alloys without Ni-59 are being irradiated side by side with the doped material. The doped alloys will produce 10–100 times more helium than the control alloys. The materials include ternary and quaternary alloys in the form of transmission electron microscope (TEM) discs and miniature tensile specimens. The helium-to-dpa (displacements per atom) ratio will be in the range 5 to 35 and will be nearly constant throughout the irradiation. The exposures will range from 0.25 to 50 dpa over the duration of the experiment. The irradiation temperatures cover the range of 360–600°C.

Instrumented in-reactor test capabilities in FFTF

Fusion materials research and development requires high neutron fluxes coupled with a large irradiation volume. Both of these needs can be provided by the Fast Flux Test Facility. The Materials Open Test Assembly, the vehicle used for materials irradiations in the FFTF, is described.