Low Temperature Neutron Irradiation Research Articles

Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

Effect of low-temperature neutron irradiation on the corrosion resistance of 12Kh18N9T austenitic stainless steel

12Kh18N9T austenitic stainless steel was tested for the intercrystalline corrosion. The testing was carried out for samples irradiated in a BN-350 reactor with a dose of 27 dpa in a temperature range of 280–340°C. The defects formed during the corrosion tests were studied by means of optical microscopy. The phase composition of the irradiated steel was established using X-ray structural analysis, and its microstructure was investigated by electron microscopy.

Radiation defects and hydrogen in austenitic and austenitic-martensitic steels under low-temperature neutron irradiation

The experimental data concerning the effect of hydrogen (300 appm), radiogenic helium, and low-temperature neutron irradiation (77 K) on the properties of the promising austenitic 16Cr15Ni3Mo1Ti and austenitic-martensitic 16Cr9Ni3Mo steels have been reported. It has been found that hydrogen saturation causes an increase in the yield stress, with this increase being larger in the martensitic than in the austenitic phase. The yield stress of both steels increases substantially after exposure to fast neutrons. The variation of the yield stress of the two-phase steel and its phase components under low-temperature neutron irradiation has been estimated. The displacement cascades begin overlapping under irradiation at a fluence larger than 1.5 × 1018 cm−2.

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10 −5 to 0.28 dpa at ∼80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10 −5 to 1 × 10 −2 s −1. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at ∼0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at ∼0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.

Effect of neutron irradiation on the mechanical properties of 316L(N) steel for International Thermonuclear Experimental Reactor applications

One of the features peculiar to the application of 316L(N) steel for International Thermonuclear Experimental Reactor components is that it will be used in such components as the divertor and the first wall, mainly as part of the Cu–stainless steel bimetal composition. The main purpose of this investigation was to estimate the effect of heat treatment on the radiation hardening and embrittlement of 316L(N) steel. The report presents the results from an experimental investigation into the effect of low-temperature neutron irradiation on the radiation hardening and embrittlement in 316L(N) steel after different heat treatments. Specimens of 316L(N) steel were irradiated at T irr=150 and 80 °C in the dose range 0.001–2 dpa in the RBT-6 and SM-2 reactors. It is shown that heat treatment does not cause a significant effect on the radiation hardening and embrittlement of 316L(N) steel.

Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation

The thermal conductivity degradation due to low-temperature neutron irradiation is studied and quantified in terms of thermal resistance terms. Neutron irradiation is assumed to have no effect on umklapp scattering. A theoretical model is presented to quantify the relative phonon-scattering effectiveness of the three dominant defect types produced by neutron irradiation: point defects, dislocation loops and voids. Several commercial ceramics have been irradiated with fission reactor fast neutrons at low temperatures to produce defects. Materials include silicon carbide, sapphire, polycrystalline alumina, aluminum nitride, silicon nitride, beryllium oxide, and a carbon fiber composite. The neutron dose corresponded to 0.001 and 0.01 displacements per atom (dpa) for a ∼60°C irradiation and 0.01 and 0.1dpa for a ∼300°C irradiation. Substantial thermal conductivity degradation occurred in all of the materials except BeO following irradiation at 60°C to a dose of only 0.001dpa. The data are discussed in terms of the effective increase in thermal resistance caused by the different irradiation conditions. Evidence for significant point defect mobility during irradiation at 60 and 300°C was obtained for all of the ceramics. The thermal stability of the radiation defects was investigated by isochronal annealing up to 1050°C.

Plastic instability behavior of bcc and hcp metals after low temperature neutron irradiation

Plastic instability in uniaxial tensile deformation has been investigated for the body centered cubic (bcc) and hexagonal close packed (hcp) pure metals, V, Nb, Mo, and Zr, after low temperature (60–100 °C) neutron irradiation up to 0.7 dpa. Relatively ductile metals, V, Nb, and Zr, experienced uniform deformation prior to necking at low doses and prompt plastic instability at yield at high doses. Mo failed in a brittle mode within the elastic limit at doses above 0.0001 dpa. V showed a quasi-brittle failure at the highest dose of 0.69 dpa. In the ductile metals, plastic instability at yield occurred when the yield stress exceeded the plastic instability stress (PIS), which was nearly independent of dose. The PIS values for V, Nb, Mo, and Zr were about 390, 370, 510, and 170 MPa, respectively. The coincidence of plastic instability at yield and dislocation channeling cannot be generalized for all metallic materials.

Low-temperature low-dose neutron irradiation effects on beryllium

Mechanical property results are presented for high quality beryllium materials subjected to low-temperature, low-dose neutron irradiation in water-moderated reactors. Materials chosen were the S-65C ITER candidate material produced by Brush Wellman, Kawecki Berylco Industries P0 beryllium, and a high-purity zone refined beryllium. Mini-sheet tensile and thermal diffusivity specimens were irradiated in the temperature range of ∼100–300 °C with a fast ( E>0.1 MeV) neutron fluence of 0.05–1.0 × 10 25 n/m 2 in the high flux isotope reactor (HFIR) at the Oak Ridge National Laboratory and the high flux beam reactor (HFBR) at the Brookhaven National Laboratory. As expected from earlier work on beryllium, both materials underwent significant tensile embrittlement with corresponding reduction in ductility and increased strength. Both thermal diffusivity and volumetric expansion were measured and found to have negligible changes in this temperature and fluence range. Of significance from this work is that while both materials rapidly embrittle at these ITER relevant irradiation conditions, some ductility (1–2%) remains, which contrasts with a body of earlier work including recent work on the Brush-Wellman S-65C material irradiated to slightly higher neutron fluence.

The effect of tritium and neutron irradiation on the structure and properties of reactor steels and alloys

Studies performed under ISTC Project No. 019-94 provided a database on the effect of heavy hydrogen isotopes (concentration up to 0.03 at.%), radiogenic helium and neutron irradiation predominantly on the mechanical properties of FCC 16Cr15Ni3Mo1Ti radiation-resistant reactor steel, which is produced in Russia. It was shown that loading of the steel with tritium and deuterium up to 0.03 at.% and subsequent neutron irradiation at 77 or 320 K with fluence of 10 19–10 20 n/cm 2 led to an abrupt increase in strength characteristics, while a high plasticity (specific elongation δ=30–40%) was preserved even at extremely low temperatures. The microautoradiographic examination showed that strengthening was due to blocking of dislocations by tritium atmospheres, formed at dislocation cores. Comparative studies were performed on BCC alloys (Fe–13Cr, V–4Ti–4Cr, and V–10Ti–5Cr). Exposure to the tritium effect and low-temperature neutron irradiation was shown to lead to an almost complete degradation of their plasticity characteristics at 77 K.

The effect of low temperature neutron irradiation and annealing on the thermal conductivity of advanced carbon-based materials

Several carbon-based materials (carbon fibre composites NB 31, NS 31 and UAM-92, doped graphite RGTi-91), were irradiated at about 90 °C in the damage dose range 0.0021–0.13 dpa. Significant reduction of the thermal conductivity of all materials was observed (e.g. at damage dose of ∼0.13 dpa the thermal conductivity degraded up to level of ∼2–3% of the initial values). However, saturation of this effect was observed starting at a dose of ∼0.06 dpa. The effect of annealing at 250 and 350 °C on the recovery of thermal conductivity of NB 31 and NS 31 was studied and it was shown this annealing can significantly improve thermal conductivity (∼2.5–3 times). The data on the degradation of the thermal conductivity after additional irradiation after annealing is also reported.

Effects of low-temperature neutron irradiation on mechanical properties of vanadium-base alloys

Vanadium-base alloys were irradiated in three experiments to determine the effects of low-temperature neutron irradiation on their mechanical properties. The properties studied were tensile, Charpy impact and irradiation creep. The result of this study showed the alloys tested incurred significant hardening and embrittlement due to the irradiation. Heat-to-heat variations appeared to have little effect on hardening and embrittlement. Creep strains were measured from the specimens and a preliminary data set on creep strain rate was generated.

Critical issues and current status of vanadium alloys for fusion energy applications

Vanadium alloys are widely regarded as possessing desirable mechanical and physical properties for application as structural materials in fusion power systems. The bulk of the recent research on vanadium is focussed on ternaries containing 4–5% Cr and 4–10% Ti. The aim of this paper is to review significant results generated by the international research and development community on this alloy system and to highlight the critical issues that must be resolved before alloy development can proceed to the next stage. Recent progress on understanding the physical metallurgy, fabrication and joining behavior, and compatibility with hydrogen and oxygen containing environments of unirradiated vanadium alloys is discussed. The effect of low-temperature neutron irradiation on mechanical properties and their relationship to the observed microstructure are briefly summarized. Current efforts to characterize the high-temperature mechanical properties, develop constitutive equations describing flow and fracture, and understand and mitigate the effects of non-metallic impurities on properties are presented.

Microstructural evolution and hardening of neutron irradiated vanadium alloys at low temperatures in Japan Material Testing Reactor

In order to clarify the mechanism of the large increment of DBTT due to low temperature neutron irradiation, the mechanical properties and the microstructural evolution were investigated in vanadium binary alloys irradiated in JMTR. From a series of microhardness tests conducted on these alloys after irradiation at 90–200°C, the hardness increased with increasing irradiation temperature, except for V–5Ti and V–5Nb. However, the hardening decreased with increasing irradiation temperature from 350°C to 400°C. The dislocation loops were observed by TEM after irradiation from 90°C to 200°C, while no void was observed at these irradiation temperatures. Positron annihilation spectroscopy (PAS) exhibited the nucleation of vacancy clusters in undersized binary alloys. The dislocation loop density increased with increasing irradiation temperature. It is considered that the irradiation hardening is mainly caused by dislocation loops. It has also been suggested that interstitial oxygen promote dislocation loop nucleation since interstitial oxygen and solute atoms interacted strongly and bound at around 200°C leading to enhanced nucleation of interstitial loops.

Low Temperature Ductility Loss and Microstructural Evolution of Vanadium-Titanium-Chromium-Silicon Type Alloy after Helium Ion Implantation

Vanadium alloy as candidate structural material would be used at temperatures ranging above 600°C. However, the components might be exposed to low temperature and low neutron flux irradiation during transient operation such as start-up and shutdown period of the reactor. Helium generation also has a possibility to affect the irradiation behavior of the alloy at the low-temperature. Helium effect on loss of ductility at low-temperature neutron irradiation for vanadium alloy was pointed out from the results of tensile tests using various combinations of helium charging and neutron irradiation. It is possible so far that pre-implantation at ambient temperature increases radiation hardening of the vanadium alloy after irradiation at 400°C. From the viewpoint of low-temperature and low-flux irradiation behavior of vanadium alloy, to clarify the helium effect on microstructure evolution, especially small complexes of interstitial impurity and irradiation defect is needed.

The effect of tritium and low-temperature neutron irradiation at 77 K on the structure and mechanical properties of reactor steels

The effect of tritium and deuterium (∼0.03 at%) and also low-temperature (77 K) neutron irradiation on the physical and mechanical properties of stainless steels type Cr16Ni15Mo3Ti1 has been studied. Using an autoradiographic method, it was shown that tritium located mainly at dislocations, pinning the dislocations and slightly strengthening the steel. The resistance to small plastic deformations (the yield stress σ 0.2) was found to increase considerably when the fluence of fast neutrons was as high as 1.5 × 10 19 n/cm 2. This is attributed to a large number of low-temperature clusters, which appear during the cascade-forming irradiation. A much lesser increase in the strength σ B may probably be explained by dissolution of the strengthening clusters when they interact with dislocations as the test samples are being tensioned at 77 K. The degradation of the plastic characteristics was shown to be an order of magnitude larger in BCC steels and V alloys than in reactor materials possessing an FCC lattice.

The influence of low-temperature neutron irradiation on physicomechanical properties of new reactor alloys saturated with tritium

The influence of tritium (∼0.03 at.%) and low-temperature (77 K) neutron irradiation (fluence 1.5×10 18–15×10 18 n cm −2) on changes in physical and mechanical properties of stainless steels at 77 K was studied. Strength of the steels was found to be considerably enhanced by increasing the fluence of fast neutrons (power dependence). This increase in strength is explained by the appearance of a large number of clusters formed at low temperatures (hence termed ‘low-temperature clusters’) under cascade-forming irradiation. The deterioration of plasticity of BCC steels and alloys was shown to be an order of magnitude faster than that for reactor materials having an FCC lattice.

Effects of Neutron Irradiation on Tensile Properties of V-Ti-Cr-Si Type Alloys with Pre-Existing Helium

AbstractHelium effect on loss of uniform elongation after low-temperature neutron irradiation of the V-Ti-Cr-Si-Al-Y alloy was studied. Helium implantation to about 30 atomic ppm was carried out before neutron irradiation to 50 dpa at 406°C. The yield stress of the irradiated specimen with helium pre-implantation was slightly smaller than that of irradiated specimens without helium. The uniform elongation and the increase in yield stress of the irradiated specimens were not affected by helium pre-implantation at ambient temperature. It might be possible that the helium effect appears after lower temperature irradiation such as 300°C or lower, which corresponds to the temperature range where the loss of uniform elongation of the alloy is particularly pronounced.

Radiation defect formation and evolution in C0.03Cr20Ni16Mn6 steel under low-temperature neutron irradiation and their effect on physical and mechanical properties of the steel

Radiation defect formation in austenitic steel under low-temperature irradiation was analysed. Certain approximations were made to calculate the neutron interaction with a multicomponent system, such as austenitic steel, and atomic displacements in cascade regions. These approximations were used to calculate a threshold energy for stable Frenkel pair formation. The level of gas transmutant generation by neutron irradiation was estimated. A qualitative interpretation of the microstructural changes in austenite stainless steel under low-temperature neutron irradiation was formulated. The radiation defect types dominating physical and mechanical properties were considered. A model to describe an influence of these defects on steel strength properties, plasticity, and thermal expansion characteristics was suggested. C0.03Cr20Ni16Mn6 steel property variation under neutron irradiation to the fluence of 1 × 10 23n/m 2 ( E ≥ 0.1 MeV) at 77 K was calculated within the mechanisms selected. The results obtained were compared to data on low-dose mechanical properties, Young's modulus and the thermal linear expansion coefficient of C0.03Cr20Ni16Mn6 steel specimens, irradiated to the fluence 1 × 10 22 n/m 2, 5 × 10 22 n/m 2, and 1 × 10 23n/m 2 at 77 K and tested at the same temperature as well as after heating to room temperature.

Effects of low temperature neutron irradiation on deformation behavior of austenitic stainless steels

Two experiments have been conducted to quantify the effects of neutron irradiation on the deformation and fracture behavior of solution annealed austenitic stainless steels irradiated to doses ranging from 3 to 19 dpa at temperatures from 60 to 400°C. For all alloys, yield strength increases rapidly with dose in the 60–300°C regime. Radiation hardening is accompanied by changes in the flow properties with the appearance of an initial yield drop and a significant reduction in strain hardening capacity. The magnitude of the changes is dependent upon both neutron dose and irradiation temperature, with reductions in strain hardening capacity occurring most rapidly in the range 250–350°C. It is shown that for neutron doses up to about 3 dpa, although the changes in deformation mode reduce the fracture thoughness, the toughness remains satisfactorily high.

Microstructure of V-4Cr-4Ti Following Low Temperature Neutron Irradiation

AbstractThe V-4Cr-4Ti alloy displays excellent mechanical properties, including a ductile-to-brittle transition temperature (DBTT) below -200°C in the unirradiated condition. Specimens of this alloy were fission neutron-irradiated in the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) to a dose of 0.4 dpa at temperatures from 100 to 275°C. Mechanical testing showed significant irradiation hardening in the alloy which increased with increasing irradiation temperature. Charpy impact testing also showed a dramatic increase in the DBTT on the order of 200 to 350°c. The mechanical property changes are correlated with preliminary results from the transmission electron microscopy (TEM) analysis of the defect microstructure resulting from the low-dose neutron irradiations. The TEM analysis of the irradiated material showed a nearly constant defect density of ∼1.6 × 1023 m−3, with an average defect diameter of slightly greater than 3 nm.