Mixed-spectrum Fission Reactor Research Articles

Mechanical properties of single-crystal tungsten irradiated in a mixed spectrum fission reactor

To collect data for fusion applications and understand the basic properties of tungsten, single-crystal tungsten was neutron irradiated in the mixed-spectrum High Flux Isotope Reactor at the Oak Ridge National Laboratory at temperatures of 90–830 °C to fast fluences of 0.01–9 × 1025 n/m2 (E > 0.1 MeV). For tensile tests at room temperature of the irradiated material in both orientations, <110> and <100> tensile orientation, initially there was strengthening that peaked at 0.02 dpa and was followed by progressive modulus of toughness reduction, approaching zero at higher doses. For all irradiation temperatures, in elevated temperature tensile tests there was a distinct transition from ductile-to-brittle behavior between 0.1 and 0.4 dpa, accompanied by an increase in indentation hardness. The ductile-to-brittle transition with increasing dose was particularly critical because it presented as a total loss in elongation and modulus of toughness. The critical transition dose was well below the dose (∼1 dpa) where irradiation-induced precipitates are visible in the TEM. The extent of Vickers microhardness increase was significant at higher doses and did not depend on irradiation temperature or crystal orientation, reaching 12.9 GPa after 2.8 dpa. The significant mechanical property degradation above 0.1 dpa is believed to have been caused by the accumulation of irradiation-induced clusters and eventually precipitates of the transmutation elements Re and Os.

Stored energy release in neutron irradiated silicon carbide

The purpose of this investigation is to experimentally quantify the stored energy release upon thermal annealing of previously irradiated high-purity silicon carbide (SiC.) Samples of highly-faulted polycrystalline CVD β-SiC and single crystal 6HSiC were irradiated in a mixed spectrum fission reactor near 60 °C in a fluence range from 5 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.05–20 dpa, in order to quantify the stored energy release and correlate the release to the observed microscopic swelling, lattice dilation, and microstructure as observed through TEM. Within the fluence of this study the crystalline material was observed to swell to a remarkable extent, achieving 8.13% dilation, and then cross a threshold dose for amorphization at approximately 1 × 1025 n/m2 (E > 0.1 MeV) Once amorphized the material attains an as-amorphized swelling of 11.7% at this irradiation condition. Coincident with the extraordinary swelling obtained for the crystalline SiC, an equally impressive stored energy release of greater than 2500 J/g at the critical threshold for amorphization is inferred. As expected, following amorphization the stored energy in the structure diminishes, measured to be approximately 590 J/g. Generally, the findings of stored energy are consistent with existing theory, though the amount of stored energy given the large observed crystalline strain is remarkable. The overall conclusion of this work finds comparable stored energy in SiC to that of nuclear graphite, and similar to graphite, a stored energy release in excess of its specific heat in some irradiation conditions.

Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa

Five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700 °C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Thermal resistivity increase as a result of irradiation was ∼20% higher when compared to CVD-SiC.

Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (∼90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage, providing insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

High-sensitivity quadrupole mass spectometry system for the determination of hydrogen in irradiated materials

High levels of helium and hydrogen are generated in metals in fusion reactors, and fusion simulations in mixed-spectrum fission reactors, spallation neutron sources, and high-energy charged particle environments. Although hydrogen generation rates are generally higher than those of helium, hydrogen is thought to quickly diffuse out of metals, especially at elevated temperatures. There appear to be some conditions, however, where significant hydrogen retention can occur. To assess this potential, a high-sensitivity analysis system has been developed for the measurement of hydrogen in small samples of irradiated materials. The system is based on a low-volume extraction furnace in combination with a quadrupole mass spectrometer, and has a detection limit of ∼1 appm for steel. Hydrogen levels have been measured in high-energy proton-irradiated tungsten and Inconel 718, in iron-based alloys and vanadium alloys from fusion materials programs, and in stainless steels and pure metals irradiated in light water reactors. Details of the system and typical results are presented.

Neutronics aspects of a DHCE experiment

The DHCE (dynamic helium charging experiment) irradiation experiment was conceived to simulate fusion-relevant helium production in a fission reactor irradiation. The main objective is to maintain the helium-to-dpa ratio at, roughly, the same level as expected in a fusion environment. The problem in fission reactor irradiation is that helium production is very low, because the fission neutrons, for basically all structural materials relevant for fusion applications, do not have enough energy to trigger the helium producing reactions. A DHCE experiment involves the decay of tritium to He-3 to produce the required helium during irradiation. This paper describes an analysis of the most important aspects of a DHCE experiment and compares different types of fission reactors and their suitability for performing such an experiment. It is concluded that DHCE experiments are feasible in a certain class of mixed-spectrum fission reactors, but a careful and detailed evaluation, for each facility and condition, must be performed to ensure the success of the experiment.

Microstructure of copper and nickel irradiated with fission neutrons near 230°C

The microstructures of pure copper and nickel specimens were examined by transmission electron microscopy (TEM) following irradiation at ∼ 230°C in a mixed spectrum fission reactor to damage levels between 0.01 and 0.25 displacements per atom (dpa). A high density of small stacking fault tetrahedra (SFT) and dislocation loops was observed in both materials, and a moderate density of small voids was observed in the irradiated copper specimens. From a comparison with published studies, the proportion of SFTs and dislocation loops in copper was observed to be nearly constant over a wide range of damage levels at temperatures between 200 and 250°C. This suggests that nucleation and growth of interstitial loops to visible sizes is rather difficult in copper in this temperature regime. On the other hand, the loop size and density in nickel increased steadily with increasing dose. The defect clusters in nickel formed a {001} planar wall pattern which became visible after damage levels of ∼ 0.1 dpa. Defect cluster patterning was not observed in the irradiated copper specimens.

Tensile Behavior of Neutron-Irradiated Martensitic Steels: A Review

Chromium-molybdenum martensitic (ferritic) steels such as 9 Cr-1 Mo-V-Nb and 12 Cr-1 Mo-V-W are candidates for fast reactor and fusion reactor applications. In a fast reactor, the effect of neutron irradiation is caused by displacement damage, that is, by the interstitials and vacancies that are created by the high-energy neutrons. Increases in strength occur for irradiation up to [approximately]450 C. This hardening is largely attributed to the dislocation loops that form from the agglomeration of the interstitials. Precipitates that form during irradiation can also contribute to the hardening. At higher temperatures, most of the displacement damage anneals out. Irradiation effects expected in the first wall of a fusion reactor differ from those in a fast reactor. In addition to displacement damage, large amounts of transmutation helium will also be produced. The simultaneous effects of displacement damage and helium can be simulated by irradiating nickel-doped ferritic steels in a mixed-spectrum fission reactor. Helium is produced by transmutation reactions between thermal neutrons and nickel, and displacement damage is formed by the fast neutrons of the spectrum. Results using this technique indicate that hardening occurs as in a fast reactor, but the helium causes a strength increase in addition to that caused bymore » displacement damage alone. This effect of helium could have a significant effect on other properties, especially toughness, and must be considered in the design of fusion reactors.« less

Structural materials for fusion reactors

The fusion reactor represents the most difficult challenge of any energy system from the viewpoint of developing materials which will allow fusion to be realized as an economic, safe and environmentally acceptable energy source. Substantial progress has been made over the past decade towards an understanding of the effects of the fusion environment on the properties of alloys, in identifying alloy systems which have the greatest potential for development as useful structural alloys, and in formulating and testing a metallurgical approach for developing irradiation resistant alloys with acceptable chemical and mechanical properties. Results to date give confidence that the long term objective – materials that will allow fusion to be an economical, safe and environmentally acceptable energy source – can be achieved. The next experimental fusion reactor, which will be either the International Thermonuclear Experimental Reactor (ITER), the Fusion Experimental Reactor (FER) or the Next European Torus (NET), will be the first device in which the effects of irradiation on the properties of materials will be a critical and limiting factor. At present, fission reactors are the only source of neutrons for the investigation of neutron irradiation damage and the development of radiation resistant materials. Spectral tailoring experiments in mixed spectrum fission reactors in which the neutron spectrum is adjusted to control the balance of transmutation produced helium and atomic displacements provide the closest approximation of fusion irradiation damage that can presently be achieved. These experiments suggest that for ITER conditions the most significant irradiation effects are in the areas of irradiation creep, loss of fracture toughness and possibly irradiation assisted stress corrosion cracking. A major expansion of the experimental programme will be required to develop a database for supporting the engineering design of ITER.

Radiation facilities for fusion reactor first wall and blanket structural materials development

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

Fatigue Behavior of Type 316 Stainless Steel Irradiated in a Mixed Spectrum Fission Reactor Forming Helium

In a Tokamak reactor that operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys, but none on irradiated materials containing irradiation-induced helium, which will be characteristic of fusion service. Specimens of 20% cold-worked Type 316 stainless steel were irradiated in the High Flux Isotope Reactor, which produces atomic displacement damage as well as helium through a two-step neutron absorption reaction with nickel. The specimens were irradiated at 430/sup 0/C to up to 15 dpa and 900 at. ppm helium. Following irradiation, specimens were tested in a vacuum at the irradiation temperature with total strain ranges from 0.30 to 2.0%. The irradiated specimens exhibited a reduction in fatigue life of a factor of 3 to 10 compared to unirradiated material. An endurance limit was observed at a total strain range of 0.3% for irradiated material.

Radiation effects in materials for fusion reactors

The 14-MeV neutrons produced in a fusion reactor result in different irradiation damage than the equivalent fluence in a fast breeder reactor, not only because of the higher defect generation rate, but because of the production of significant concentrations of helium and hydrogen. Although no fusion test reactor exists, the effects of combined displacement damage plus helium can be studied in mixed-spectrum fission reactors for alloys containing nickel (e.g., austenitic stainless steels). The presence of helium appears to modify vacancy and interstitial recombination such that microstructural development in alloys differs between the fusion and fission reactor environments. Since mechanical properties of alloys are related to the microstructure, the simultaneous production of helium and displacement damage impacts upon key design properties such as tensile, fatigue, creep, and crack growth. Through an understanding of the basic phenomena occurring during irradiation and the relationships between microstructure and properties, alloys can be tailored to minimize radiation-induced swelling and improve mechanical properties in fusion reactor service.

Microstructures developed in ‘simulated’ fusion irradiations

Microstructural information is reviewed for irradiation conditions which ‘simulate’ one or more salient features of the irradiation environment anticipated at first wall and blanket positions in a fusion reactor. The effects of high energy neutron damage from T[d,n], Be[d,n] and 16 MeV proton irradiations, and the effects of simultaneous atom displacement damage and helium and hydrogen production from mixed-spectrum fission reactor and dual-ion irradiation experiments are summarized for a variety of metals. It is shown that observations of microstructural sensitivity to damage rate, irradiation temperature and helium are providing important clues for developing damage resistant structural alloys for fusion reactor appiications.

Temperature and Fluence Limits for a Type 316 Stainless-Steel Controlled Thermonuclear Reactor First Wall

Results of a series of neutron irradiation experiments conducted on annealed and 20% cold-worked Type 316 stainless steel in a high-flux mixed-spectrum fission reactor to simulate a controlled ther...