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A key conception of the remote maintenance (RM) method is presented of the target assembly (TA) of Advanced Fusion Neutron Source (A-FNS), that is an accelerator-driven fusion neutron source utilizing the Li(d,xn) nuclear reactions. The RM method, newly proposed here, is aimed at (i) enhancing the modularity of the TA RM and (ii) keeping the compatibility with the materials irradiation tests, compared to the preceding TA RM design of International Fusion Material Irradiation Facility (IFMIF). A 3-dimensional kinematics simulation indicates that the large clearance of 52 mm between the TA and high flux test module (HFTM) brings the independent removal from and the installation in the test cell, regardless of the presence of the test modules in the test cell. A neutronics performance of the HFTM with the enlarged clearance is also studied. The gradient of the calculated neutron irradiation damage is found to be decreased and satisfies a users’ requirement of the irradiation environment when the clearance is enlarged from 2 to 52 mm. (The former value was the IFMIF engineering design value.) These simulation results suggest that the proposed TA RM method will enhance not only the modularity of the TA, but also will improve the materials irradiation test environment.

Using neutrons generated by d-Li reaction in the A-FNS, we analyzed radionuclidic purity of technetium-99m (99mTc). It is one of the most-used medical isotopes and the daughter radionuclide of molybdenum-99. The angular distribution of the neutrons by the d-Li reaction is anisotropic, and the high energy neutrons around 14 MeV have strong angular dependence in the forward direction. The high energy neutrons produce various radioisotopes, because their energies are over threshold energies of nuclear reactions. We calculated neutron spectra in the A-FNS test cell by using the Monte Carlo transport code MCNP5 with the extension McDeLicious-11 and the nuclear data library FENDL-3.1d. Using the spectra, activation calculations were performed with the inventory calculation code FISPACT-2010. Ratios of the activity of 99mTc to that of other Tc radioisotopes were evaluated for the cooling time after the neutron irradiation. It was found that the purity met a medical demand after an adequate cooling time because of decrease due to the decay of radioisotopes. In addition, we clarified that the purity became higher after a chemical separation process of Tc.

The alleviated limits of impurities of nitrogen and hydrogen isotopes in lithium are suggested to be 400wppm and 550appm, respectively for early realization of Advanced Fusion Neutron Source planned in Japan. The applicability of the suggestion is discussed by comparing to that of IFMIF. As the results, not less than 4-year buffer for R&D of nitrogen trap are brought in, which is thought as one of bottle neck. Additionally, the application of this suggestion would diminish the beryllium (a radioactive isotope will generate in A-FNS process) concentration in processing lithium, as well. On the other hand, it is indicated that there would be no fatal problems by application of the present suggestion.

The latest beta version of ENDF/B-VIII, ENDF/B-VIIIβ4 was released in February 2017 from the Cross Section Evaluation Working Group. We benchmarked the iron and concrete shielding experiments with 40 and 65 MeV neutrons at QST/TIARA by using ENDF/B-VII.1, JENDL-4.0/HE and ENDF/B-VIIIβ4 to investigate whether the 56Fe and 16O data in ENDF/B-VII.1 were properly improved in ENDF/B-VIIIβ4 or not. The drastic overestimation in the calculated result with ENDF/B-VII.1 was resolved in that with ENDF/B-VIIIβ4 in the iron experiment with 40 MeV neutrons, because of the revision of the non-elastic scattering cross section data of 56Fe in ENDF/B-VIIIβ4. However, the calculated result with ENDF/B-VIIIb4 significantly underestimated the measured one in the experiment with 65 MeV neutrons. The drastic overestimation in the calculated results with ENDF/B-VII.1 still appeared in those with ENDF/B-VIIIβ4 in the concrete experiments with 40 and 65 MeV neutrons, due to no improvement of 16O data particularly above 20 MeV in ENDF/B-VIIIβ4. We figured out that the effect of the 16O data replacement was very large and those of 28Si and 27Al data replacements were also not so small. The official release of ENDF/B-VIII should be produced based on this study.

Accelerator-driven fusion neutron sources, e.g. Advanced Fusion Neutron Source (A-FNS) and DEMO Oriented Neutron Source (IFMIF-DONES), are being designed, based on the achievements and lessons learnt from the IFMIF/EVEDA project. The purpose of the study is to analytically evaluate the deformation of the lithium free surface, potentially created by the intense beam injection. We have developed a two-dimensional (2-D) fluid model of the “beam-on” target of an IFMIF-type fusion neutron source by extending the phenomenological 2-D model of the Li target flow of the EVEDA Lithium Test Loop (ELTL). We have derived, from the newly developed model, a practical and analytical formula of the thickness of the “beam-on” target in the flow direction with accounting for the beam injection effect. We have also evaluated the thickness of the beam-on target of A-FNS as an example of application of the derived formula. The evaluation indicates that under the reference design conditions of the accelerator and beam-on Li target of A-FNS, the deformation of the free surface of the target by the beam injection is negligibly small. The analytical results presented here may be used for a benchmark of more complicated CFD simulation for the “beam-on” target design of IFMIF-type fusion neutron sources like A-FNS and IFMIF-DONES. Keywords: IFMIF-type fusion neutron source, Design, Liquid lithium target, Beam injection, Free surface, Modeling

A plan of an advanced fusion neutron source (A-FNS) by using d-Li reaction is in progress at Rokkasho in Japan. We investigate multipurpose usages of the A-FNS in addition to fusion material irradiation test. Production of medical isotope 99Mo is considered as one of the usages. We conducted a conceptual study on a module for radioisotope production which was composed of a neutron spectrum shifter and a neutron reflector. We examined impacts of materials of the shifter and reflector on amounts of the 99Mo production, and their thicknesses. It was concluded that beryllium is the most suitable material both for the shifter and the reflector from the viewpoint of the 99Mo production. It was shown that we produced an enough amount of the 99Mo for the demand in Japan. We can apply natural molybdenum for this purpose. It was also shown that we could use a part of irradiation capsules in high flux test module, which was for the fusion material irradiation test originally, by using isotopically enriched 100Mo to meet that. Keywords: d-Li reaction, Fusion neutron source, Medical isotope, 98Mo(n,γ)99Mo, 100Mo(n,2n)99Mo, A-FNS

Liquid lithium, a candidate material for the target of International Fusion Material Irradiation Facility (IFMIF), easily contains nitrogen. Nitrogen in liquid lithium should be removed as it accelerates corrosion of tubing materials. Hot trapping by iron-titanium alloys is expected to be a practical method. In order to understand the characteristics of nitrogen trapping with iron-titanium alloys in liquid lithium, nitrogen distribution was observed by Soft X-ray Emission Spectroscopy (SXES). Samples of Fe-5Ti alloys were immersed for 9–256 h in liquid lithium which contains 1000 w-ppm of nitrogen at 823 K and studied by X-ray Powder diffraction and SXES. The results show that nitrogen is trapped mainly in the grain boundaries and on the surfaces of the Fe-5Ti alloys and diffuses in Fe-5Ti much faster than in pure titanium. In addition, it is assumed that the amount of nitrogen trapping by the surface of Fe-5Ti alloys does not depend on the immersion time.

Advanced Fusion Neutron Source (A-FNS) [11] is an accelerator-based neutron source utilizing Li(d,xn) nuclear stripping reactions to simulate D-T fusion neutrons for testing and qualifying structural and functional materials of fusion reactor components, which is to be constructed at the Rokkasho site of National Institutes for Quantum and Radiological Science and Technology, Japan, in the near future. The purpose of the study reported here is to demonstrate the ultimate safety margins of A-FNS in the worst case of release of radioactive materials outside the A-FNS confinement system. For this purpose, we analyzed a ‘bounding event’ postulated in A-FNS. The postulated event sequence consists of fire of the purification system of the liquid Li loop during the maintenance, of mobilization of the tritium and 7 Be, which are the impurities of the loop, and of the entire loss of confinement of the radioactive materials. We have calculated the early doses to the public due to the release of the tritium and 7 Be source terms to the environment. The UFOTRI/COSYMA simulations have been performed considering the site boundary of 500 m away from the facility. The obtained results indicate that the early dose is below the level that requires the emergent public evacuation. Such results demonstrate that the A-FNS complies with the defined safety objective against its radiation hazard. The simulation results suggest that the inherent, ultimate safety characteristic found by this study may assist a licensing process for installation of A-FNS.

Liquid lithium (Li) jet is planned as a beam target in fusion neutron sources developed in Japan and EU. We have been studying the hydrodynamic characteristics of the Li jet experimentally and numerically. For the safety and the efficiency of fusion neutron sources, it is necessary to understand the heat transfer inside the Li target irradiated by ion beams. In this paper, in order to evaluate the heat transfer under the turbulent flow with complex vortex structure inside the Li jet, 3-dimensional simulation of it with beam irradiation is conducted using LES. As the first step of this study, the Li jet is modeled as single-phase for reducing the computational load, in which the Li surface is treated as the free-slip boundary condition. Simulation model is large-scale with over thirty million meshes for resolving the detailed vortex structure. In this simulation, it was found that the temperature peak was disturbed by wall turbulence near the bottom wall of the flow channel.

We had pointed out that the elastic scattering and capture reaction data of the recent nuclear data libraries for copper had included some problems in the resonance energy region, which had caused a large underestimation of reaction rates of non-threshold reactions. In order to corroborate this issue, we carried out a new benchmark experiment on copper in the neutron field with more low energy neutrons. We measured reaction rates using the activation foils and analyzed the experiment with MCNP5.140 and ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0. As a result, the calculated reaction rates related to low energy neutrons excessively underestimated the measured ones as in the previous experiment. The calculated reaction rates with the modified copper nuclear data in the previous study reproduced the measured ones well. However, it was revealed that the modified copper data needed more revision.

KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

An integral experiment on molybdenum is performed with a DT neutron source at JAEA/FNS. A Mo assembly is covered with lithium oxide blocks in order to reduce background neutrons inside the assembly. Several reaction rates and fission rates are measured along the central axis inside the assembly and compared with calculated ones with the Monte Carlo transport code MCNP5-1.40 and recent nuclear data libraries of ENDF/B-VII.1, JENDL-4.0, and JEFF-3.2. The calculated results generally show underestimation. From our detailed analysis, it is concluded that the (n,2n) cross section data of all the Mo stable isotopes in JEFF-3.2 are more suitable than those in JENDL-4.0 and the (n,γ) cross section data of 92Mo, 94Mo, 95Mo, 96Mo, 97Mo, and 100Mo in JENDL-4.0 are overestimated.

Fusion reactors require advanced neutron multipliers with high stability at high temperatures. Beryllium intermetallic compounds (beryllides) have a universally robust potential for high temperature use. Fabrication of beryllide pebbles of Be–Zr compositions was conducted from the viewpoint of pebble mass production. Prototypic pebbles were successfully fabricated using the plasma-sintered electrodes with high Be and Zr contents for enhanced of the thermal shock resistivity of the electrodes during granulation by a rotating electrode method. From the results of granulation examinations, it was revealed that granulation yields varied greatly depending on compositions of the plasma-sintered electrodes themselves. By utilizing plasma-sintered electrodes with high Be and Zr contents, Be13Zr pebbles had a high granulation yield of 84%. Moreover, the Be13Zr pebbles displayed better oxidation properties as compared to pure Be pebbles.

The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

In order to validate a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), not only for DT neutrons but also for neutrons with energy of less than 14 MeV, we perform an integral test with a Li 2 O rectangular assembly of 60.7 cm in thickness and a DT neutron source at JAEA/FNS. We place a lot of activation foils at depths of 10.1 cm and 30.4 cm for measurements of dosimetry reaction rates in small space along the central axis in the assembly, measure decay gamma-rays from the activation foils with high-purity Ge detectors after the DT neutron irradiation by the foil activation technique, and deduce a variety of dosimetry reaction rates. We calculate the reaction rates by using a Monte Carlo code MCNP5-1.40 and the nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.05 as the response functions for the dosimetry reactions. The calculation results generally show good agreements with the measured ones, and it can be confirmed that most of the data in IRDFF-v.1.05 are valid for the neutron field in the Li 2 O assembly with the DT neutrons.

The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in Rokkasho, Japan, and the deuteron beam commissioning of the injector began in July 2015. Due to the huge beam current of 125 mA, a large amount of d-D neutrons are produced in the commissioning. The neutron streaming effect through pipe penetrations and underground pits may dominate the radiation dose at the outside of the accelerator vault during the injector operation. In the present study the effective dose rate expected during the injector commissioning was analyzed by a Monte Carlo calculation and compared with the measured value. For the comparison it is necessary to know the total neutron production yield in the accelerator vault, thus a dedicated neutron production yield monitoring system was developed. The yield obtained was smaller than that previously reported in a literature by a factor of a few and seems to depend on some beam conditions. From the comparison it was proved that the calculation always provides a conservative estimate and the dose rates in places where occupational works can always access and the controlled area boundary are expected to be far less than the legal criteria throughout the injector commissioning.

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

In order to validate copper nuclear data, an integral experiment on copper with the DT neutron source at JAEA/FNS had been performed over 20 years ago. The experiment had showed that ratios of the calculated values to the experimental ones (C/Es) related to lower energy neutrons had been drastically smaller than unity. In order to reveal reasons of the small C/Es, we newly performed the integral experiment on copper with the DT neutron source at JAEA/FNS. A quasi-cylindrical copper assembly of 315 mm in radius and 608 mm in depth was covered with Li 2 O blocks of 51 mm in thickness for the front and side parts and 153 mm in thickness for the rear part to exclude background neutrons which might affect the measured data. We measured reaction rates with 5 activation foils and fission rates with 2 micro fission chambers at the center of the assembly. The experiment was analyzed by using MCNP5-1.40 with the recent nuclear data libraries, ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0. As a result, the C/E of the reaction rate of the 197 Au(n,γ) 198 Au reaction improved by 10% from the previous result and the combination of the 63 Cu data in JEFF-3.2 and 65 Cu data in JENDL-4.0 increased the C/E by more 10% because of the resonance data of the 63 Cu in JEFF-3.2. Moreover, the calculated result with the 63 Cu data in JEFF-3.2 and 65 Cu data in JENDL-4.0 with 10% larger elastic scattering cross section data and 10% smaller capture cross section data between 100 eV and 0.3 MeV agreed with the measured one related to lower energy neutrons very well. The copper data should be reassessed.

Li2TiO3 with excess Li (Li2+xTiO3+y) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li2CO3, which promotes grain growth. To inhibit the generation of Li2CO3, calcined Li2+xTiO3+y pebbles were sintered under vacuum and subsequently under a 1% H2–He atmosphere. The average grain size of the sintered Li2+xTiO3+y pebbles was less than 5 μm. Furthermore, the tritium release properties of Li2+xTiO3+y pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H2–He purge gas was passed through the Li2+xTiO3+y pebbles. The Li2+xTiO3+y pebbles exhibited good tritium release properties, similar to those of Li2TiO3 pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li2TiO3 during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li2TiO3. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li2TiO3 as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li2TiO3 by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li2TiO3 irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li2TiO3 was the dominant parameter to control the chemical form of tritium released from the Li2TiO3.

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.

Liquid lithium (Li) jet is planned as a beam target in fusion neutron sources (FNSs), such as IFMIF in Japan and EU, A-FNS in Japan and IFMIF-DONES in EU. For the safety and the efficiency of such FNSs, it is desirable to keep the high-speed Li jet stable. In addition to many experimental researches, numerical approaches using CFD simulation have been also required to evaluate the detailed flow pattern inner the Li jet because of the opacity. In our previous simulation, it was found that the vortex structure under the free surface of the Li jet had the strong influence on the surface fluctuation using LES simulation. In this study, for evaluating the interaction between the Li surface behavior and the flow pattern inner the Li jet, two simulations are conducted: one is the LES simulation considering wall roughness, and another one is the RANS simulation using full-scale model with the nozzle and the Li jet. The influence of wall roughness on the vortex structure becomes an issue in long-term operation of actual FNSs, and the influence of the secondary flow generated due to the side wall on the surface shape is evaluated in the latter simulation.

An accelerator based intense neutron sources such as the IFMIF, A-FNS and DONES produce neutron flux by a stripping reaction between lithium (Li) and deuteron for testing various materials for fusion reactor. In these fusion neutron sources, a liquid Li free-surface jet is applied as the liquid Li target. For the development of the Li target system, the variation in the thickness of the Li jet was measured using the Li loop of Osaka University. At the Osaka Univ., a lot of knowledge about the surface fluctuation of the Li jet had been obtained by using the electro-contact type distance meter (referred to as the probe apparatus herein). Besides, for the stable and safety operation of the facility, the probe apparatus is planned to be used as a safety interlock system for beam irradiation in actual intense neutron source because of its simple structure. Therefore, it is desirable to verify its accuracy. Especially, it is expected that false detection and undetected waves, which is particular to the contact-type apparatus, could cause errors. Then in this study, the measurement using non-contact type distance meter using optical comb (referred to as the laser probe herein) was conducted, and then the error of the results of the probe apparatus is evaluated by comparing results.

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests.

We have conducted a conceptual design on the test modules for Advanced Fusion Neutron Source A-FNS. Fusion neutron irradiation tests are performed using a variety of the test modules in A-FNS. We acquire the irradiation data on the fusion reactor materials, based on which the construction phase of the fusion DEMO DT reactor is to be decided. We established basic specifications and concepts on the test modules. Structural design on the test module was conducted taking into account the replacement with the remote handling, and we established the basic structural concept and the replacement methodology. By applying the methodology in this study, only capsules can be easily replaced from the test module. Nuclear analyses were performed for the test module. The maximum value of the neutron displacement damage on the blanket structural material test specimen is 10 dpa in accumulated irradiation for one year, and we can acquire the irradiation data on the cumulative damage of 20 dpa, which is requirement value as the initial irradiation data for the construction of a DEMO reactor, in accumulated irradiation for two years. Thermal analyses were conducted for the test specimens in irradiation, and it was confirmed their temperatures could satisfy the design specifications.

We perform an integral test of a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), which has been released from the International Atomic Energy Agency recently. A pseudo-cylindrical graphite assembly is used with a DT neutron source at the fusion neutronics source facility of Japan Atomic Energy Agency. A lot of foils for measurements of dosimetry reaction rates are placed in the small space along the central axis in the assembly. After the DT neutron irradiation, forty dosimetry reaction rates are measured and are compared with calculated values, which are obtained by using a Monte Carlo neutron transport code MCNP5-1.40 and a nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.03 as the response functions for the dosimetry reactions. The calculated results generally show good agreements with the experimental ones. The cross section data of the 68Zn(n,np)67Cu and 68Zn(n,d)67Cu reactions and the 113In(n,γ)114mIn reaction, which are not included in IRDFF-v1.03, are needed for the analysis of the 67Zn(n,p)67Cu reaction and the 115In(n,2n)114mIn reaction as the competitive reactions, respectively.

The temperature of Li 2 TiO 3 pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li 2 TiO 3 pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li 2 TiO 3 pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li 2 TiO 3 layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H 2 /He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H 2 O/He gas (H 2 O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H 2 , it is indicated that the increasing tendency of HT release is similar to that of the dry H 2 /He.

IAEA released a new Fusion Evaluated Nuclear Data Library, FENDL-3.0, in 2012. FENDL-3.0 extends the neutron energy range from 20 MeV to greater than 60 MeV. Now there is increasing interest in nuclear data above 20 MeV. Thus we have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which are supplied from JAEA, LANL and IAEA, respectively, were used for this analysis. The collimated neutron beam and test shields were modeled in the analysis. The measured source neutron data were adopted in the analysis. The followings are found out from the results; (1) Iron experiments: The calculation result with FENDL-3.0 agrees with the measured one best. That with JENDL/HE-2007 fairly agrees with the measured one. On the contrary that with ENDF/B-VII.1 drastically overestimates the measured one. It is confirmed that this overestimation is due to the smaller non-elastic scattering data of 56Fe in ENDF/B-VII.1. (2) Concrete experiments: The calculation result with ENDL/HE-2007 agrees with the measured one best, while those with FENDL-3.0 and ENDF/B-VII.1 drastically overestimate the measured one. It is confirmed that this overestimation is due to both the larger elastic and smaller non-elastic scattering data of 16O in FENDL-3.0 and ENDF/B-VII.1.

We have fabricated a particle detector using single crystal diamond grown by chemical vapor deposition. The irradiation dose dependence of the output pulse height from the diamond detector was measured using 3 MeV protons. The pulse height of the output signals from the diamond detector decreases as the amount of irradiation increases at count rates of 1.6–8.9 kcps because of polarization effects inside the diamond crystal. The polarization effect can be cancelled by applying a reverse bias voltage, which restores the pulse heights. Additionally, the radiation hardness performance for MeV energy protons was compared with that of a silicon surface barrier detector.

In the Broader Approach (BA) framework, research and development on blanket related materials and technologies have been carried out between the EU and Japan. Those activities are implemented mainly at the Rokkasho BA site in collaboration with universities in Japan. In the R&D on SiCf/SiC composites for an advanced blanket material, mechanical properties of chemically vapor infiltrated (CVI) SiCf/SiC composites have been obtained in high temperature vacuum environment up to 1000 °C. As to R&D on the tritium technology, tritium retention of the fine-grained re-crystallized tungsten has been evaluated. On reduced activation of ferritic/martensitic (RAFM) steels as the blanket structural material, a 20-ton heat of the F82H RAFM steel has been successfully produced in an electric arc furnace. Advanced neutron multiplier pebbles of beryllide (beryllium–titanium alloy) have been fabricated using a dedicated rotating electrode apparatus followed by annealing from the beryllide rod produced with the plasma sintering method. Also advanced tritium breeder pebbles of lithium-rich lithium titanate have been fabricated by an emulsion method, where the grain size, confirmed by SEM, was less than 5 μm. Tritium recovery tests of advanced tritium breeder pebbles will be investigated under 14 MeV neutron irradiation as one of the new activities.

The IAEA supports and promotes the gathering of the best data from evaluated nuclear data libraries for each nucleus involved in fusion reactor applications and compiles these data as FENDL. In 2012, the IAEA released a major update to FENDL, FENDL-3.0, which extends the neutron energy range from 20 MeV to greater than 60 MeV for 180 nuclei. We have benchmarked FENDL-3.0 versus in situ and TOF experiments using the DT neutron source at FNS at the JAEA and TOF experiments using the DT neutron source at OKTAVIAN at Osaka University in Japan. The Monte Carlo code MCNP-5 and the ACE file of FENDL-3.0 supplied from the IAEA were used for the calculations. The results were compared with measured ones and those obtained using the previous version, FENDL-2.1, and the latest version, JENDL-4.0. It is concluded that FENDL-3.0 is as accurate as or more so than FENDL-2.1 and JENDL-4.0, although some data in FENDL-3.0 may be problematic.

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

The Li2TiO3 is one of candidate breeding materials of a water cooled ceramic breeding (WCCB) blanket. In order to clarify the tritium recovery property of the WCCB blanket with the Li2TiO3 breeding material, we have performed the tritium recovery online experiment with the DT neutron source at the Fusion Neutronics Source facility in Japan Atomic Energy Agency (JAEA-FNS). We irradiated an experimental assembly simulating the WCCB blanket and recovered the tritium recovered from the Li2TiO3 pebbles put into the assembly with a heater system, sweep gases and bubblers. The activity of the recovered tritium was measured with a liquid scintillation counter. From our tritium recovery online experiment and calculation, the followings were found out: (1) the recovered tritium corresponded to the calculated tritium production within the experimental error in the range of 573–1073 K and (2) the recovered HTO tended to be easily recovered at lower temperature and high water moisture. The recovered HT increases at higher temperature and dry hydrogen circumstance. However, the maximum level of the tritium gas recovery is around 90% even at higher temperature and 1% H2 circumstance.

To evaluate the nuclear properties of the International Thermonuclear Experimental Reactor (ITER) JA Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) and to ensure its design conforms to nuclear licensing regulations, nuclear analyses have been performed for the WCCB-TBM's components, including its frame, shield, flange, port extension, pipe forest, bio-shield and Ancillary Equipment Unit (AEU). Utilising Monte Carlo code MCNP5.14, activation code ACT-4 and the Fusion Evaluated Nuclear Data Library FENDL-2.1, this paper focusses on the shutdown dose rate calculation for the WCCB-TBM. Monte Carlo N-Particle Transport Code (MCNP) geometry input data for the TBM are created from computer-aided design (CAD) data using the CAD/MCNP automatic conversion code GEOMIT, and other geometry input data are created manually. The ‘Direct 1-Step Monte Carlo’ method is adopted for the decay gamma-ray dose rate calculation. Behind the bio-shield, the effective dose rates 1 day after shutdown are about 0.2 μSv h−1, which are much lower than 10 μSv h−1, the upper limit for human access. Behind the flange, the effective dose rates 106 s after shutdown are 50–80 μSv h−1, which are lower than 100 μSv h−1, the upper limit for human hands-on access for workers performing maintenance.

Reduced activation ferritic/martensitic (RAFM) steels have been developed as the structural material for the fusion demonstration reactor, DEMO. These steels contain elements that produce radioactive isotopes and decay to low levels in timeframe required by the waste management scenario. Developments within the past quarter-century suggest a practical limit to the removal of undesired impurities such as Co, Cu, Ni, Mo and Nb. The concentrations of elements essential for the mechanical properties of RAFM steels, such as Al and N, required a compromise between the waste disposal scenario and performance demand. The limits and potential of RAFM steel pertaining to reducing the activation levels after service are discussed based on the actual achievements of F82H, Japanese RAFM steel, and numerical analyses of the activity. It was found that in order to achieve the shallow land burial limits 100 years after a reactor shutdown, Ni is the most significant impurity that must be removed (Mo in the case of the first wall). Limiting N below concentrations of 100 ppm will not be possible for a large scale melt, but concentrations of Al up to the maximum amount that has been achieved present no problems.

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li2O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of 235U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′cont) reaction cross section data and resonance parameters were improved.

In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.

To perform post irradiation tests of superconducting strands, a 15.5 T superconducting magnet and a variable temperature insert (VTI) were installed at a radiation control area in Oarai center, Institute for Materials Research, Tohoku University. Both of the 15.5 T magnet and the VTI are conductively cooled with GM refrigeration. The commissioning test of the system is still ongoing because unexpected troubles occurred during the commissioning. Also, a SQUID system with the maximum field of 7 T has been installed at another radiation control area to investigate the magnetic property of uranium and its isotopes. These devices are very useful to study the electro-magnetic properties of the neutron irradiated superconducting strands. This paper will introduce the new superconducting test facility, and some data recently obtained will be presented together with the data of magnetization evaluated with the SQUID, and the discussion on the irradiation effect on superconducting properties will be performed.

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li 2 TiO 3 ) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

A new integral experiment with a deuteron–triton fusion (DT) neutron beam started in order to validate scattering cross section data. First the DT neutron beam was constructed with a collimator. The performance of the collimator system and the characteristics of the DT neutron beam were measured. Second a new integral experiment for type 316 stainless steel (SS316) was carried out with this DT neutron beam. The DT neutron beam of 3.5 cm in diameter was injected to the front surface center of an SS316 cylindrical assembly. Reaction rates of the 93Nb(n,2n)92mNb reaction in the assembly were measured with the activation foil method and were calculated with the Monte Carlo transport calculation code. The measurement points were located at three positions, on the center of the beam axis and at 15 cm and 30 cm apart from the axis. The ratio of calculation to experiment of the 93Nb(n,2n)92mNb reaction rate became smaller than 1 with the distance from the beam axis. Then, the dependency of each reaction rate on scattering angle was calculated. It was proved that at off-axis positions, where C/E is smaller than 1, 90° scattering contribute relatively larger than at on-axis positions and backward scattering made little contribution to the results in this experiment. The reasons of the discrepancy between the measured and calculated data will be investigated.

In order to measure the 2.5 MeV neutrons produced by DD nuclear fusion reactions, we have developed a compact neutron detector based on nuclear emulsion. After optimization of development conditions, we evaluated the response of the detector to an accelerator-based DD neutron source. The absolute efficiency at an energy of 2.5 MeV was estimated to be (4.1±0.2)×10 -6 tracks/neutron.

A fusion reactor will yield a lot of high energy neutrons, and some of them will stream out of a plasma vacuum vessel and penetrate a blanket system and reach superconducting magnets which provide high magnetic field to confine high energy ionized particles. Under the neutron irradiation, the magnet materials will be activated and the properties will change. In this study, a research network on study of the irradiation effects is introduced and some data recently obtained are presented. By neutron irradiation, the critical current of the Nb3Sn wire increased and the critical field did not change. The organic insulation materials were degraded at neutron fluence of 1 × 1022 n/m2. In addition, the outline of an installation plan of 15.5 T superconducting magnet into radiation control area is described and study issues are explained. The project is undergoing as a program of Nuclear Basic Infrastructure Strategic Study Initiative.

The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0, was released in May, 2010. As one of benchmark tests, we have carried out a benchmark test of JENDL-4.0 iron data, which are very important for radiation shielding in fusion reactors, by analyzing the iron fusion neutronics integral experiments (in situ and Time-of-Flight (TOF) experiments) at JAEA/FNS. It is demonstrated that the problems of the iron data in the previous version of JENDL, JENDL-3.3, are solved in JENDL-4.0; the first inelastic scattering cross section data of 57Fe and the angular distribution of the elastic scattering of 56Fe. The iron data in JENDL-4.0 are comparable to or are partly better than those in ENDF/B-VII.0 and JEFF-3.1.

A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

In the neutronics experiment for the ITER test blanket module with a 6 Li-enriched Li 2 TiO 3 layer and a beryllium layer conducted at the FNS facility of Japan Atomic Energy Agency, the calculated tritium production rate (TPR) was approximately 10% larger than the measured value only when a neutron source reflector composed of SS316 was attached. On the other hand, the influence of the reflector on the TPR prediction accuracy was not seen in the recent blanket experiment with a natural Li 2 TiO 3 layer, beryllium layers and the reflector. We investigated the former experiment in detail, and found an unphysical tendency in the measured TPR distribution. In order to clarify whether the deterioration of the TPR prediction accuracy originates from the reflector or not, we have conducted the same experiment as the previous experiment again. In the present experiment, the measured TPR distribution inside the 6 Li-enriched Li 2 TiO 3 layer agreed well with the calculated value within an estimated experimental error of 6%. We conclude that the overestimation of TPR observed in the previous experiment would be due to some experimental faults and that the TPR prediction accuracy is good even in the case with the reflector.

Nuclear data are indispensable for development of fusion reactor candidate materials. However, benchmarking of the nuclear data in MeV energy region is not yet adequate. In the present study, benchmark performance in the MeV energy region was investigated theoretically for experiments by using a 14 MeV neutron source. We carried out a systematical analysis for light to heavy materials. As a result, the benchmark performance for the neutron spectrum was confirmed to be acceptable, while for gamma-rays it was not sufficiently accurate. Consequently, a spectrum shifter has to be applied. Beryllium had the best performance as a shifter. Moreover, a preliminary examination of whether it is really acceptable that only the spectrum before the last collision is considered in the benchmark performance analysis. It was pointed out that not only the last collision but also earlier collisions should be considered equally in the benchmark performance analysis.

Some neutron KERMA factors in ACE files are negative and extremely large if nuclear data libraries do not keep energy-balance. The status of neutron KERMA factors in the official ACE file of ENDF/B-VII.1 is examined. As a result, it is found out that neutron KERMA factors of nuclei more than 200 in ENDF/B-VII.1 have some problems. Effects of the inadequate KERMA factor are also investigated, which are large for neutron heat while those are small for total (neutron + gamma) heat. Users who use only neutron KERMA factors should check if the factors are adequate or not before they use the factors.