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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 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.

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

Conceptual design of the Radio Isotope Production Module (RIPM) in Advanced Fusion Neutron Source (A-FNS) is conducted especially by taking horizontal transfer of irradiation samples from a lateral access cell into consideration. Neutron measurements are also planned to be performed with the RIPM by replacing the samples with activation foils. A method with combination of two kinds of foils is proposed for the measurement of distribution of neutrons produced at the lithium target in A-FNS. One is a small foil to accurately measure an absolute value of a reaction rate by using a HPGe detector and the other is a large one to measure its relative spatial distribution by using an imaging plate. The ratio of contribution from product nuclei of a dosimetry nuclear reaction to others is important because the imaging plate is usually unable to discriminate energies of γ-rays. The ratio was investigated for cooling time by using the MCNP and FISPACT codes. It was clarified that the spatial distributions of reaction rates can be monitored with good accuracy over 90% for 209Bi(n,4n)206Bi, 209Bi(n,3n)207Bi, 93Nb(n,2n)92mNb, 27Al(n,α)24Na, and 115In(n,n’)115mIn reactions.

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.

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.

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.

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.

For the development of the International Fusion Materials Irradiation Facility (IFMIF) aiming at material tests for fusion power plants, the construction of the Linear IFMIF Prototype Accelerator (LIPAc) has been conducted at Rokkasho, Japan under the Broader Approach Agreement. The commissioning of LIPAc has been progressing significantly. The world most powerful RFQ successfully accelerated proton beam of 58 mA, and an important project milestone of the acceleration of deuteron beam to 5 MeV with the beam current of 125 mA in pulse mode was successfully achieved. The result proves the validity of the present RFQ design. The adjustment and tuning of the RF power system and the injector for enabling an operation at maximum performances played an important role in achieving this result. For the next step, the preparation of the high duty cycle operation of RFQ is underway. The installation of the HEBT (High Energy Beam Transport line) and the beam dump accepting 1 MW, CW beam, has been completed and a new beam transport line is under manufacturing. Also the assembly of the Superconducting RF (SRF) linac to accelerate the beam up to 9 MeV started in a clean room in Rokkasho.

In order to analyze self-efficiency of tritium fuel cycle (TFC) and share the systems thinking of TFC among researchers and engineers in the vast area of fusion reactor technology, we develop a system dynamics (SD) TFC model using a commercial software STELLA. The SD-TFC model is illustrated as a pipe diagram which consists of tritium stocks, such as plasma, fuel clean up, isotope separation, fueling with storage and blanket, and pipes connecting among them. By using this model, we survey a possibility of D-D start-up without initial loading of tritium on two kinds of fusion plant having different plasma parameters. The D-D start-up scenario can reduce the necessity of initial loading of tritium through the production in plasma by D-D reaction and in breeding blanket by D-D neutron. The model is also used for considering operation scenario to avoid excess stock of tritium which must be produced at tritium breeding ratio over unity.