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A concept of a sodium cooled fast spectrum materials testing reactor is presented as one of the potential design configurations to achieve the performance versatility with respect to supporting a range of testing environments from current thermal-spectrum LWRs to Generation IV advanced fast-spectrum reactors. The objective is accomplished by designing a core with region-wise varying neutron spectra. This paper presents the design details and explores spectral variability examining feasibility to maintain thermal and fast energy spectra in the system. The core assemblies are designed using the EBR-II assembly as a prototype while implementing certain design modifications to meet performance needs. The fuel is 19.9%-enriched U-10Zr. Compare to EBR-II, the proposed core has larger pins, a taller active core, and thick reflector instead of depleted uranium breeding blankets. The MgO pins are analyzed as the reflector material to meet the needed versatility requirements. The design effort successfully concluded yielding a feasible 600 MW th -configuration with sufficient neutron fluxes in material testing locations. A fast flux greater than 5.6E15 n/cm 2 -s is maintained over the core lifetime in the central irradiation position. The volume of 63,000 cm 3 with a fast flux of ∼ 4.2E15 n/cm 2 -s or greater is provided for materials irradiation options. The core has regions of high magnitude thermal flux in a large graphite region at the core periphery. Spectral variability, design choices and reactor physics characteristics are discussed to demonstrate the concept viability.

A dedicated neutronics computational scheme is a crucial tool in the preparation and the post-analysis of an irradiation experiment, as it defines the key parameters of the experiment. The complex and very dynamic environment of a Materials Testing Reactor (MTR) core, on the other hand, represents a challenging object for neutronics calculations. Within the framework of the HELIOS experiment conducted at the High Flux Reactor (HFR) in Petten (The Netherlands) in 2009 under the European Project EUROTRANS, a new neutronics scheme has been developed with the objective of finding the optimum balance between accuracy and calculation performance. This scheme is based on the Monte-Carlo technique and employs a two-level approach by separating the problem into overlapping spatial and temporal domains. The underlying physical phenomena are presented, together with elements of validation which gives a further insight into the effects governing the neutron fluxes in the core.

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.

A prototype 3 He-based Passive Neutron Albedo Reactivity (PNAR) counter was developed and tested at Los Alamos National Laboratory (LANL) in collaboration with the Korea Atomic Energy Research Institute (KAERI) to measure the fissile content in electrochemical recycling (ER) product materials. The counter consists of 16 3 He cylindrical gas-filled proportional counters at 4 atm of pressure embedded in high-density polyethylene. In this work, experimental measurements were performed at LANL to characterize the performance of the PNAR counter using surrogate materials for the uranium metal ingot. The purpose of these experiments was to: 1) measure the operating and calibration parameters of the PNAR counter (e.g. efficiency profiles, coincidence gate fractions, die-away time) and 2) evaluate the accuracy and sensitivity of the PNAR method and the time correlated induced fission (TCIF) method for quantifying the 235 U mass in PWR fresh LEU fuel rods and Materials Testing Reactor (MTR) HEU fuel plates. A small 244 Cm reference source (13,373 n/s) was placed in the center of the fuel rods and fuel plates to simulate spontaneous fission from sub-ppm (parts per million) levels of Cm contamination in the U ingot. In order to compare the relative accuracy of the PNAR and TCIF methods for quantifying 235 U mass, calibration curves were generated for the net doubles rate and the doubles Cd ratio using the Deming software. The results from this experiment will be used to obtain a better understanding of the sensitivity of the PNAR and TCIF methods for samples with low neutron multiplication. Furthermore, this experimental measurement data will also help inform safeguards research and development (R&D) efforts on the viability of nondestructive assay (NDA) techniques and detector designs for quantifying fissile content in ER product materials. Future work will include performing measurements with the PNAR counter on small samples of U/TRU materials.

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.

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.

Vitrocerus is a processing method for MTR type fuel plates (Materials Testing Reactor) developed in the laboratory of Nuclear Materials at Bariloche Atomic Center which aims at producing a safe immobilization matrix for the potential provision of interim dry spent fuel from research reactors such type MTR.The present paper deals with the sintering of simulated spent fuel, that is composed of uranium silicide (U 3 Si 2 ) and aluminum, isotopically diluted in natural U3O8 and also with small additions of borosilicate glass (designated VG98/12), with composition: SiO 2 VG98/12 (56.7%), B 2 O 3 (12.4%), Al 2 O 3 (2.6%), Na 2 O (17.5%), CaO (4.1%), MgO(2.1%) and TiO 2 (4.6%) and Tg ∼ 520° C. The glass has the function to induce a liquid phase sintering at low temperatures, resulting in an enhancing of the hardness and densification of the final material. The simulated spent fuel is pre-oxidized before mixing with the glass.Studies were carried out on the thermal behaviour through dilatometry measurements, differential thermal analysis, hot-stage microscopy and X-ray diffraction, resulting in partial expansions at temperatures near 800°C. Then, these system probably densified by the action of a liquid phase from the glass VG98/12 or reaction with the spent fuel that will be immobilized.

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.

The Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency is a testing reactor with first criticality in March 1968. The reactor has been utilized 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. For the resumption of the new JMTR, the new irradiation facilities, the usability improvements, the target, and the expected roles of the new JMTR have been discussed. As one of the new irradiation facilities, the cryogenic irradiation facility, which is used for the investigation on the low-temperature irradiation behavior of materials such as superconducting magnet materials for fusion reactors, has been desired. In this study, the feasibility of low-temperature irradiation tests with the cryogenic irradiation facility was investigated, and the conceptual design of the facility was carried out. As a result, it was found that irradiation tests at temperatures below 20 K for the development of the superconducting magnet materials can be realized by the installation of an irradiation capsule into an irradiation hole with low γ heating and by the adoption of vacuum jacketed tubes to connect between the capsule and a cooling system.

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.

The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

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.

A small-sized motor with a resistance to neutron and gamma-ray radiation was developed. This motor was rotating normally at a gamma-ray dose of 7 × 10 8 Gy, a fast neutron ( E > 1 MeV) fluence of 2 × 10 21 m −2 and a thermal neutron ( E 22 m −2 . The gamma-ray dose for this motor was about 700 times as high as the specification of a commercial motor. Much progress in radiation resistance was attained by using radiation-proof materials without organic lubricant. It became clear that this small-sized motor is able to be used for in-pile tests of a fusion blanket using the Japan Materials Testing Reactor (JMTR).

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.

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.

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.

Li 2 TiO 3 is a candidate material for the tritium breeder blanket in the fusion reactor. The knowledge of Li 2 TiO 3 pebble bed properties is necessary to design the fusion blanket. In this study, the effective thermal diffusivity of Li 2 TiO 3 pebble bed under the neutron irradiation was evaluated. According to the irradiation test conducted in the Japan Materials Testing Reactor (JMTR), the effective thermal diffusivity of Li 2 TiO 3 pebble bed decreased with increasing the irradiation temperature and the thermal neutron fluence. However, horizontal temperature difference (DT) was almost constant at 400 °C under neutron irradiation up to 1×10 24 n/m 2 . Therefore, we can conclude that the effective thermal conductivity of Li 2 TiO 3 pebble bed is constant for thermal neutron fluence up to 1×10 24 n/m 2 at least.

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.

Irradiation tests of a mica substrate bolometer were carried out at the Japan Materials Testing Reactor up to the fast neutron fluence of 0.1 dpa. A performance of the bolometer was investigated during the irradiation. Significant increase of the meander resistance from 275 to 446 Ω was observed in the fast neutron fluence of 0.06 dpa. In the post-irradiation examination (PIE), it was confirmed that the gold meander contained 46% mercury which was produced by the nuclear transmutation reaction of 197Au(n, γ) 198Au→198Hg+β−. The bolometer sensitivity was measured by the change of the meander resistance for the input power. The sensitivity decreased slightly at constant reactor power. There was not significant change in the bolometer response time. During the cool down phase of the first cycle all connections went open circuit. Some recovered on start-up of the second cycle, however, all connections went open circuit again near the end of the second cycle. They were not recovered any more at the third cycle. Breaks of the gold meander were observed in the PIE. The use of gold meanders might be problematic in ITER.

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.

In order to conduct the integrated irradiation experiment of fusion blanket with testing reactor, it is indispensable to develop a small motor with high radiation resistance as the irradiation device. Therefore, the small motor was irradiated at the Japan Materials Testing Reactor. From this result of the irradiation test of a small motor, it became clear that the small motor was rotating normally up to the gamma-ray dose of 3.1×107 Gy, the fast neutron fluence of 2.8×1020 n/m2 and the thermal neutron fluence of 5.2×1021 n/m2 at 50 °C. The conductor resistance and the insulation resistance of polyimide coated wire, which did not change until the total gamma-ray dose of 3.1×108 Gy, the fast neutron fluence of 2.8×1021 n/m2 and the thermal neutron fluence of 5.2×1022 n/m2 at 50 °C, were 12 Ω and 1×108 Ω, respectively.

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.

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.

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.

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.

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.