Search

An in-pile functional test of a fusion blanket with lithium containing ceramics as the tritium breeder is planned in Japan Materials Testing Reactor (JMTR). Therefore, the getter is necessary for the recovery of the tritium gas on in-pile functional test. In this study, ZrNi alloys added other elements, i.e. Fe, Co, etc. were selected as a getter. The equilibrium hydrogen pressure of these alloys and breakthrough properties of hydrogen with the particle packed bed were estimated for its application. Resulting of the examination on plateau pressure of the P–C–T, the same plateau pressure as the uranium was obtained in the Zr1Ni0.2Co0.7Fe0.1 and Zr1Ni0.3Co0.5Fe0.2 in which the pressure was about 4×10−4 Pa at 298 K. The absorption band and amount of hydrogen absorption in the packed bed were about 0.7 cm and 170 cm3/g Zr1Ni0.3Co0.5Fe0.2 in 3 cm/s with 9940 ppm H2+Ar gas at 298 K, respectively. From these results, it was obvious that Zr1Ni0.3Co0.5Fe0.2 was the preferable material for the recovery of tritium gas for in-situ irradiation test of fusion blanket.

Lithium titanate (Li2TiO3) is one of the most attractive tritium breeders for breeding blanket in fusion reactor from view points of low tritium inventory, high chemical stability and so on. The data on the performance of a blanket mockup with pebble bed under neutron irradiation is needed for the design of breeding blanket. To obtain such data, two kinds of the blanket in-pile mockups with Li2TiO3 pebble bed were developed and the in-situ tritium recovery experiments were carried out in the Japan Materials Testing Reactor (JMTR). In these studies, effects of various parameters, i.e., irradiation temperature, sweep gas flow rate, etc., on the tritium recovery behavior from Li2TiO3 pebble bed were evaluated. It was found that the tritium recovery (R) to tritium generation (G) ratio (R/G) increased with increasing the temperature of Li2TiO3 pebble bed and was saturated when the temperature of Li2TiO3 pebble bed at the outside edge exceeded 300°C. Additionally, the sweep gas flow rate in the range of 100 to ...

The effects of neutron irradiation on tensile properties of the Cu alloys to type 316 stainless steel (SS316LN-IG) joints fabricated by friction welding method were examined. Additionally, irradiation tests of Al 2 O 3 -dispersed copper alloy (Al-25) and SS316LN-IG were conducted. The Al-25/ SS316LN-IG joints were irradiated in the Japan Materials Testing Reactor (JMTR), and tension tests and metallographical observation of these joints have been performed as the post irradiation examination (PIE). In the tension tests of Al-25/ SS316LN-IG joints, all irradiated specimens fractured at the copper sides. In the tension tests at 293K, tensile strength of the Al-25/ SS316LN-IG joints irradiated at each temperature was almost similar to that of the base metal (Al-25) irradiated in the same condition. On the other hand, tensile strength of Al-25/ SS316LN-IG joints was smaller than that of Al-25 in the tension tests at the same temperature as each irradiation temperature. In this study, useful data of the irradiated Al-25/ SS316LN-IG joints fabricated by friction welding method were obtained to design fusion reactors.

The study of neutron irradiation damage is inevitable for the development of fusion reactors such as the International Thermonuclear Experimental Reactor (ITER). In general, it is difficult to simulate the characteristics of irradiation damage, such as the relationship between helium production and displacement damage, expected in fusion reactors by using the existing test reactors. This is due to the neutron spectrum of test reactors, which are more thermalized than expected in fusion reactors. Despite such limitation, irradiation damage expected to occur in fusion reactors can be simulated in test reactors by spectrum tailoring techniques and/or a combination of irradiation at different conditions, so called irradiation–coupling. In Japan Atomic Energy Research Institute (JAERI), the technique of re-irradiation for irradiated specimens was developed to realize irradiation–coupling as stated above. The key technology in establishing such a technique is the development of a specially designed coupling capsule which can be assembled and loaded with irradiated specimens by remote handling devices in the hot cell. JAERI has recently developed the coupling capsule and successfully completed its irradiation in the Japan Materials Testing Reactor (JMTR).

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.

The austenitic stainless steel type AISI 316L was selected as the main structural material of the next-step International thermonuclear experimental reactor (ITER) fusion device, i.e., the first wall, blanket modules, and vacuum vessel components. Although this steel was extensively investigated under different aspects, most results concern irradiation temperatures above 300°C. In the present work, tensile and fatigue specimens were irradiated in the BR2 materials testing reactor at 42°C up to a maximum neutron fluence of 8×10 21 n/cm 2 ( E >0.1 MeV), corresponding to 5 dpa. The European reference AISI 316L in the solution annealed condition and the TIG-metal deposit were tested in the baseline and irradiated conditions. The tensile specimens were tested at 25°C, 250°C and 450°C, while the low-cycle fatigue tests were performed at room temperature. The tensile test results obtained in this work are consistent with published data: substantial radiation hardening combined with some reduction of elongation. No specimen orientation effect could be evidenced. The amount of hardening decreases with increasing test temperature. By contrast, the low-cycle fatigue data show no or little effect of irradiation, independent from irradiation and testing conditions. No major difference was found between the plate and the weld metal.

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.

Lithium titanate (Li2TiO3) is one of the candidate tritium breeding materials for breeding blanket of the fusion reactor. Therefore, tritium release experiments from Li2TiO3 packing region were carried out to evaluate the effects of various parameters, i.e. sweep gas flow rate, irradiation temperature, and hydrogen content in sweep gas, etc. on tritium release. As for the shape of the Li2TiO3, a small spherical form (pebbles) is preferred to reduce the induced thermal stress in the breeding material. The diameter and total weight of Li2TiO3 pebbles were 1 mm and ∼135 g, respectively. And the integrated experiment of blanket in-pile mockup with Li2TiO3 pebble bed was carried out at the Japan Materials Testing Reactor (JMTR). The tritium released from Li2TiO3 pebble bed was swept by the helium gas with hydrogen. The total tritium concentration (HT+HTO) and gaseous tritium concentration (HT) of tritium released from Li2TiO3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated under various conditions.

Magnetic probes for plasma diagnostics in the International Thermonuclear Experimental Reactor (ITER) were developed and tested in a fission reactor, the Japan Materials Testing Reactor. Performance of coils made of mineral insulating cables could be analyzed by a proposed equivalent electrical circuit but were disturbed by electrical drift, presumably caused by a phenomenon called radiation induced electromotive force. However, results were promising and development of an ITER-relevant magnetic probe is within view.

The effect of alloying elements, Si (-5.8%: the volume size factor in Ni), Ge (+14.76%) and Sn (+74.08%), on void swelling in neutron irradiated Ni at 573 K was studied by transmission electron microscope (TEM) observation and positron annihilation lifetime measurement. Neutron irradiation dose was changed widely from 0.001 to 0.4 dpa using two reactors, the Kyoto University reactor (KUR) and the Japan materials testing reactor (JMTR). Voids were observed in pure Ni by TEM even after very small irradiation dose of 0.001 dpa. With increasing dose, the density of voids did not change much while their size increased. The same tendency was observed in Ni-2at.%Ge. In Ni-2at.%Sn and Ni-2at.%Si, however, no voids were observed by TEM at a damage dose of 0.4 dpa. But positron lifetime measurement revealed the existence of microvoids at a medium dose of irradiation. When irradiation dose increased to 0.4 dpa in Ni-2at.%Si and 0.13 dpa in Ni-2at.%Sn, their existence was not detected. Suppression of microvoids in these alloys is discussed from the standpoint of solute point defect interactions.

This paper consists of two parts. In part 1, the experimental results of damage evolution of neutron-irradiated Cu and Ni are described. In part 2, results of computer simulations are described with linkage of experimental data to explore the atomistic process of damage evolution. To study experimentally the atomistic processes of damage evolution in neutron-irradiated Cu and Ni in part 1, we prepare two types of specimens for both metals. One is as-received specimen from manufacturer. Another is a residual-gas-free specimen which is prepared by melting as-received metals in highly evacuated vacuum at 10−5 Pa. Specimens are irradiated with fission neutrons in the temperature-controlled-irradiation capsule at JMTR (Japan Materials Testing Reactor). TEM (Transmission Electron Microscope) observation shows that the dislocation structure is developed by the aggregation of interstitial clusters in irradiated metals. It is found that the number density of void which are observed in specimens, both as-r...

A precipitation hardening type 630 stainless steel was irradiated in the Japan Materials Testing Reactor (JMTR) in contact with the reactor primary coolant. The temperature of the irradiated specimens was about 330 K. The fast neutron (E > 1 MeV) fluence for the specimens ranged from 1024 to 1026 m−2. Tension tests and fracture toughness tests were carried out at room temperature, while Charpy impact tests were done at temperatures of 273–453 K. Tensile strength data showed a peak of 1600 MPa at around 7 × 1024 m−2, then gradually decreased to about 1500 MPa at 1.2 × 1026 m−2. The elongation decreased with irradiation from 12% for unirradiated material to 6% at 1.2 × 1026 m−2. The fractography after the tension test revealed that the fracture was ductile. Fracture toughness decreased to about a half of the value for unirradiated material with irradiation. The cleavage fracture was dominant on the fractured surface. Charpy impact tests showed an increase of ductile–brittle transition temperature (DBTT) by 60 K with irradiation.

Miniature Charpy V-notch impact test specimens of commercial reactor pressure vessel (RPV) steels having high and low copper contents were irradiated at the different irradiation positions with neutron flux levels of ∼6 × 10 14 , ∼7 × 10 15 , and ∼8 × 10 16 n m −2 s −1 ( E > 1 MeV) to fluence levels ranging from ∼6 × 10 21 to ∼7 × 10 22 n m −2 ( E > 1 MeV) at temperatures of about 50°C to 150°C in the Japan Materials Testing Reactor (JMTR). The results showed that the radiation-induced increases in ductile-to-brittle transition temperature (ΔDBTT) at a neutron flux level of ∼6 × 10 14 n m −2 s −1 were greater than those for neutron flux level of ∼7 × 10 15 n m −2 s −1 . The neutron flux effect on embrittlement tended to be more pronounced in the lower neutron fluence range of ∼6 × 10 21 –∼1 × 10 22 n m −2 than in the higher fluence level of ∼7 × 10 22 n m −2 , and also to be larger for the low copper steel than for the high copper steel, although the ΔDBTT for the high copper steel was larger than that for the low copper steel regardless of neutron fluence or flux. The displacement dose rate effect identified by the data converted to the ΔDBTT for the full size Charpy specimens from those for the miniature Charpy specimens was consistent with that based on the comparison of the results in the literature.

Various carbon fibers with different microstructures were irradiated in the Japan Materials Testing Reactor (JMTR). Tensile tests were conducted and the microstructures were also examined before and after neutron irradiation. Clear correlations between the tensile properties and the microstructures were found before and after irradiation.The data obtained here will be useful for the development of C/C composite materials that have higher resistanceto radiation damage.

Mg-Al spinel ceramics with different stoichiometry (Al 2 O 3 /MgO=n; n = 1.00, 1.01, 1.10 and 1.48) were irradiated with fast neutrons up to 3.8 x 10 23 n/m 2 (E > 0.1 MeV) at 170°C, and 5.7 x 10 23 n/m 2 at 190°C in the Japan Materials Testing Reactor. There was no obvious difference in property change between the two irradiation conditions. The amount of macroscopic length change due to these irradiations increased almost linearly with increasing Al 2 O 3 content. These specimens were isochronally annealed up to 1000°C, During the annealing, macroscopic length of the specimens started to decrease at around 200°C and reduced monotonically up to 550°C, After that, the length increased slightly up to 650°C and then decreased again up to 800°C, showing a small peak. These final values corresponded to the length of the specimens before the irradiation. The shrinkage in length between 200-550°C and 650-800°C may be attributed to the recombination of Frenkel pairs of anions, and the expansion in length between 550-650°C to the order-disorder transition of cation distribution.

Oxygen sensors based on zirconia solid electrolyte stabilized by MgO, CaO or Y2O3 were irradiated with neutrons in the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Research Institute (JAERI), and the characteristics of electromotive force of these sensors under and after irradiation were measured. The electromotive force of YSZ sample decreased with increasing irradiation fluence up to 1 × 1023 m−2 (E > 1 MeV). The electromotive force of the MSZ sensor irradiated for fluences up to 9 × 1021 m−2 (E > 1 MeV) was almost equal to the reference value of the electromotive force. It was shown that after irradiation, the decrease in the electromotive force of the CSZ sensor was smaller than those of MSZ and YSZ sensors, although the electromotive forces of MSZ, CSZ and YSZ sensors were smaller than the reference value.

As a part of a program to screen several V–4Cr–4Ti containing Si, Al and Y alloys and optimize the amounts of Si, Al and Y, the Charpy impact test of five kinds of V–4Cr–4Ti–Si–Al–Y alloys by an instrumented Charpy impact testing machine using miniaturized specimens (1.5 mm × 1.5 mm × 20 mm) have been conducted before and after neutron irradiation. Charpy impact specimens were encapsulated in an aluminum vial filled with high purity He and irradiated up to 1.06 × 10 1 9 n/cm 2 ( E > 1 MeV, 156 h) at low temperatures (about 150°C) in Japan Materials Testing Reactor (JMTR). The ductile brittle transition temperature (DBTT) of each alloy was determined by various methods on absorbed energy, brittle fracture ratio and lateral expansion from a quantitative analysis of fractography for broken specimens after the Charpy impact test. Almost all specimens were embrittled after low temperature irradiation. Decomposition of primary precipitates could result in migration of interstitial elements to irradiation defects and many precipitates are formed under irradiation. Radiation hardening then caused the substantial degradation of its fracture toughness.

Neutron irradiation effects on a Ti–Ni shape memory alloy (SMA) are not understood regarding the engineering behavior of tightening and connecting core parts. Irradiation response on shape memory capabilities for Ti–Ni SMA couplers was investigated by measuring the inner diameter of a SMA coupler. Ti–Ni SMA couplers were irradiated to a fast neutron fluence ( E > 1 MeV) of 1.4 × 10 25 m −2 at 323 and 503 K in the Japan Materials Testing Reactor of JAERI. The inner diameter versus temperature curves showed a large hysteresis for the SMA coupler irradiated at 323 K. At a neutron fluence above 10 25 m −2 , the hysteresis after 323 K irradiation increased by 50% more than that after 503 K irradiation. The maximum amount of reversible stroke of the irradiated two-way Ti–Ni SMA coupler during heating and cooling was reduced to 70% that of an unirradiated one and the two-way Ti–Ni SMA coupler is more resistant to neutron irradiation than the one-way Ti–Ni SMA coupler.

Ceramic coatings on the surface of structural materials such as 316SS have been considered for electrical insulators and tritium permeation barriers in fusion reactor designs. Y 2 O 3 is one of the most promising materials as a coating from a point of high electrical resistivity, etc. In this report, the electrical conductivity of a Y 2 O 3 coating was investigated under neutron irradiation with the Japan Materials Testing Reactor (JMTR). The specimen was 316SS with a Y 2 O 3 coating, and was irradiated at 300°C in a He atmosphere. From the results of in situ measurements on the electrical conductivity of a Y 2 O 3 coating, the Radiation Induced Conductivity (RIC) was observed at JMTR power-up. The electrical conductivity of the coating before neutron irradiation and under neutron irradiation were about 3 × 10 −13 and about 1 × 10 −9 1/Ω/cm, respectively. The electrical conductivity of the coating was constant under neutron irradiation but after neutron irradiation recovered up to that before neutron irradiation. The Radiation Induced Electrical Degradation (RIED) in the Y 2 O 3 coating was not recognized up to a fluence of about 5 × 10 20 n/cm 2 (>1 MeV). It was clear that the coating had good electrical resistivity under neutron irradiation corresponding to about 4 × 10 10 Gy.

Acquisition of systematic irradiation data are essential for understanding fundamental processes of irradiation effects and for establishment of a reliable database for irradiation effects in fusion reactor materials. It will take several years with expensive several different irradiation rigs in a fission reactor irradiation. There, it will take a long time to carry out the needed iterations between irradiation tests and evaluation and materials development. An irradiation rig was developed to carry out irradiation under multiple temperatures and irradiation fluences. Irradiation tests of fusion reactor materials were successfully carried out using the rig in Japan Materials Testing Reactor.

A beryllium irradiation study with the Japan Materials Testing Reactor of JAERI is reviewed from the point of view of tritium release, thermal properties, mechanical properties and reprocessing technology by using the neutron irradiated sample with a total fast neutron fluence up to 10 21 (n cm −2 ). It becomes clear that problems arise for each of the properties, for example, effects of the surface oxide layer, impurities, grain size, and/or swelling. Moreover, future plans of beryllium irradiation studies are discussed in this paper.

Irradiation hardening and embrittlement of reactor pressure vessel (RPV) steels having different contents of copper and nickel have been investigated as a part of the phase II of IAEA Coordinated Research Program (CR P). Seven kinds of materials which were made by Japanese steel manufacturers for the CRP were used. Neutron irradiation was conducted in Japan Materials Testing Reactor (JMTR). Mechanical properties of the materials were examined by conducting the hardness, tensile, Charpy impact and fracture toughness tests before and after irradiation. Some relationships between the changes of the mechanical properties due to irradiation were established. The increase in yield strength was correlated with hardness increase and the shift of Charpy transition temperature. In the upper shelf temperature range, the decrease in fracture toughness was well correlated with the degree of the increase in yield strength.

An accelerator-driven deuterium-lithium (d-Li) stripping reaction-type neutron source, such as the International Fusion Materials Irradiation Facility (IFMIF) planned by the International Energy Agency is recognized as one of the most promising facility to obtain test environments of high-energy neutrons for fusion reactor materials development. The limitation on the available irradiation volume of the irradiation facility requires the development of the small specimen test techniques (SSTT). Application of SSTT to evaluate the degradation of various components in the light water reactor for the life extension is expected to be also quite beneficial.A remote-controlled testing machine for the Small Punch (SP) and miniaturized tensile tests was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR). The machine is designed for testing at temperatures ranging between 93 and 1, 123K to evaluate the temperature dependence of the strength of materials including the embrittlement at low temperatures and the softening at elevated temperatures. The tests are performed in a vacuum or in an inert gas environment.The machine has been installed in a hot cell and is being used for the round robin test program of the SP test method. The round robin test program is planned to identify the capability of the test method and to establish a standard test procedure. The configuration and the specifications of the test machine are introduced and the results of the SP tests are also shown.

Stainless steel is a candidate material for the structural material in fusion reactors. Rewelding of irradiated materials will have a large impact on the design and the maintenance of in-vessel components. In the present work, the welding specimens made of type 316 stainless steel were irradiated in JMTR (Japan materials testing reactor) to a fast neutron fluence of ∼2.0 × 10 20 n/cm 2 ( E > 1 MeV) at a temperature of ∼200°C. The rewelding of unirradiated and/or irradiated stainless steel was performed by the tungsten inert gas (TIG) welding method and the weldments of unirradiated and/or irradiated SS316 were characterized by tensile testing (test temp.: 20°C and 200°C), hardness, metallographical observation and SEM/XMA analyses.

The austenitic stainless steels of type 304 and type 316 and the precipitation hardened type 630 stainless steel were irradiated to doses ranging from 5 × 1024 to 1.2 × 1026 at 323 K in the Japan Materials Testing reactor (JMTR). Tensile and fracture toughness tests have been carried out at 293 K. Type 304 and 316 steels in the solution annealed condition hardened by irradiation to about 700 MPa. Fracture toughness values for these steels decreased with dose. Only a small hardening occurred after irradiation for cold worked type 304 and precipitation hardened type 630 steels. The effect of irradiation on the fracture surfaces was small.

Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop plasma facing components which can resist these. We have established electron beam heat facility (‘OHBIS’, Oarai hot-cell electron beam irradiating system) at a hot cell in JMTR (Japan materials testing reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50 kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30 kV (constant) and 1.7 A, respectively. The loading time of the electron beam is more than 0.1 ms. The shape of vacuum vessel is cylindrical, and the main dimensions are 500 mm in inside diameter, 1000 mm in height. The ultimate vacuum of this vessel is 1 × 10−4 Pa. At present, the facility for the thermal shock test has been established in a hot cell. The performance of the electron beam is being evaluated at this time. In the future, the equipment for conducting static heat loadings will be incorporated into the facility.

TiNi and TiPd shape memory alloys (SMA) reveals good workability and shape memory properties and these SMAs seem to be one of hopeful functional materials in a severe irradiation field. The effects of irradiation induced residual strain on the deformation processes of TiNi and TiPd SMAs after neutron irradiation with fluences ( E > 1 MeV) up to 3.9 × 10 24 m −2 at Japan Materials Testing Reactor (JMTR) were investigated by the remote controlled X-ray diffraction measurement. Residual strains of TiNi SMAs took place over damage of 0.1 dpa and the strains were not completely removed by post-irradiation annealing at 473, 523 and 573 K. On the other hand, residual strains of TiPd SMAs scarcely occurred after irradiation, and furthermore, the TiPd SMA seems to be an irradiation-resistant material. This may be explained by a difference between the irradiation response of a parent phase and that of a martensitic phase to neutron irradiation. Deformation processes of SMAs are associated with stress fields generated by irradiation in a parent phase with B2 type ordered structure or in a martensitic phase with 2H and 9R type close-packed structures.

High thermal diffusivity and insulating aluminum nitride (AIN) ceramic was neutron-irradiated in the Japan Materials Testing Reactor. Fast neutron fluence and irradiation temperature of the specimens were 1.8 × 10 23 n/m 2 at 100°C and 5.3 × 10 24 n/m 2 at 470°C ( E > 0.1 MeV), respectively. Change in thermal diffusivity due to isochronal thermal annealing up to 1400°C was measured by the laser-flash method. As compared with macroscopic length change, it was concluded that thermal diffusivity of irradiated AIN was controlled by concentration of irradiation-induced vacancies.

An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

Microstructures and tensile properties of a martensitic steel F82H (Fe-8Cr-2W-0.1C-0.2V-0.04Ta) were examined as a function of time and temperature of tempering. A heat treatment was performed at temperatures from 750 to 800°C for 0.5 h after the normalizing at 1040°C for 0.5 h. The tempering time at 750°C was varied from 0.5 to 10h. The tensile specimens were irradiated at 250°C to a neutron dose of 1.9 dpa in the JMTR (Japan Materials Testing Reactor), and tensile tests were carried out at 250°C after the irradiation. The microstructures of the non-irradiated specimens were observed by a transmission electron microscope. The density of dislocations decreased with increasing time and temperature of the tempering, and the size of M 2 3 C 6 carbide increased with it. While the yield stress of the non-irradiated specimens decreased with increasing time and temperature of tempering, the yield stress of the irradiated specimens tended to increase with increasing temperature of the tempering. The yield stress of the irradiated F82H steel changed from about 525 to 750 MPa and depended on the conditions of tempering treatment before irradiation.

Dependence of fracture properties and hardening was examined as a function of helium production in tensile specimens of a martensitic steel F82H (Fe-8Cr-2W-0.1C-0.04Ta) irradiated at 300°C to 2.3 dpa by neutron irradiation in the JMTR (Japan Materials Testing Reactor). The specimens used in this study were F82H, F82H+60ppm 1 1 B, F82H+30 ppm ( 1 1 B+ 1 0 B) and F82H+60ppm 1 0 B. The helium range produced from 1 0 B (n,α) 7 Li reaction was from 5 to 330 appm in the specimens. The tensile testing was performed at 25°C. The radiation hardening due to helium production was detected at 330appmHe. The degradation of fracture stress due to helium production was approximately evaluated from the fracture strength and the reduction area. Effect of specimen size on tensile and Charpy impact properties in F82H doped with 60ppm boron and 200 ppm nitrogen was also examined. The JIS 14A and SS-J3 (Small Size-Japanese-3 type) were used for the tensile specimens, and half size (55 mm in length, 10 mm in height and 5 mm in width) and 0.5-1/3CVN (18 mm in length, 3.3 mm in height and 1.65 mm in width) were used for the Charpy impact testing. The tensile properties were a similar to each other. However, the ductile-brittle transition temperature measured in smaller size specimen was somewhat lower than that in the standard size specimen.

An in-situ irradiation test of a fusion blanket is planned in JMTR (Japan Materials Testing Reactor) using lithium containing ceramics as the tritium breeder and beryllium as the neutron multiplier. Therefore, it is necessary to develop getters for the recovery of the tritium gas. In this study, Zr 9 Ni 11 alloy was chosen as a getter, and pressure-composition isotherms between 20°C and 300°C, and other properties were measured for its application. The breakthrough test of hydrogen was carried out with Ar+H2 gas. The evaluation of the examined getter materials leads to the conclusion that the Zr 9 Ni 11 alloy may be used for the in-situ irradiation test of fusion blanket.

Tritium release apparatus with the function of pulse mode heating was developed by using the infrared ray furnace to demonstrate the pulse mode heating of tritium breeder blanket for the fusion reactor. This apparatus was installed in the glove box of the beryllium PIE facility that constructed the hot-laboratory of Japan Materials Testing Reactor. The performance of this apparatus is that the minimum time of rapid heating up to 1015 {degree}C is about 119 s and maximum heating rate reached at 1000 {degree}C/min. The maximum temperature depends on the crucible materials because of the differences for infrared ray absorption. The conversion efficiency of the gaseous water by ceramic electrolysis cell is above 99.99%. The pulse mode heating of the tritium breeder and neutron multiplier materials of the blanket could be demonstrate by using this apparatus. 7 refs., 7 figs., 1 tab.

MgO·3Al2O3 single crystals were irradiated with neutron fluences of 8.3 × 1022 n/m2 at 100°C and 2.4 × 1024 n/m2 at 470°C (E > 1.0 MeV) in the Japan Materials Testing Reactor. The Knoop microhardness of several orientations on the (100) plane of both the irradiated and unirradiated crystals were measured with different indentation loads. The change in hardness profile of the crystals was almost the same after the two irradiation conditions. The hardness increased by 4–15% because of the irradiations depending on the crystallographic orientation, the larger change being observed at orientations between the (001) and (011) directions. While both the {111} and {110} slip systems are simultaneously active in the unirradiated MgO·3Al2O3, the {111} system may be the dominant slip system in the neutron-irradiated crystals. It is concluded that the restriction of the {110} slip system is caused by irradiation-induced interstitial ions.

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

Thermal diffusivity of two kinds of two-directional carbon-carbon composites were studied after neutron irradiation at temperatures 640–1070°C to a maximum neutron fluence 1.98 × 10 25 n/m 2 ( E > 29 fJ , 0.18 MeV) in the JMTR (Japan Materials Testing Reactor). The decrease of thermal diffusivity was observed after irradiation, and its significant decrease occurred at low neutron fluence, followed by gradual decrease with neutron fluence. The changes in thermal diffusivity, which were normalized to unirradiated values, were large at low measurement temperatures and decreased with temperature and also increased with neutron fluence. The changes in thermal conductivity related to thermal diffusivity are also discussed.

Tensile properties of reduced-activation 10Cr-30Mn austenitic steels with carbon levels from 0.003 to 0.55% were investigated over the temperature range from room temperature to 873 K after neutron irradiation in the Japan Materials Testing Reactor at 573 K to 8.5 × 10 22 n/m 2 . Irradiation-induced increase in yield stress increased significantly with carbon concentration up to about 0.1% and it was constant above 0.1% carbon. A high density of dislocation loops with small (below 10 nm) and large (20–30 nm) sizes formed during irradiation. The high density, small loops caused a large irradiation hardening, while the large loops contributed only slightly to irradiation hardening. It was considered that carbon atoms formed the small loops together with irradiation defects. The deformation channeling was observed in the irradiated high carbon steels, 0.11 and 0.55% carbon, but not in the very low carbon steel, 0.003% carbon, after deformation near room temperature.

Two kinds of optical fibers were irradiated in a fission reactor, JMTR (Japan Materials Testing Reactor), up to a 1.55 × 10 19 n/cm 2 fast neutron fluence and a 3.3 × 10 9 Gy ionizing dose at 370 K. Optical transmission spectra were measured in the wavelength range of 450–1750 nm, in situ. Growth of strong optical absorption bands and a peak were observed in the range of wavelength shorter than 750 nm. In the meantime, the fibers showed good radiation-resistance in the range of wavelength larger than 750 nm. Some optical radiation from the fibers was observed. The main cause of the optical radiation is thought to be so-called Cherenkov radiation.

The accumulated results of radiation effects on materials have clearly shown that temperature history is very important for the evolution of microstructures induced by irradiation. The Japan Materials Testing Reactor has several preferable features for temperature-controlled irradiation. Two kinds of techniques were applied to control the irradiation temperature. The control of thermal insulation along with heating by the electrical heater, which is the other technique of the temperature control, permits us to control the temperature in the range of 360–1100 K, regardless of whether the reactor is in the operation or out of operation.

The characteristics of pebble packing and the sweep gas pressure drop have been investigated for the design of the in-pile mock-up in Japan Materials Testing Reactor, and the results obtained are the following. The packing fraction of single diameter pebbles is kept at constant, i.e., about 63%, under the condition that the ratio of tube inside diameter to pebble diameter is above 10. The pebble distribution in the bed is not homogeneous, i.e., the mixture of close packing and loose packing zones both in the middle of bed and near wall. The packing fraction is about 77% for two-size pebble packing consisting of O1 and 5 mm pebbles. The measured pressure drops agree with those predicted by the Kozeny-Carman equation within the range of (+25)−(−60)%. The pressure drop is not affected by moisture concentration (< 100 ppm) and does not change for tests lasting as long as 300 hours.

To investigate the effects of cyclic temperature changes during neutron irradiation upon radiation induced microstructure evolution and resulting property changes of materials is very important from both fundamental and engineering viewpoints. Therefore, a technique that allows us to do the controlled temperature-cycle irradiation was developed in the Japan Materials Testing Reactor (JMTR). The technique meets the following requirements: (1) the temperature-cycle irradiation is to be performed under three different conditions by changing lower and upper temperatures; 200⇓400°C, 300⇓400°C and 300⇓450°C; (2) the number and period of the temperature-cycles are to be six for 24-day full irradiation and approximately 44 h/44 h at the lower/upper temperatures; (3) the temperatures of each specimen assembly are to be maintained at the lower temperatures before start-up of the reactor and at the upper temperatures during shutdown until the complete absence of reactor power. In this paper, the details of the irradiation rig, successful results and several problems to be overcome for future improvement are presented.

Magnetic probes are installed on the first wall of a fusion reactor to measure the magnetic field. These magnetic probes are exposed to the neutrons generated by the D-T reaction and γ-rays. However, electrical properties and durability of the magnetic probe under irradiation have not yet been satisfactorily investigated. Therefore, irradiation tests have been performed for two types of magnetic probes at the JMTR (Japan Materials Testing Reactor). It has resulted from these irradiation tests that the inductance of the magnetic probe with ceramic coated platinum wire is stable below about 550°C, and the inductance of the magnetic probe with MI-cable of 0.5 mm diameter has a temperature dependency and maintains reproducibility even above 550°C.

The dimensional changes of two very different graphites under fast neutron irradiation are analysed in detail. One of the graphites is anisotropic, manufactured from needle coke (Pile Grade A graphite), while the other is isotropic, manufactured from Gilsonite pitch coke (AGR moderator graphite). Samples of both graphites were irradiated in the temperature range 400–600°C in the PLUTO Materials Testing Reactor at AEA Technology Harwell in their normal form and several variants. The latter included a series of Pile Grade A base stock (i.e., unimpregnated), heat treated to various final temperatures and both graphites doped substitutionally with boron eleven. It is found that the dimensional changes, pore volume changes and various structure factors relating macroscopic to microscopic properties are controlled by the parameter XT, the difference in crystallite strains parallel and perpendicular to the hexagonal axis. A number of correlations between the structure factors connecting various properties are proposed and a qualitative model described which, with appropriate parameters, should apply to all graphites under neutron irradiation.

Two kinds of hot-pressed and cold-pressed boronated graphites, of which boron contents are 3wt%and 30wt%, were irradiated up to a maximum neutron fluence 2×1025n/m2 (E>29fJ) at 550-1070°Cin the JMTR (Japan Materials Testing Reactor).Irradiation-induced changes in length, thermal expansivity and thermal conductivity were measured after the irradiation. The changes depended on the direction measured and the fabrication method inaddition to irradiation conditions. Thermal conductivity decreased and its temperature dependence changed after the irradiation.

Samples of highly oriented graphite heat-treated to final temperatures of 2,000, 2,200, 2,400, 2,800 and 3,000°C have been successively irradiated with fast neutrons at controlled temperatures of 430° and 600°C in the PLUTO Materials Testing Reactor. The dimensional changes of these samples, measured parallel and perpendicular to the deposition plane, correspond closely to those of the component crystallites parallel and perpendicular to the basal planes. The changes are found to agree well with expectation from current theories of irradiation damage.

The tensile properties of reduced-activation martensitic 9Cr-IWVTa and 9Cr-3WVTa steels for fusion reactors were investigated over the temperature range from room temperature to 873 K after neutron irradiation in the Japan Materials Testing Reactor at 538 K to fast neutron fluences of 2×10 23 and 3×10 24 n/m 2 (E>1 MeV). A conventional 9Cr-1MoVNb steel was also examined for comparison. The irradiation caused an increase in yield stress and a decrease in total elongation. Irradiation-produced defect clusters could not be seen directly by transmission electron microscopy. The increase in yield stress caused by irradiation Δσ y was proportional to the 1/4 power of the neutron fluence

Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.

Aluminum nitride was neutron-irradiated in the Japan Materials Testing Reactor up to 5.2×10 24 n/m 2 ( E > 1.0 MeV). The specimens were isochronally annealed under a vacuum up to 1400°C. The macroscopic length and lattice parameters of both a - and c -axis of the specimen irradiated to 8.3 × 10 22 n/m 2 at about 100°C decreased concurrently with an increase in annealing temperature from 100°C, and returned to the pre-irradiated values after annealing at 1100°C. The macroscopic lengths of the specimens irradiated to 2.4×10 24 n/m 2 at 470°C and 5.2×10 24 n/m 2 at 785°C, both containing dense interstitial dislocation loops, were not significantly influenced by annealing up to 800–900°C. Then they decreased rapidly with increasing annealing temperature up to 1400°C. The change in c -axis lattice parameter by annealing was larger than that of a -axis above 900–1000°C, and finally reduced to the pre-irradiated values after annealing at 1400°C.

Tensile properties of the reduced activation martensitic steels 9Cr-1WVTa and 9Cr-3WVTa were investigated over the range from room temperature to 873 K after irradiation in the Japan Materials Testing Reactor at about 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The increase in yield stress due to irradiation measured at room temperature was smaller in the 9Cr-1WVTa and 9Cr-3WVTa steels than in conventional 9Cr-1MoVNb steel and was smaller in the low-W steel than in the high-W steel. With increasing test temperature, the irradiation hardening exhibited a further increase at around 573 K and then decreased to zero over the range 673 to 873 K. The increase in irradiatio hardening at around 573 K was believed to be due to vacancy-carbon complexes. The present results were also compared with earlier irradiation hardening experiments.

The Japan Materials Testing Reactor (JMTR) is now under refurbishment, and the operation of the new JMTR will start in FY 2011. The new JMTR has a plan to produce 99Mo, which is the parent nuclide of 99mTc, and two 99Mo production technologies have been developed: one is a solid irradiation method, and the other is a solution irradiation method. In the solid irradiation method, it was found that JMTR can provide about 20% of the 99Mo imported into Japan. In the solution irradiation method, the fundamental characteristics of the aqueous molybdate solutions selected as candidates for the irradiation target were cleared by the gamma-ray irradiation test. Aiming at the domestic production of 99Mo in Japan, the development of 99Mo production technologies in JMTR is kept going.