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The Jules Horowitz Reactor (JHR) is a new material testing reactor under construction at CEA in Cadarache, France, to become one of the major infrastructures for scientific research on nuclear materials and fuels under irradiation as well as for medical isotope production. The reactor is designed to host various R&D programs dedicated to the optimization of operation of the existing nuclear power plants (NPPs), to assess the irradiation induced ageing of non-replaceable and safety related components in the operating NPPs, to support the improvement, development and deployment of the third generation of NPPs and small modular reactors (SMR) and to offer irradiation capabilities for GEN IV and fusion technologies. This paper will give the status of the construction of the experimental capacity and describe the works done by the International Consortium. Also, the ongoing operation planning and the future schemes for co-operation will be described.

The Jules Horowitz Reactor (JHR) is a new material testing reactor under construction at CEA in Cadarache, France, to become one of the major infrastructures for scientific research on nuclear materials and fuels under irradiation as well as for medical isotope production. The reactor is designed to host various R&D programs dedicated to the optimization of operation of the existing nuclear power plants (NPPs), to assess the irradiation induced ageing of non-replaceable and safety related components in the operating NPPs, to support the improvement, development and deployment of the third generation of NPPs and small modular reactors (SMR) and to offer irradiation capabilities for GEN IV and fusion technologies. This paper will give the status of the construction of the experimental capacity and describe the works done by the International Consortium. Also, the ongoing operation planning and the future schemes for co-operation will be described.

The Brazilian Multipurpose Reactor - RMB is a 30 MW pool type research reactor, that uses Materials Testing Reactor - MTR type fuel assemblies. It has a 5x5 square array core with 23 fuel assemblies and two in-core irradiation positions, operating with upward flow and average velocities nearly 10 m/s in the fuel plates channels. The IEA-R1 is a 5 MW pool type research reactor, which also has a 5x5 square array core with 19 standard fuel assemblies, four control fuel assemblies and a central beryllium irradiation device. It operates with downward flow nearly 1.8 m/s in the channels. In order to verify and provide data and information about the dynamical behavior of fuel assemblies under nominal and critical conditions, the experimental circuit ORQUÍDEA is being designed. This circuit will permit upward and downward flow and dynamical behavior of the fuel assemblies and its parts will be tested and verified. Flow rate, temperature, pressure and differential pressure transducers are the instruments of the circuit. Endurance and critical flow velocity tests will be performed. The COLIBRI experimental circuit is being designed to make tests that allow the studies of the fluid-structure phenomenology of fuel plates similar to those of the RMB fuel assemblies when subjected to high flow velocities, which can induce pressure differences between the channels formed by the fuel plates. This work presents a preliminary design for the ORQUÍDEA and COLIBRI experimental circuits to be built at the Instituto de Pesquisas Energéticas e Nucleares - IPEN of the Comissão Nacional de Energia Nuclear - CNEN.

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...

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.

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.

Up to now, predicting accurately the Fission Gas Release (FGR) from high burn up UO2 and/or MOX (Mixed Oxide) fuels at off-normal conditions, such as power transient, reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA), is still a significant and very challenging task. For this purpose, different R&D programs have been carried out in France, as well as in other countries. This has been done with a specific emphasis on mechanisms which promote the FGR under accidental conditions. These studies can be performed thanks to dedicated integral experiments conducted in-pile (i.e. in Materials Testing Reactor) with the corresponding cost and constraints, or at the laboratory scale with annealing tests which allow to be representative of specific parameters (thermal history for instance). During these annealing tests under well-known conditions (temperature, atmosphere), both the absolute level and the time dependence of the released gases should be monitored, together with the corresponding fuel micro-structural changes, since experimental knowledge of fission gas release alone is not efficient enough. This approach requires more and more accurate on-line measurements. This corresponds to the driving force of the present work. In this contribution, we will present our progress in developing an experimental platform that can submit nuclear fuel and cladding samples to annealing tests involving very high temperatures (up to 2500°C) and very fast temperature ramp (up to thousands of °C/s) with controlled thermal gradients and temporal dynamics. This new platform implements innovative instrumentation, such as optical diagnostics to measure fuel fragmentation kinetics and infrared pyrometry for temperature monitoring. This experiment is based on a high-power laser (1.5kW) coupled to an experimental chamber with controlled atmosphere (Ar, N2, or vacuum) and specific optical components. Based on the spatial beam profile and temporal power function of the laser, it is possible which such a system to produce complex spatio-temporal temperature gradients, relevant for addressing different research needs. It provides access to extreme conditions that are very difficult to reach with other means. Particularly, one of main objectives of this work is to investigate conditions of Reactivity Initiated Accident (RIA). The first experiments performed on inactive materials, non-irradiated uranium dioxide, is presented in order to highlight the capabilities of this technique.

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.

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.

The defect structures in Ni, Fe–15Cr–16Ni and Ti-added modified SUS316SS (modified SUS316) were examined after neutron irradiation below 0.3 dpa by the Japan Materials Testing Reactor and Belgian Reactor 2 to compare their defect structural evolution. The growth behaviour of interstitial-type dislocation loops (I-loops), stacking fault tetrahedra (SFTs) and voids was found to be quite different among these specimens. I-loops developed at lower temperatures in Ni than in Fe–15Cr–16Ni and modified SUS316, and more swelling occurred in Ni than in Fe–15Cr–16Ni. Finally, there were no voids in modified SUS316. These results were analysed in terms of the I-loop energy. A large discrepancy was found between the analytical results and experimental observations for Ni and modified SUS316, which suggests the formation of unfaulted I-loops directly from collision cascades. The growth of SFTs was detected in Fe–15Cr–16Ni and modified SUS316, and can be explained by a change in the dislocation bias of SFTs resulting f...

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.

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.

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.

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).

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.

The BR2 reactor, put into operation in 1963 and refurbished in 1995-1997, is a 100 MW high-flux 'materials testing reactor' which provides thermal neutron fluxes up to 10 15 n/cm 2 s. The production of radioisotopes, characterized by high specific activities, is carried out in dedicated in-core devices within standard irradiation cycles of 3-4 weeks and in accordance with a 'quality system' that has been certified to the requirements of the 'EN ISO 9001:2000' in November 2003. Due to its operating flexibility, its reliability and production capacity, the BR2 reactor is considered as a major facility for a routine supply of radioisotopes such as 99 Mo ( 99m Tc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 169 Er, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 79 Kr, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, etc. It will continue to play this key role with the possibility of an extra operating cycle from 2006 and the realization of various 'test' irradiations to develop the production of new radioisotopes. Some irradiation devices allow the loading and unloading of irradiated targets during the operation of the reactor. Hot cells and storage facilities are available to prepare and organize the shipment of the irradiated targets to dedicated processing facilities.

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.

Radiation induced changes in electrical properties of calcium-zirconium-indium oxide ceramics (CaZr0.9In0.1O3-d) were investigated using a fission reactor of Japan Materials Testing Reactor (JMTR). Electrical conductivity of CaZr0.9In0.1O3-d under the reactor irradiation increased by the Radiation Induced Conductivity (RIC), Radiation Induced Electromotive Force (RIEMIF)and Radiation Induced Electrical Degradation (RIED), caused by the electronic excitation. The RIC and the RIEMF decreased gradually for the constant value of the reactor power with the increase of irradiation time. Moreover, it was confirmed by Post-irradiation Examination (PIE) experiments that the initial electrical and thermal properties were completely degraded by the irradiation. These radiation effects are attributed to point defects or their clusters produced by atomic displacements caused by neutron collisions and also by the radiolysis.

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.

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.

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.

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.

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.