Unirradiated Specimens Research Articles

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 ∼0.7displacements per atom (dpa) at 1073, 1223, and 1373K and ∼1.4 dpa at 1073K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1×10−3s−1 in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100h 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 1100K, 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.

A STUDY OF STRESS RELAXATION RATE IN UN-IRRADIATED AND NEUTRON-IRRADIATED STAINLESS STEEL

Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2–12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

Irradiation energy dependence of stress relaxation rate and activation volume in polycrystalline nickel

Stress relaxation in un-irradiated and electron beam irradiated polycrystalline nickel (99.99%) was studied at room temperature. The specimens were irradiated by high-energy electrons with energies in the range 8–18 MeV for 12 min at 300 K. The tensile tests of both un-irradiated and irradiated specimens were carried out at room temperature using a universal testing machine. During deformation, straining was frequently interrupted to observe stress relaxation at fixed loads. Measurements were made over the entire stress-strain curves up to fracture. The stress-relaxation was found to be logarithmic with time and its rate was found to decrease with an increase in the energy of incident electrons. The activation volume Vσ was observed to decrease with an increase in stress levels (σ0) both in un-irradiated and irradiated specimens, however Vσ values were higher in irradiated specimens than those of un-irradiated specimens. The intrinsic height of energy barrier (Uo) to the movement of dislocations increased with an increase in incident electron energy. The results are analysed in terms of a single barrier model of stress relaxation proposed by Feltham.

Study on irradiation-induced magnetic transition in FeRh alloys by means of Fe K-edge XMCD spectroscopy

We have studied the effects of swift heavy ion irradiation on the magnetic properties of Fe–50at.%Rh intermetallic compound by using Fe K-edge X-ray magnetic circular dichroism (XMCD) at 20K. For an unirradiated specimen, no signal of XMCD spectrum has been observed because Fe–50at.%Rh intermetallic compound is intrinsically anti-ferromagnetic below room temperature. For the specimens irradiated with 120–200MeV Ni, Kr, Xe or Au ions, we have found XMCD spectra which are characterized by a dispersion-type profile with a positive peak and a negative peak near Fe K-edge. The intensity of XMCD spectra remarkably depends on ion-fluence and ion-mass. The experimental result implies that the ferromagnetic state is induced by the swift heavy ion irradiation. The change in magnetic state of Fe element is discussed in terms of energy deposition through the elastic and electronic processes.

Role of grain boundary engineering in the SCC behavior of ferritic–martensitic alloy HT-9

This paper focuses on the role of grain boundary engineering (GBE) in stress corrosion cracking (SCC) of ferritic–martensitic (F–M) alloy HT-9in supercritical water (SCW) at 400°C and 500°C. Constant extension rate tensile (CERT) tests were conducted on HT-9 in as-received (AR) and coincident site lattice enhanced (CSLE) condition. Both unirradiated and irradiated specimens (irradiated with 2 MeV protons at 400°C and 500°C to a dose of 7dpa) were tested. Ferritic–martensitic steel HT-9 exhibited intergranular stress corrosion cracking when subjected to CERT tests in an environment of supercritical water at 400°C and 500°C and also in an inert environment of argon at 500°C. CSL-enhancement reduces grain boundary carbide coarsening and cracking susceptibility in both the unirradiated and irradiated condition. Irradiation enhanced coarsening of grain boundary carbides and cracking susceptibility of HT-9 for both the AR and CSLE conditions. Intergranular (IG) cracking of HT-9 results likely from fracture of IG carbides and seems consistent with the mechanism that coarser carbides worsen cracking susceptibility. Oxidation in combination with wedging stresses is the likely cause of the observed environmental enhancement of high temperature IG cracking in HT-9.

Surface modification of silicone medical materials by plasma-based ion implantation

Silicone (polydimethylsiloxane) sheets and tubes for medical use were irradiated with inert gas ions using plasma-based ion implantation (PBII). The affinity of the surface with tissue examined by an animal test was improved by the irradiation at optimal conditions. The cell attachment percentage increased at an applied voltage of less than −7.5kV; however, it decreased at higher voltage. The specimens irradiated at higher voltages were more hydrophobic than unirradiated specimens. The surface became rough with increasing voltage and textures, and small domains appeared. This effect was caused by different etching speeds in the amorphous and crystalline areas.

Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions

Laboratory-scale experiments were performed with Zircaloy-4 fuel cladding, irradiated to 39 and 44 GWd/t at a PWR, for evaluating high burn-up fuel rod behavior in a loss-of-coolant accident. Short test rods, fabricated with the cladding were heated, isothermally oxidized at 1,303 to 1,451 K in steam flow, and quenched with water flooding. Two cladding specimens, oxidized to about 26 to 29% ECR, were fractured during the quench, while four cladding specimens, oxidized to about 16 to 22% ECR, survived the quench. Threshold of fracture is comparable with that of unirradiated cladding specimens which are hydrided to the same level. Accordingly, in the burnup level of the present study, differences were not significant between irradiated and unirradiated cladding specimens in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases.

Changes in hardness and elasticity of a Ti 6Al 4V alloy under helium irradiation

Ti 6Al 4V 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 Ti 6Al 4V 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 Ti 6Al 4V 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 Ti 6Al 4V 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 Cooling History on Cladding Ductility under LOCA Conditions

In a Loss-of-Coolant Accident of LWRs, heated and oxidized fuel cladding is slowly cooled and then quenched by re-flooding water. Effects of cooling rate during the slow-cooling process and quench temperature were investigated on post-quench ductility of the oxidized cladding. Unirradiated Zircaloy-4 cladding specimens were oxidized in steam at 1,373 or 1,473 K, cooled at a rate from 2 to 7 K/s, and finally quenched from temperatures ranging 1,073 to 1,373 K. Post-quench ductility was evaluated by ring-compression test, and microscopic properties were examined by metallurgical examination, hardness test, and oxygen analysis. Morphology of oxygen-rich α phase in the metallic prior-β phase layer changed depending on the cooling history. Area fraction of α phase region in the cross section obviously increased and post-quench ductility reduced with decrease in the quenching temperature. Since α phase region with low ductility can be preferential path for crack propagation, increase in area fraction of α phase region possibly decreases resistance for fracture, which results in ductility reduction. On the other hand, the area fraction was nearly constant irrespective of the rate of the slow cooling, and consequently, effect of the cooling rate on the post-quench ductility was negligible.

Hydrogen Accumulation in Surface of Perfluorosulfonic Acid Membranes after γ-Ray Irradiation using Elastic Recoil Detection Techniques

Electrical conductivity measurements and elastic recoil detection (ERD) analysis were carried our on perfluorosulfonic acid membranes that were irradiated by 60Co γ-ray. Maximum electrical conductivity of irradiated specimen was obtained at 40–60 °C, though that of unirradiated specimen increased with increasing temperature. And 4000-fold higher conductivity was obtained in irradiated membrane at 40 °C under dry condition. ERD analysis indicated that hydrogen concentration of the irradiated membrane is 1.5-fold higher than that of the unirradiated specimen. It is considered that the γ-ray generates a lot of site that can trap hydrogen and the trapped hydrogen cannot easily move. As the results, irradiated membrane retains more hydrogen than unirradiated one and the electrical conductivity of irradiated membrane is higher than that of unirradiated membrane in any conditions. γ-Ray irradiation is effective for improving the electrical conductivity of proton-conducting membranes.

Mechanical properties and the evolution of matrix molecules in PTFE upon irradiation with MeV alpha particles

The morphology, chemical composition, and mechanical properties in the surface region of α-irradiated polytetrafluoroethylene (PTFE) have been examined and compared to unirradiated specimens. Samples were irradiated with 5.5 MeV 4He 2+ ions from a tandem accelerator to doses between 1 × 10 6 and 5 × 10 10 Rad. Static time-of-flight secondary ion mass spectrometry (ToF-SIMS), using a 20 keV C 60 + source, was employed to probe chemical changes as a function of α dose. Chemical images and high resolution spectra were collected and analyzed to reveal the effects of α particle radiation on the chemical structure. Residual gas analysis (RGA) was utilized to monitor the evolution of volatile species during vacuum irradiation of the samples. Scanning electron microscopy (SEM) was used to observe the morphological variation of samples with increasing α particle dose, and nanoindentation was engaged to determine the hardness and elastic modulus as a function of α dose. The data show that PTFE nominally retains its innate chemical structure and morphology at α doses <10 9 Rad. At α doses ≥10 9 Rad the polymer matrix experiences increased chemical degradation and morphological roughening which are accompanied by increased hardness and declining elasticity. At α doses >10 10 Rad the polymer matrix suffers severe chemical degradation and material loss. Chemical degradation is observed in ToF-SIMS by detection of ions that are indicative of fragmentation, unsaturation, and functionalization of molecules in the PTFE matrix. The mass spectra also expose the subtle trends of crosslinking within the α-irradiated polymer matrix. ToF-SIMS images support the assertion that chemical degradation is the result of α particle irradiation and show morphological roughening of the sample with increased α dose. High resolution SEM images more clearly illustrate the morphological roughening and the mass loss that accompanies high doses of α particles. RGA confirms the supposition that the outcome of chemical degradation in the PTFE matrix with continuing irradiation is evolution of volatile species resulting in morphological roughening and mass loss. Finally, we reveal and discuss relationships between chemical structure and mechanical properties such as hardness and elastic modulus.

Effect of low energy helium irradiation on mechanical properties of 304 stainless steel

Mechanical properties of low energy He irradiated 304SS have been investigated by means of tensile test and nano-identation hardness tests. Specimen size is 2.8 mm × 5.4 mm × 0.1 mm in gauge and 12.4 mm in total length. Specimens are irradiated with 8 keV helium ions at R.T., 573 K and 873 K up to the fluence of 3 × 10 21 He/m 2. Tensile tests are performed on unirradiated and irradiated specimens at a strain rate of 3.33 × 10 −3 s −1. In every irradiation condition, one specimen is tested to failure and the other two specimens are strained plastically to 10% and 20%. Formation of blisters with a diameter of about 500 nm is observed on the surface of tensile-tested specimens after R.T. irradiation. Some exfoliation of the blisters is also observed. When the irradiation temperature increases, a large number of cracks with a size of about several micrometers are formed on the surface together with exfoliation during the tensile test. The yield stress of specimens irradiated at R.T. and 573 K increases about 10% in comparison with that of the unirradiated specimen. The result of nano-indentation tests indicate that hardness near the surface increases depending on the irradiation temperature.

Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions

Laboratory-scale experiments were performed with Zircaloy-4 fuel cladding, irradiated to 39 and 44 GWd/t at a PWR, for evaluating high burn-up fuel rod behavior in a loss-of-coolant accident. Short test rods, fabricated with the cladding were heated, isothermally oxidized at 1,303 to 1,451 K in steam flow, and quenched with water flooding. Two cladding specimens, oxidized to about 26 to 29% ECR, were fractured during the quench, while four cladding specimens, oxidized to about 16 to 22% ECR, survived the quench. Threshold of fracture is comparable with that of unirradiated cladding specimens which are hydrided to the same level. Accordingly, in the burnup level of the present study, differences were not significant between irradiated and unirradiated cladding specimens in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases.

Effects of irradiation on the microstructure and mechanical properties of nanostructured materials

Nanostructured materials should present a good resistance to irradiation because the large volume fraction of grain boundaries can be an important sink for radiation-induced defects. The objective of the present study is to experimentally investigate the irradiation impact on the microstructure and mechanical properties in nanostructured materials. Nickel and Cu-0.5Al2O3 specimens were synthesized by electro deposition (ED) and severe plastic deformation (SPD). Mean grain size of the unirradiated specimen is about 30 nm for the ED Ni and about 115 nm for the SPD Ni. 590 MeV proton irradiation and 840 keV nickel ion irradiation were conducted at room temperature. Vickers hardness measurements and transmission electron microscope observation were performed to examine the impact of irradiation on nanocrystalline materials. It appears that the irradiation induced microstructure in Ni and in Cu-0.5Al2O3, which leads to hardening, consists exclusively of stacking fault tetrahedra. Their density appears much lower than in the case of coarser grained material. These results, experimentally showing the resistance of nanostructured material to radiation damage, are presented here.

Bend-fatigue properties of 590 MeV proton irradiated JPCA and 316F SS

A beam window of a spallation target will be subjected to proton/neutron irradiation, pressure wave and thermal stresses accompanied by high-energy proton beam injection. To obtain irradiation data, the SINQ target irradiation program (STIP) was initiated in 1996 at PSI. JAERI takes part in STIP and conducted the post-irradiation examination of JPCA, 316F. Irradiation conditions of JAERI specimens were as follows: proton energy was 590 MeV. Irradiation temperature ranged from 135 to 360 °C and irradiation dose from 6.3 to 12.5 dpa. The fatigue life of irradiated specimens is almost the same as that of unirradiated specimens. On the other hand, fracture surfaces varied with irradiation conditions. Specimens irradiated at low temperature fractured in a ductile manner. However, intergranular fractured surfaces were observed for 316F irradiated up to 12.5 dpa at 360 °C.

Fatigue properties of F82H irradiated at 523 K to 3.8 dpa

Fatigue properties were examined on a reduced activation ferritic/martensitic steel (F82H), and preliminary results are presented. F82H steel was irradiated at 523 K to 3.8 dpa, and then fatigue-tested at 298–573 K in vacuum with total strain range of 0.4–1.0%. The fatigue life of the irradiated specimen tested at 298 K with total strain range of 0.4% was revealed to be reduced to about 1/7 of that for the unirradiated specimen. The reduction of the fatigue life was attributed to the change of the fatigue mechanism to the channel fracture. The effect of test temperature is also discussed.

In-beam mechanical testing of CuCrZr

In the ITER design, CuCrZr has been selected as the heat sink material for components of the first wall and the divertor. The objective of this work is to check the material fatigue performance when the CuCrZr alloy is cyclically deformed concurrently with irradiation, using an in situ testing machine placed in a 590 MeV proton accelerator. Three fatigue experiments have been conducted at 100 °C, under strain control, at a total strain range of 0.8%. The in-beam specimen reached the longest life. The post-irradiation tested specimen had the shortest life. The total plastic strain measured in the in-beam specimen was larger than the plastic strain measured in the statically irradiated specimen or in the unirradiated specimen.

In-beam fatigue behavior of F82H steel at 500 °C

Stress-controlled fatigue tests were conducted for side-notched small specimens of F82H steel under in-beam and unirradiated conditions with 17 MeV protons at 500 °C. The environment for the fatigue tests was impure He gas containing 10–3000 ppm oxygen. The number of cycles to fracture ( N f) was slightly reduced under irradiation. SEM examination of fracture surfaces indicated the formation of an oxide film in the vicinity of notch tip only for the in-beam specimens. Optical microscopic observation demonstrated oxidation not only along the grain boundaries but also on the crystallographic slip planes on the specimen surfaces of both in-beam and unirradiated specimens. The oxidation appeared to be enhanced to form numerous oxidation pits on the surface, especially for the in-beam specimens. In this study, the in-beam fatigue behavior at 500 °C is discussed with respect to the dynamic irradiation effect in the low oxygen potential atmosphere.

Fracture surface topography analysis of in-beam fatigue behavior for 20% cold-worked 316 stainless steel

Topography on the fatigue fracture surface was examined by using a confocal laser microscope for the side-notched small specimens of 20% cold-worked 316 stainless steel fractured in stress-controlled fatigue tests under the in-beam (dynamic), post-irradiation (static) and unirradiation conditions with 17 MeV protons at 60 °C. A technique of the fracture surface topography analysis (FRASTA) was adopted to extract the information of plastic deformation at the crack tip for the in-beam, post-irradiation and unirradiated specimens. The progress in plastic deformation at the crack tip was significantly delayed in the dynamic irradiation condition while not so much as in the static irradiation condition. The present results of FRASTA would not only support the superiority of dynamic irradiation effect to the static irradiation effect on fatigue behavior but also provide good evidence for the mechanism of the dynamic irradiation effect based on the interaction between continuously induced defect clusters and mobile dislocations.

Thermal conductivity of neutron irradiated Be 12Ti

Beryllides are expected to be used as the advanced neutron multiplier for the DEMO blanket owing to their peculiar properties like higher melting point, high oxidation resistance, etc. which can improve the blanket design. In this study, thermal conductivity of neutron irradiated Be 12Ti was measured. The Be 12Ti specimens were fabricated by HIP process using beryllium and titanium powder, and were irradiated in the JMTR up to a fast neutron fluence ( E>1 MeV) of 4×10 20 n/cm 2 at temperatures of 330, 400 and 500 °C. Thermal diffusivity and specific heat of un-irradiated and irradiated specimens were measured up to 1000 °C by laser flash method. Then, the thermal conductivity was evaluated from these data. The effective thermal conductivity in the pebble bed of fusion blanket was also estimated and interesting prospects for blanket design were obtained.