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Activation analysis of impurities was performed on a test specimen of the reference low-activation vanadium alloy NIFS-HEAT-2 (V-4Cr-4Ti) which has been irradiated in Japan Materials Testing Reactor (JMTR) ~16 years before. Since the radioactive nuclides from V, Cr and Ti are produced with short half-lives, intense peaks of 60Co (T1/2 = 5.27 years) were dominant in the measured gamma-ray energy spectra and peaks of 94Nb (2.03 × 104 years) were also observed. The concentrations were evaluated to be 0.41 wppm and 0.71 wppm respectively from comparison with results of radioactivity calculation using the FISPACT-2005 code. In the analysis, detection of 108mAg (418 years), which would have a considerable influence on the material recycling after the usage in a fusion reactor, was intense motivation. Although the peaks of 108mAg has not been detected, the concentration of Ag was estimated to be

The performance of uranium-plutonium mixed carbide fuel was analyzed based on experimental data produced from the Japan Research Reactor No. 2, the Japan Materials Testing Reactor, and the Fast Flu...

The application of best-estimate codes [coupled neutron kinetics (NK)/thermal hydraulics (TH)] for safety analyses of research reactors (RRs) has gained considerable momentum during the past decade. Application of these codes is largely facilitated by the high level of technological maturity and expertise that these codes allow as a safety technology in nuclear power plants, and it is largely driven by International Atomic Energy Agency activities. The present study belongs in this framework and presents the development and application of the coupled NK and TH code THERMO-T to the analysis of protected reactivity insertion accidents and loss-of-flow accidents in a typical RR with standard Materials Testing Reactor plate-type fuel elements. The coupling is realized by considering the neutronic reactivity feedbacks of the fuel and coolant temperatures and a heat generation model for the reactor power. The neutron flux in the reactor core is solved by applying point reactor kinetic equations and employing radial and axial power distributions calculated from a three-dimensional full-core model by the continuous-energy Monte Carlo reactor physics code Serpent. The evolution of temporal and spatial distributions of the fuel, cladding, and coolant temperatures is calculated for all fuel channels by using a finite volume time implicit numerical scheme for solving a three-conservation equation model. In this study, additional features, such as critical heat flux ratio prediction and decay heat model, are implemented for both highly enriched uranium and low-enriched uranium cores, and a comprehensive comparison of THERMO-T results is performed against other codes.

A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

Data on the microstructural development of tungsten (W) and tungsten rhenium (Re) alloys were obtained after neutron irradiation at 400–800 °C in the Japan Materials Testing Reactor (JMTR), the experimental fast test reactor Joyo, and the High Flux Isotope Reactor (HFIR) for irradiation damage levels in the range of 0.09–1.54 displacement per atom (dpa). Microstructural observations showed that a small amount of Re (3–5%) in W–Re alloys is effective in suppressing void formation. In W–Re alloys with Re concentrations greater than 10%, acicular precipitates are the primary structural defects. In the HFIR-irradiated specimen, in which a large amount of Re was expected to be produced by the nuclear transmutation of W to Re because of the reactor's high thermal neutron flux, voids were not observed even in pure W. The synergistic effects of displacement damage and solid transmutation elements on microstructural development are discussed, and the microstructural development of tungsten materials utilized in fusion reactors is predicted.

Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.2 × 1020 n/cm2 (E > 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation, which was analyzed using Auger electron spectroscopy (AES) on intergranular facets, increased with increasing neutron fluence. A rate theory model based on four diffusion-reaction equations for substitutional P atoms, octahedral interstitial P atoms, vacancies, and self-interstitial atoms was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%, using parameters optimized by the present AES data. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.%, which is the maximum P concentration in RPV steels used in Japanese nuclear power plants intended for restart, was estimated to be less than 0.1 in monolayer coverage at 1 × 1020 n/cm2 (E > 1 MeV). A comparison of the PWR data with the MTR data, including that from the literature, showed that neutron flux had no effect upon grain-boundary P segregation for A533B steels. The relationships of the ductile-brittle transition temperature (DBTT) shifts to grain-boundary P segregation and to yield strength were also discussed. A linear relationship between the yield strength and the DBTT shift with a slope of 0.63 was obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs.

This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening.

To evaluate neutron-irradiation-induced crystalline defects and its thermal stability, α-SiC sintered bodies consisting of mainly 6H polytype were neutron irradiated using the Japan Materials Testing Reactor up to 1.9 × 1023 n/m2 (E > 0.1 MeV) at a low temperature of ∼423 K. Due to very low dose irradiation at low temperature, expected defects induced into crystalline lattice should only be simple point defects. Changes in the lattice parameters and macroscopic lengths resulting from post-irradiation isothermal annealing up to 6 h between room temperature and 1673 K were measured. Macroscopic volume swelling of α-SiC specimens coincided well to those of unit cell volume changes, indicating the presence of only simple defects. The recovery behavior of the macroscopic length and lattice parameters showed almost the same tendencies, although residual changes in the c-axis length slightly exceeded that of the a-axis length at temperatures lower than 1273 K. Difference in these axes lengths diminished at temperatures above 1273 K. Calculation of activation energies, obtained from precise length measurement using a dilatometer during each isothermal annealing step, revealed that length recovery behavior between 473 and 1573 K could almost be completely explained by a first-order reaction, and three stages with activation energies of 0.14, 0.26, and 1.13 eV. Changes in activation energy were related to the behavior of lattice parameter changes. The results indicated the presence of several kinds of point defects, some of which induced anisotropic lattice expansion.

A water chemistry research program using the in-pile loop in the Japan Materials Testing Reactor (JMTR) has been launched to develop data that are applicable for model verification as well as model...

Hot-pressed α- and β-SiAlON ceramics were neutron-irradiated in the Japan Materials Testing Reactor up to a fluence of 8.5 × 1024 n/m2 (E > 0.1 MeV) at 563 K. Irradiated SiAlON ceramics were isothermally and isochronally annealed up to 1473 K in a precision dilatometer, and macroscopic length change of α- and β-SiAlON ceramics were examined. During isochronal annealing, the recovery of both SiAlONs started at a temperature about 50–100 K higher than the irradiation temperature. Although the macroscopic length change for isochronal annealing decreased gradually up to 1473 K, the recovery was not completed. The macroscopic length change of β-SiAlON decreased more continuously than that of α-SiAlON. During isothermal annealing, the length gradually recovered as increasing annealing time at each annealing temperature. The activation energy of macroscopic volume recovery at lower temperature range was calculated as the first-order reaction. For α- and β-SiAlON from 723 to 1073 K, the values were 0.09 and 0.06 eV, respectively. On the other hand, for higher temperature range, the activation energy was calculated as the second-order reaction, and to be estimated as 0.49 eV for α-SiAlON from 1273 to 1373 K, and 0.33 eV for β-SiAlON from 1123 to 1473 K, respectively. These values suggested that the recovery process was governed by the recombination of different type of Frenkel pairs and depended on its polymorphs.

Aluminum nitride (AlN) is a candidate ceramic for electric or MHD insulator materials of liquid Pb cooled system of fusion reactors since it shows high electrical resistivity, high thermal conductivity, high strength and good compatibility with Pb alloys. In this study, recovery behavior of macroscopic length and lattice parameter of neutron-irradiated AlN was observed. Polycrystalline AlN specimens were neutron-irradiated in the Japan Materials Testing Reactor at the following conditions. (1) 4.4 × 10 24 n/m 2 ( E > 0.1 MeV) at 1193–1223 K (920–950 °C), (2) 8.8 × 10 24 n/m 2 at 1273 K (1000 °C) and (3) 2.8 × 10 24 n/m 2 at 573 K (300 °C). In the case of specimen irradiated to 2.8 × 10 24 n/m 2 at 573 K, macroscopic length slightly shrunk up to 1273 K, whereas the other specimens showed slight increase in length up to 1273 K. Rapid recovery of the macroscopic length was observed beyond ∼1273 K in all the specimens. Furthermore, beyond 1673 K, small increase in length was observed for the specimens irradiated to 4.4 × 10 24 and 8.8 × 10 24 n/m 2 , which should be caused by He bubble swelling. The lattice parameter of the specimens showed anisotropic swelling, i.e., expansion of the c -axis was significantly greater than that of the a -axis due to the irradiation in all the specimens. The recovery of the c -axis parameter was found to proceed above ∼1273 K, corresponding to the decrease in macroscopic length. The difference in macroscopic volume change and the unit-cell volume change can be attributed for the formation of microcrack along grain boundaries in the lower fluence specimen and possibly voids formation for the higher fluence specimens.

The Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency is a testing reactor for various neutron irradiation tests on nuclear fuels and materials, as well as for radioisotope production. The operation of JMTR stopped temporarily in August 2006 for refurbishment and improvement. The renewed JMTR will resume operation in Japanese fiscal year 2011. The renewal of aged reactor components, the preparation of new irradiation facilities, and the development of irradiation technologies have been carried out for the resumption of the new JMTR. The new JMTR with the new irradiation facilities and the irradiation technologies will be utilized for the research and development of fission and fusion reactor fuels and materials. This paper describes the present status of the refurbishment and the irradiation technologies focused on instrumentation such as the multi-paired thermocouple which is applicable to irradiation temperature control and a ceramic oxygen sensor in JMTR.

Innovative in-pile instrumentation is crucial for advanced experimental programs in research reactors. In this field, we developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in Material Testing Reactors (MTR). In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: 1) fuel temperature with a centreline thermocouple type C 2) internal pressure with a specific counter-pressure sensor, 3) fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector, which has been developed and is patent pending by CEA, SCK·CEN (Belgian Nuclear Research Center) and IES (French research laboratory of Montpellier II University and French National Research Center), is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced on-line from this measurement. The principle of this non destructive and on-line acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in MTR conditions. For this purpose some technical points have been studied in details: 1) fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, 2) size of the gas cavity to avoid any perturbation of the acoustic waves, 3) miniaturization of the sensor because of narrow in-pile experimental devices, 4) appropriate cables to transmit high frequency signal under nuclear conditions.

The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 1025 n/m2 under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specime...

Effects of neutron irradiation on microstructural evolution and radiation hardening were examined for fine-grained W–0.3 wt%TiC (grain size of 0.9 μm) and commercially available pure W (20 μm). Both materials were neutron irradiated at 563 K to 9 × 1023 n/m2 (E > 1 MeV) in the Japan Materials Testing Reactor (JMTR). Post-irradiation examinations showed that the microstructural changes and the degree of hardening due to irradiation were significantly reduced for fine-grained W–0.3TiC compared with pure W, demonstrating the significance of grain refinement to improve radiation resistance. In order to develop ultra-fine grained W–TiC compacts with nearly full densification, the fabrication process was modified, so that W–(0.3–0.7)%TiC with 0.06–0.2 μm grain size and 99% of relative density was fabricated. The achievable grain refinement depended on TiC content and milling atmosphere. The three-point bending fracture strength at room temperature for ultra-fine grained W–TiC compacts of powder milled in H2 reached approximately 1.6–2 GPa for composition near 0.5%TiC.

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core.

Effects of neutron irradiation on YAG laser welded V–4Cr–4Ti alloy were studied in JMTR (Japanese Materials Testing Reactor). The samples with and without post-weld heat treatment (PWHT) at 1073 K were irradiated in the temperature range of 673–873 K up to the dose of 0.45 dpa. After the irradiation, the microstructure and Vickers hardness of the welded samples were compared of the base metal, which were simultaneously irradiated at the same irradiation cycle.

In-pile flow-boiling experiments were performed to investigate the possible enhancement of heat transfer by the radiation induced surface activation (RISA) effect. The test section was a 2-mm diameter 100-mm long bore in a SUS-316L stainless steel block heated electrically. The test section, housed in an irradiation capsule, was inserted into one of the irradiation holes in the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). Quasi-steady state experiments were conducted before irradiation (out-of-pile and in-pile before reactor operation), during irradiation and after irradiation (in-pile), for the same boundary conditions using the same test section block. This approach allowed direct evaluation of the RISA effect through comparison of experimental data. Boiling curves were obtained up to the onset of dryout in an annular dispersed flow, for mass fluxes ranging from 180 to 630 kg/(m 2 s) under a fixed pressure of 420 kPa. The critical heat flux obtained during and after irradiation indicated an about 17% increase, on average, from that before irradiation. Meanwhile, the wall superheat at subcritical heat fluxes generally became greater than that before irradiation.

A small pebble has been selected as the shape of lithium titanate (Li 2 TiO 3 ) in the Japanese design of the fusion blanket. In the present study, fabrication tests of Li 2 TiO 3 pebbles were carried out by two kinds of indirect wet processes (a wet process using a dehydration reaction and another wet process using a substitution reaction), and the basic characteristics of each kind of Li 2 TiO 3 pebbles were evaluated for irradiation tests with the Japan Materials Testing Reactor (JMTR). From results of the fabrication tests and characterization, bright prospects were obtained concerning mass production of Li 2 TiO 3 pebbles with a target diameter of 0.2–2.0 mm and a target density of 80–85% theoretical density (TD) by these two kinds of indirect wet processes.

The dependence of tensile properties on tempering time and temperature was examined for a martensitic steel F82H irradiated at 250 °C to a neutron dose of 1.9 dpa in the Japan Materials Testing Reactor. The specimens were first normalized at 1040 °C for 0.5 h and tempered at 750 °C for 1 h. A second 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 second tempering time at 750 °C was varied from 0.5 to 10 h. The tensile specimens of F82H + 2Ni, pure iron and Fe + 0.1C were also irradiated. Tensile tests were carried out at room temperature. The radiation hardening of F82H was significantly relieved by the second heat treatment, and it was smaller than that of the first heat treated other steels. The radiation hardening depended on the tempering conditions and tended to increase with increasing the temperature and time of tempering.

A binary pebble bed of lithium titanate (Li 2 TiO 3 ) was irradiated in the Japan Materials Testing Reactor (JMTR), and its tritium recovery characteristics bed was studied under pulsed neutron operations. The temperature at the outside edge of the pebble bed increased from 300 to 350 °C immediately after the window of hafnium (Hf) neutron absorber was turned toward the reactor core, while the tritium recovery rate increased gradually. The ratio of tritium recovery rate to generation rate at the high-power, ( R / G ) high , approached the saturated value of unity at about 20 h of operation. Overall tritium recovery behavior under the pulsed operation was similar to that under the steady state power operation. An estimated time constant of about 3 h for the tritium recovery was much longer than the thermal time constant of about 100 s.

Radiation effects in components and materials will be one of the most serious technological issues in nuclear fusion systems realizing burning-plasmas. Especially, diagnostic components, which should play a crucial role in controlling plasmas and understanding the physics of burning-plasmas, will be exposed to high-flux neutrons and gamma rays. Dynamic radiation effects will affect the performance of components substantially from the beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their service. High-power-density fission reactors will be the only realistic tools to simulate the radiation environments expected to occur in burning-plasma fusion machines such as the International Thermonuclear Experimental Reactor (ITER), at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibres, were irradiation-tested in a fission reactor, to evaluate their performances in heavy radiation environments. Results indicate that ITER-relevant radiation-resistant diagnostic components could be developed in time, although there are still some technological problems to be overcome.

In order to investigate effects of neutron irradiation on fatigue properties of nuclear materials, a remote-controlled high temperature fatigue test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). A small-sized fatigue specimen having double blades to measure strain with a laser extensometer was designed for this machine. A strain amplitude in fatigue tests of a completely reversed push-pull type using a triangular wave was controlled with an accuracy of ±3% of the total strain range during test. Low cycle fatigue tests of type 304 stainless steel irradiated in JMTR at 823 K up to a fast neutron fluence of 1×1025 n/m2 (E>1 MeV) were performed in total strain ranges of 0.7–1.4% at 823 K using the designed small-sized specimens.

To develop a non-destructive testing technique for evaluating embrittlement of irradiated materials, the correlation between ultrasonic characteristics and embrittlement was investigated from the results of the ultrasonic wave measurement and the Charpy impact test of irradiated specimens of commercial A533B-1 steel and welded material at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). After irradiation at 523 or 563 K up to a fast neutron fluence of 1×1024 N/m2 (E>1 MeV), velocities of both shear and longitudinal waves in the irradiated specimen were lower than those in the unirradiated one. The decrease in the velocities may be caused by the reductions of the shear and Young’s moduli in the irradiated specimen. The attenuation coefficient of the longitudinal wave in the irradiated specimens increased compared with unirradiated ones. With increasing the shift amount of the Charpy transition temperature at 41 J absorbed energy, the velocity and attenuation coefficient of the ultrasonic waves decreased and increased, respectively.

Lithium titanate (Li 2 TiO 3 ) is one of the candidate materials as tritium breeder in the breeding blanket of fusion reactors, and it is necessary to show the tritium release behavior of Li 2 TiO 3 pebble beds. Therefore, a blanket in-pile mockup was developed and in situ tritium release experiments with the Li 2 TiO 3 pebble bed were carried out in the Japan Materials Testing Reactor. In this study, the relationship between tritium release behavior from Li 2 TiO 3 pebble beds and effects of various parameters were evaluated. The ( R / G ) ratio of tritium release ( R ) and tritium generation ( G ) was saturated when the temperature at the outside edge of the Li 2 TiO 3 pebble bed became 300 °C. The tritium release amount increased cycle by cycle and saturated after about 20 pulse operations.

Fused silica core optical fibers are expected to play crucial roles especially in the size-reduced International Thermonuclear Experimental Reactor (ITER-FEAT). Several radiation resistant optical fibers have been developed in Japan and the Russian Federation. The task force on radiation effects in diagnostic components in the ITER-EDA (engineering and design activity) promoted international round-robin irradiation experiments on the developed optical fibers. Ten different optical fibers were tested in a cobalt-60 gamma-ray irradiation facility and in the Japan Materials Testing Reactor. The paper reports results obtained on five different optical fibers, which include purified, hydrogen loaded, and fluorine doped ones. Results show that the developed optical fibers could be deployed in remote handling and out-of-vessel applications. But, for the in-vessel diagnostics in the visible range optical spectroscopy, further improvement of the radiation resistance of optical fibers will be needed.

Stress relaxation of tensile type specimens under fast neutron irradiation at 288 °C has been studied for 316L stainless steel (SS) in the Japan Materials Testing Reactor. In-pile stress-relaxation tests were carried out at fast neutron fluence levels of 1.3×10 24 , 5.5×10 24 and 1.5×10 25 n/m 2 ( E >1 MeV). These tests were carried out at the applied total strain levels of 0.06%, 0.1%, 0.3% and 0.75%. In order to evaluate the thermal stress-relaxation behavior and to distinguish it from the irradiation induced stress-relaxation behavior, out-of-pile stress-relaxation tests were also performed at 288 °C in air using an electric furnace. This paper describes results of in-pile and out-of-pile stress-relaxation tests on 316L SS tensile specimens. These results are compared with the literature data by Foster et al. [J. Nucl. Mater. 252 (1998) 89] which were mainly obtained from bend beam specimens. Moreover, these experimental results are compared with analytical results obtained using Nagakawa's model [J. Nucl. Mater. 212–215 (1994) 541].

In the Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using the Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensors under irradiation environment, heat-up and performance tests at elevated temperatures before irradiation and in-pile tests were performed in JMTR. It was determined that it is possible to measure strain under irradiation environment below 1×1023 n m−2 (E>1 MeV) by a fiber optic grating sensor, because in-pile temperature characteristics were in good agreement with out-of-pile test results.

Gamma dose-rate profiles have been measured by means of a self-powered gamma detector (SPGD) in a gamma irradiation facility using /sup 60/Co sources, situated in the side pool of the BR2 materials testing reactor. Because of its small size (diameter of only 4 mm), the SPGD is ideally suited to perform very localized gamma dose-rate measurements. It has been shown that the SPGD can provide precise relative measurements of the local gamma dose rate in the range 0.5-20 kGy/h. A very detailed Monte Carlo code has been developed to relate the gamma dose rate (for a given gamma spectrum) to the current induced by the SPGD. The obtained conversion factor could thus be used to determine the absolute gamma dose rate.

A `common basis' systems study of superconducting (SC) and normal conducting (NC) DT burning fusion power and materials testing reactor designs is presented. The figures of merit for power and materials testing reactors are, respectively, projected cost of electricity (COE) and direct cost. A common 0-D plasma modelling basis is used and the plasma geometry and engineering aspects of the SC and NC designs are treated in an equivalent manner that is consistent with the limitations of their respective magnet technologies. Aspect ratios A in the range 1.2≤A≤6 and plasma elongations κ in the range 1.5≤κ≤3 are explored and an MHD stability (beta limit) physics basis that accurately describes the increase of normalized beta βN and toroidal beta βT with decreasing A and/or increasing κ is incorporated. With this MHD basis taken into account and with the usual reactor geometry, physics and engineering constraints and costing bases applied, the results of the study show that for SC power reactor designs with κ = 2 the COE has a minimum for 2≤A≤3 and increases with a further increase in A(A>3). For NC power reactors the COE has minima at A≈2. For both SC and NC power reactors, the minima are more apparent with lower κ. While SC options appear to offer lower COE for power plants, the direct cost for NC test reactors with similar fusion power output is significantly lower. Within the NC category, test designs that combine modest A and maximum elongation show promise for achieving ITER-like testing capabilities at low direct cost. For example, an NC coil design with A = 2, κ = 3 could produce fusion power of 200 MW at 1.23 MW/m2 average neutron wall loading at a total direct cost of about 650 million US dollars. This NC design with a fissile blanket could also convert about 1270 kg of fission reactor waste per full power year. A possible cost effective development scenario for fusion power is identified for NC A = 2 toroidal devices for physics and material testing studies for use in the near future. The selection of the SC or NC coil option could then be made for the construction of the demonstration power reactor.

To investigate the microstructure evolution during temperature change irradiation, Ni and its binary alloys were irradiated in the Japan Materials Testing Reactor to doses up to 0.46 dpa under two cycles of temperature change of 533/693 K, i.e., 533 K, then 693 K, then 533 K, then 693 K. The irradiation dose at each temperature stage was about the same. The specimen sets were pulled out five times during irradiation. In the beginning of the first irradiation at the high temperature of 693 K after irradiation at 533 K, the densities of interstitial type dislocation loops and voids decreased in pure Ni, Ni–2at.%Cu and Ni–2at.%Ge alloys. In the second irradiation at 533 K, however, the formation of dislocation loops and voids was promoted. The void density continued to increase even after the second irradiation at 693 K. On the other hand, in Ni–2at.%Si and Ni–2at.%Sn alloys, where the formation of voids was suppressed by addition of the minor elements Si or Sn, the microstructure evolution was almost independent of irradiation temperature.

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 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.

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.

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.

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.