Japan Materials Research Articles

Behavior of optical fibers under heavy irradiation

Several kinds of optical diagnostics are planned in a fusion reactor. Complicated optical systems such as periscopes are thought to be primary candidates for optical measurements, especially for visible wavelengths. However, optical fibers have several advantages over such optical systems. Also, the optical fibers could be a far better transmission line for signals under a high electromagnetic field. However, they have been considered vulnerable to heavy irradiation. In this study, several kinds of optical fibers were irradiated in the Japan Material Testing Reactor (JMTR). The optical transmissivity in fibers was measured in situ during fast neutron and gamma irradiation, up to doses of 2×10 24 n m −2 and 5×10 9 Gy, respectively. The irradiation temperature ranged from 300 to 700 K. For pure ionizing irradiation environments, some methods for improving the radiation resistance of optical fibers were indicated. The results showed that effects of the irradiation associated with fast neutrons would be different from the effects of pure ionizing irradiation. Some fibers were found to withstand the heavy irradiation, especially in an infrared wavelength range.

Correlation of Intergranular Stress Corrosion Cracking Susceptibility with Mechanical Response of Irradiated, Thermally Sensitized Type 304 Stainless Steel

Abstract Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated, thermally sensitized type 304 (UNS S30400) stainless steels (SS) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 5.4 × 1023 n/m2 (E > 1 MeV) at 290°C and to 1.1 × 1024 n/m2 (E > 1 MeV) at 340°C in the Japan Material Test Reactor. Irradiated specimens were examined by slow strain rate tensile testing (SSRT) in 290°C pure water of 0.2 ppm dissolved oxygen (DO) concentration, by SSRT at 290°C in argon gas, and by microhardness measurements. IGSCC susceptibility increased with neutron fluences up to 1.1 × 1024 n/m2 regardless of irradiation temperatures. The 0.2% proof stress (PS) and the grain-boundary (GBHV) and grain-center microhardness (GCHV) increased. The uniform elongation (UE) decreased with neutron fluences up to 5.4 × 1023 n/m2, but 0.2% PS and GBHV and GCHV decreased, and UE increased at 1.1 × 10...

Correlation among the Changes in Mechanical Properties due to Neutron Irradiation for Pressure Vessel Steels.

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 III of IAEA Coordinated Research Program (CRP). 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.

微小試験片試験技術の開発 遠隔操作型スモールパンチ試験装置の開発

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.

Reweldability test of irradiated SS316 by the TIG welding method

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.

Effect of neutron irradiation on passive oxidation of silicon carbide

Hot-pressed SiC and pressureless-sintered SiC polycrystalline specimens were neutron-irradiated in the Japan Materials Testing Reactor up to 4.5 × 10 24 n/m 2 ( E ≫ 1.0 MeV). The bulk specimens were crushed into powders to increase surface area. Weight change of the specimen and released CO 2 gas content were measured after isothermal annealing between 500 and 1300°C in oxidizing atmosphere using an electric balance and a gas chromatograph, respectively. There was no significant difference between irradiated and unirradiated specimens, for both hot-pressed and pressureless-sintered specimens. All of the weight gain data and amount of evolved CO 2 satisfied a parabolic rate law against oxidation time. Apparent activation energy of the reaction confirmed the diffusion of O 2 molecules through amorphous SiO 2 layer. However, apparent activation energy decreased by the irradiation only in pressureless-sintered SiC containing B, suggesting the effect of transmuted atoms, or Li.

Effects of residual strain on deformation processes of neutron-irradiated TiNi and TiPd shape memory alloys

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.

Mechanical properties of austenitic stainless steels irradiated at 323 K in the Japan Materials Testing reactor

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 × 10 24 to 1.2 × 10 26 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.

New electron beam facility for irradiated plasma facing materials testing in hot cell

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.

Effect of isochronal annealing on thermal diffusivity of neutron-irradiated AIN

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.

Radiation-induced electrical degradation experiments in the Japan materials testing reactor

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/m2 (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.

Research on Neutron Activation Analysis of Biological Samples, Coals and High Purity Materials

In the first study, the analytical conditions for the determination of platinum in biological samples by instrumental neutron activation analysis (INAA) were optimized. The distributions of platinum and selenium in organs of rats treated with cisplatin and sodium selenite were determined. In the second study, selenium in soybean originating from Japan, China, Formosa, Burma, Brazil, Argentina and U.S.A. was determined by INAA. The γ-ray spectrometry was three counting methods, two conventional countings for 75Se and 77mSe, and a coincidence counting for 75Se. Concentration of selenium in analyzed soybean ranged from 5200 ppb. In the third study, the contents of about 50 elements in very bitter edible plants and the less bitter edible plants were determined by INAA. Especially, the contents of Al, Mg, Mn, K, Co, Cr, Fe and Te in very bitter edible plants were higher than in the less bitter edible plants. In the fourth study, distributions of 54 elements in coal samples were determined by INAA. In the fifth and sixth study, uranium and thorium in a high purity silica and an aluminium were analyzed by radiochemical neutron activation analysis (RNAA) with anion exchange chromatography and coprecipitation of LaF3. The lower limit for the determination of uranium or thorium was found to be on the ppt level. In the seventh study, ultra-trace amounts of uranium and thorium in high purity silica, aluminium, copper and niobium oxides were determined by RNAA using the Japan Materials Testing Reactor (JMTR).

Development of Tritium Release Apparatus Using Pulse Mode Heating

Tritium release apparatus with the function of pulse mode heating were 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 °C is about 119 s and maximum heating rate reached at 1000 °C/min. The maximum temperature is depended 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.

Development of Sweep Gas Sensor with Proton Conduction

To obtain engineering data for a design of a fusion blanket, a program in which a mock-up simulating a blanket structure is irradiated is proceeded in the JMTR (Japan Materials Tasting Reactor) project.1 It is indispensable for this irradiation test to purge the tritium produced in the mock-up by the sweep gas and to measure the tritium concentration in the sweep gas on line. However, the bad effects such as tritium adsorption on the inner surface of the pipe and tritium diffusion from the pipe are apprehended, because the distance between the mock-up and an ionization chamber is about 20m long. Therefore, it is necessary to develop a sweep gas sensor which can measure the tritium concentration in the mock-up directly without the above apprehension. A sweep gas sensor had been fabricated on trial and several performance tests were carried out. From the results of the tests, it was found that this sensor could be used at the atmosphere of high temperature of 600–700°C and had good response.

Preliminary Characterization of Zr9Ni11Alloy for Its Tritium Gettering Property in In-Ditu Irradiation Test

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.

Effect of Neutron Irradiation on Knoop Microhardness Anisotropy in MgO·3Al2O3 Single Crystals

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.

Thermal diffusivity of neutron-irradiated carbon-carbon composites

Abstract 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.

Development of experimental rigs for temperature-controlled reactor irradiations for fundamental study of radiation effects on fusion materials

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.

Effect of carbon on irradiation hardening of reduced-activation 10Cr-30Mn austenitic steels

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.

Behavior of radiation-resistant optical fibers under irradiation in a fission reactor

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.