Taiji Hoshiya

Restoration phenomena of neutron-irradiated TiNi shape memory alloys☆

Abstract It is believed that TiNi shape memory alloys can be used in fission and fusion reactor devices that may utilize shape memory capabilities. Three kinds of TiNi shape memory alloy were neutron irradiated with different doses and their electrical resistance measurements were carried out after isothermal and isochronal annealing. The temperature dependence of the electrical resistance for the irradiated alloys was found to be similar to that for unirradiated alloys after annealing at a temperature of 523 K, which gives a low homologous temperature T/Tm (where Tm is the melting point) of 0.33. The results obtained from neutron-irradiated alloys are discussed, taking into account the damage generated by both irradiation and structural alterations during annealing.

MAGNETIZATION OF CERAMIC Y-BA-CU-O AND BI-SR-CA-CU-O AFTER NEUTRON IRRADIATION

Magnetization of ceramic Y-Ba-Cu-O and Bi-Sr-Ca-Cu-O superconductors was studied after neutron irradiation in the fluence from 2.4×1021/m2 to 1.8×1022/m2 at 60°C. The area of hysteresis loops was enhanced at low neutron fluence, followed by saturation and then a decrease at high fluence. In Bi-Sr-Ca-Cu-O, the degree of enhancement was smaller than in Y-Ba-Cu-O and the enhancement reached saturation at lower neutron fluence.

Restoration phenomena of TiNi shape memory alloys in a neutron irradiation environment

Transformation properties and deformation behavior of TiNi shape memory alloys, which were neutron-irradiated at 323 and 520 K with a maximumfluence of 1.4 × 1025 m−2 and subsequently post-annealed at 473, 523 and 573 K, were measured by electrical resistance measurements and tensile tests. At an irradiation temperature of 323 K, abrupt changes in Ms temperatures and stress-strain curves of specimens were observed at a dose of over 10−2 dpa. Irradiation alterations were annihilated by post-irradiation annealing at temperatures above 523 K. On the other hand, 520 K irradiations brought about quite a few changes in those properties regardless of the magnitudes of displacement. The irradiated state of this alloy can be represented by two conflicting processes; ordering and disordering, which depends on temperature, displacement and displacement rate. The key temperature of that state is 520 K, at which the ordering becomes predominant over disordering and brings about the occurrence of a restoration phenomena.

Fast neutron irradiation of TiNi shape memory alloys

A study of the effects of neutron irradiation on shape memory capabilities in Ti-Ni shape memory alloys (SMA) has been carried out. These alloys permit the development of quick replacement techniques using SMA joints or SMA driving elements. Abrupt changes in both Ms temperatures and fracture energies in Ti-Ni alloys are brought about by neutron irradiation. On the other hand, restoration phenomena occur in this alloy both with post-irradiation annealing and upon irradiation at a low homologous temperature of 0.33 (T/Tm; where Tm is the melting point, T = 520 K). From the viewpoint of fusion reactor design, these may be useful phenomena for maintaining the irradiation resistance of Ti-Ni SMA core parts. Controlling factors governing the restoration of order in SMA are also discussed, taking into account both displacement damage and helium generation.

Irradiation effects on thermal expansion of SiC/SiC composite materials

Abstract Irradiation-induced dimensional change and thermal expansion of two kinds of composites, self-particle reinforced SiC p /SiC composites and a Hi-Nicalon™ SiC fiber reinforced SiC f /SiC composite, and monolithic α-SiC were measured after irradiation at 0.2 dpa with irradiation temperatures of 573, 673 and 843 K using the JMTR. From the measurement, swelling was observed for the SiC p /SiC composites and the monolithic α-SiC, on the contrary, the SiC f /SiC composites showed a shrinkage. The measured thermal expansion increased with increasing the specimen temperature below the irradiation temperature, and then rapidly decreased over the irradiation temperature. The so-called ‘temperature monitor effect’ of the silicon carbide was clearly observed for all specimens, the monolithic α-SiC and both composites.

High-temperature oxidation process of intermetallic compound Ti-42 at % Al

The oxidation process of two-phase (Ti3Al and TiAl) intermetallic compound, Ti-42 at% Al, in air at 1073 and 1273 K has been investigated. The oxidation at 1273 K is much faster than that at 1073 K; however, the oxidation kinetics are similar. During heating up, TiO2 scale forms initially on the compound surface at about 973 K, and then Al2O3 scale forms at about 1273 K. For the isothermal heating, TiO2 scale slowly grows up at 1073 K, while at 1273 K both TiO2 and Al2O3 scales grow up drastically. The outer oxide scale consists of TiO2 and the inner one consists of a mixture of TiO2 and Al2O3. The volume of Al2O3, which forms after TiO2 formation at the initial stage of oxidation, is larger at an area adjacent to the oxide-compound interface.

Pinning Strength of Bi–Sr–Ca–Cu–O Superconductor after Ion Irradiation

The temperature dependence of critical current was measured after removing an applied magnetic field in Bi-Sr-Ca-Cu-O films after He ion irradiation at room temperature. The curve was remarkably different from the temperature dependence of critical current without magnetic field. The difference between the temperature dependence of critical current after removing an applied magnetic field and without magnetic field is due to the presence of pinning denters produced by irradiation.

Application of shape memory alloys to compacting and element-quickly replaceable design in high-power density fusion reactors

Abstract Quick replacement has been accomplished by using a shape memory alloy (SMA) coupling in the conceptual design of a cassette compact toroid reactor (CCTR). Further, by using an SMA driving element, a compact large gate valve can be newly devised. This gate valve will enable in-situ handling without breaking vacuum, so that the baking of the vacuum boundary for every replacement becomes unnecessary, except the initial baking. In these applications of SMA, the compacting and quickly replaceable technology that are needed for dealing with the problems associated with very high neutron loading in a compact reactor become available by using presently available or promised materials rather than an assumed material.

Effect of helium to dpa ratio on fatigue behavior of austenitic stainless steel irradiated to 2 dpa

Abstract The effect of helium due to nuclear transmutation reactions during neutron irradiation on low cycle fatigue life of type 304 stainless steel was investigated. The specimens were irradiated in spectrally tailored capsules in the Japan Materials Testing Reactor (JMTR) at a temperature of 823 K to a neutron fluence of approximately 1×10 25 n / m 2 (E>1 MeV ) and helium levels of 0.8, 2.5 and 8.1 appm. The low cycle fatigue tests were performed in total axial strain ranges of 0.8–1.6% at 823 K. A laser extensometer was used for controlling the axial strain of a specimen under cyclic testing. The difference between unirradiated and irradiated specimens is quite clear and appears to be a reduction by a factor of 2–5 in fatigue life. The helium concentration of the specimen is not the main factor to shorten fatigue life in the present experimental condition.

Development of a small specimen test machine to evaluate irradiation embrittlement of fusion reactor materials

Abstract Small specimen test techniques (SSTT) are essential to use an accelerator-driven deuterium–lithium stripping reaction neutron source for the study of fusion reactor materials because of the limitation of the available irradiation volume. A remote-controlled small punch (SP) test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). This report describes the SP test method and machine for use in a hot cell, and test results on irradiated ferritic steels. The specimen was either a coupon 10×10×0.25 mm3 or a TEM disk 3 mm in diameter by 0.25 mm in thickness. Tests can be performed at temperatures ranging from 93 to 1123 K in a vacuum or in an inert gas environment. The ductile to brittle transition temperature of the irradiated ferritic steel as determined by the SP test is also evaluated.