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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.

We proposed the neutron spectrum measurement system for Advanced Fusion Neutron Source (A-FNS). It is configured with fission chambers, Self-Powered Neutron Detectors, neutron activation systems using water flow, and activation monitors. In this study, we mainly investigated the multiple activation foils which are sensitive to high energy neutrons. We selected 14 main dosimetry threshold reactions based on specific A-FNS usage conditions. The reactions are as follows: 55Mn(n,2n)54Mn, 59Co(n,p)59Fe, 59Co(n,2n)58Co, 59Co(n,3n)57Co, 75As(n,2n)74As, 89Y(n,2n)88Y, 93Nb(n,2n)92mNb, 169Tm(n,2n)168Tm, 169Tm(n,3n)167Tm, 197Au(n,2n)196Au, 197Au(n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi, and 209Bi(n,5n)205Bi. The dosimetry reaction rates were calculated using Monte Carlo code MCNP-5.15 with McDeLicious-11 and nuclear data libraries FENDL-3.1c, IRDFF-1.05 and EAF-2010. As for neutron-induced cross section data of the reactions in the high energy region, we clarified that there were limited experimental data, and proposed new experiment for their evaluations. Keywords: A-FNS, Neutron spectrum measurement, Activation foil, Dosimetry cross section

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

The validation of the Linear IFMIF Prototype Accelerator (LIPAc) is being conducted in Rokkasho Fusion Institute of QST, Aomori, Japan. The beam commissioning of the novel LIPAc RFQ has been progressing significantly, and the acceleration of the world highest current deuteron beam of 125 mA to 5 MeV (1 ms, 1 Hz pulsed beam) was achieved for the 1st time in July 2019. This study is devoted to evaluating the amount of beam loss during the high current acceleration phase by measuring the neutron produced by the interaction of the beam with accelerator components so as to verify the accelerator design. The foil activation method and the 3He neutron probes were utilized. It turned out that there was no significant trace of unexpected beam loss at the high energy section of the RFQ and the Medium Energy Beam Transport (MEBT) section.The measurement using the 3He neutron probes provided an evidence that the beam loss was well controlled in the RFQ cavity as designed. This study concludes that the LIPAc RFQ as well as MEBT is working well as predicted by the design calculation, and its concept is successful.

A new benchmark experiment on lead with DT neutrons was designed and carried out with a large lead assembly covered with Li2O blocks at JAEA/FNS to validate nuclear data of lead for measurement of ...

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.

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.

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.

Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket materials is under consideration in Japan. This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches. The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan.

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.

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.

A scenario ito obtain initial tritium to reach steady state DT operation by DD commissioning phase of the fusion plant is presented. By the DD reaction, resulted T and neutron that produces T in the breeding blanket will be accumulated. ANeutronics of LiPb liquid blanket with 90% enriched lithium-6 was evaluated with neutron transport code MCNP-5.14 using FENDL-2.1 data libraries. System dynamics code STELLA-10.0.2 was used for the analysis of tritium system and the scenario of the breeding. As the result, calculated local tritium breeding ratio in the blanket is 1.20 by DT neutron and 0.84 by DD neutron. The required extension period of DD commissioning operation without initial tritium is estimated closely 90 days, depending on the coverage and inventory. These results reveal that launching DT fusion projects without initial tritium is possible with additional DD operation period to obtain extra tritium to start up DT operation.

As a benchmark test of JENDL-4.0/HE released in 2015, we have analyzed the concrete and iron shielding experiments with the quasi mono-energetic 40 and 65 MeV neutron sources at TIARA in JAEA by using MCNP5 and ACE files processed from JENDL-4.0/HE with NJOY2012. As a result, it was found out that the calculation results with JENDL-4.0/HE agreed with the measured ones in the concrete experiment well, while they underestimated the measured ones in the iron experiment with 65 MeV neutrons more for the thicker assemblies. We examined the 56 Fe data of JENDL-4.0/HE in detail and it was considered that the larger non-elastic scattering cross sections of 56 Fe caused the underestimation in the calculation with JENDL-4.0/HE for the iron experiment with 65 MeV neutrons.

Tritium production in LiPb blanket was evaluated by neutronics analysis with Monte Carlo code, MCNP using nuclear cross-section data from the FENDL-2.1 libraries. The result suggests the sufficient TBR (Tritium Breeding Ratio) can be obtained without using additional neutron multiplier such as beryllium. Based on the calculation, a plan for the integral experiment of a small LiPb system to verify it with our DD fusion neutron beam source was made. An integral experiment on LiPb with DD neutrons properly, transport and absorption of the neutron in a small test module was evaluated.