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The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

The detailed shielding analysis has been performed in order to find a configuration of the radiation shielding in the accelerator building necessary for the operation of the Linear IFMIF Prototype Accelerator (LIPAc), which is under construction in Rokkasho, Japan, at the full power of 1.125 MW to satisfy the radiation dose limit prescribed by Japanese regulations. The final design of the beam dump constructed by CIEMAT, Spain was fully taken into consideration. The impact of all the features potentially causing neutron streaming were investigated and the optimization of local additional shieldings were conducted. As the result, the most important one was found to be the air ventilation duct shield, which completely covers the ducts and creates an additional labyrinth. By considering the additional shielding configuration determined in the present study, the legal criteria at the controlled area boundary can be completely satisfied.

The IFMIF accelerator facility consists of 2 identical linacs, each accelerating a 125 mA CW deuteron beam up to the energy of 40 MeV. In order to reach these unprecedented performances, the Linear IFMIF Prototype Accelerator (LIPAc) is under installation and commissioning at the International Fusion Energy Research Centre (IFERC) in Rokkasho, Japan, in the framework of the IFMIF/EVEDA project, which is part of the Broader Approach (BA) agreement between Japan and EU. The accelerator is designed to validate components up to 125 mA CW deuteron beam at 9 MeV. The accelerator components of LIPAc have been designed and manufactured mainly by European Institutes. The injector and superconducting linac, RFQ, MEBT, Diagnostics Plate, HEBT and beam dump have been developed respectively by CEA-Saclay, INFN-Legnaro and CIEMAT-Madrid and were delivered to Rokkasho between 2013 and 2016. The commissioning of the injector with beam started in November 2014. This paper dealt with the experimental data obtained with the beam diagnostics of the injector. The electrical measurements of the beam intensity on the beam stopper were compared with calorimetric measurements. The beam profiles measured with a CCD camera and a custom image-intensified CID camera are also addressed. The analysis of beam emittance and ion species fractions from data obtained with an Allison scanner is described and the results of species fraction measurements are compared with those obtained by using a deported spectrometer. Finally, the analysis of beam space potential from data obtained with a 4-Grid analyser is presented.

The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress at Rokkasho, Japan. The important milestone of the project is to commission the necessary components, namely RFQ, MEBT and diagnostics, to accelerate the proton beam up to 2.5 MeV or the deuteron beam up to 5.0 MeV in pulsed mode with a low duty cycle of 0.1%. The detailed radiation safety analysis has been conducted for the beam commissioning with the consideration of the specific radiation source conditions and the building configuration. As a result, it turned out that the dose rate expected outside the accelerator vault during deuteron beam operation at 5 MeV is mostly dominated by the neutron streaming effect through three major penetration sectors, i.e. HVAC ducts, water pipe holes and underground pits, while it is negligibly small during the proton beam operation at 2.5 MeV. It was confirmed that the dose rate expected does not exceed the legal limit for the controlled area boundary and no serious radiological issues exist in the beam commissioning planned.

The Linear IFMIF (International Fusion Materials Irradiation Facility) Prototype Accelerator (LIPAc) is a 9 MeV, 125 mA, continuous wave (CW) deuteron accelerator aimed to validate the technology for the IFMIF accelerators. The construction of LIPAc, which is currently the most powerful deuteron accelerator in the world, has been carried out under the Broader Approach (BA) Agreement between EU and Japan, and it is located at Rokkasho (Japan). CIEMAT is one of the five European Institutions that has participated in the design, manufacturing and commissioning/operation of the main accelerator components, among them, the Radio Frequency Power System (RFPS). The RFPS contains all the equipment necessary to generate the required RF power to feed the LIPAc cavities. These cavities demand eighteen RF power chains at 175 MHz being distributed as follows: eight 200 kW tetrode-based chains for the Radiofrequency Quadrupole (RFQ), two 16 kW solid-state chains for the re-buncher cavities, and eight 105 kW tetrode-based chains for the Superconducting RF Linac Half-Wave Resonators. The design of the RFPS main components is presented in this paper, including the tetrode-based chains, the Solid-State Power Amplifier (SSPA) for the re-buncher cavities, the High Voltage Power Supplies (HVPSs) for the final amplifiers anodes and the RF water cooling system. Additionally, the main difficulties encountered during the first months of the RFPS commissioning and operation will be described, together with the applied improvements.

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.

A neutronics benchmark experiment on copper assembly was performed end 2014–beginning 2015 at the 14-MeV Frascati neutron generator (FNG) of ENEA Frascati with the objective to provide the experimental database required for the validation of the copper nuclear data relevant for ITER design calculations, including the related uncertainties. The paper presents the pre- and post-analysis of the experiment performed using cross-section sensitivity and uncertainty codes, both deterministic (SUSD3D) and Monte Carlo (MCSEN5). Cumulative reaction rates and neutron flux spectra, their sensitivity to the cross sections, as well as the corresponding uncertainties were estimated for different selected detector positions up to ∼58 cm in the copper assembly. This permitted in the pre-analysis phase to optimize the geometry, the detector positions and the choice of activation reactions, and in the post-analysis phase to interpret the results of the measurements and the calculations, to conclude on the quality of the relevant nuclear cross-section data, and to estimate the uncertainties in the calculated nuclear responses and fluxes. Large uncertainties in the calculated reaction rates and neutron spectra of up to 50%, rarely observed at this level in the benchmark analysis using today's nuclear data, were predicted, particularly high for fast reactions. Observed C/E (dis)agreements with values as low as 0.5 partly confirm these predictions. Benchmark results are therefore expected to contribute to the improvement of both cross section as well as covariance data evaluations.

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.

The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

The R2Smesh (“Rigorous 2-step”) approach, developed by KIT for the determination of shut down dose rate distributions on high resolution mesh grids, can be utilized to calculate photon radiation dose fields around activated components which are moved from the irradiation site in ITER to some external location. In this work, the R2Smesh approach has been applied to dose rate analyses of various ITER in-vessel components of high complexity: A divertor cassette, four first wall modules, a generic diagnostic equatorial port plug and two test blanket modules. The considered components were first irradiated in ITER, and then moved to the transfer cask together with the associated decay gamma source distribution for the calculation of high resolution radiation dose fields in and around the transfer cask. The R2Smesh approach was shown to be a versatile tool for performing such analyses.

A neutron source for material and component tests is an essential tool for the DEMO reactor development. An accelerator-based neutron source such as IFMIF is regarded as the most promising one in J...

The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the computations for the five computational cases concerning safety and experiment analyses are presented.

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The Test Cell (TC) accommodates the Test Modules and the lithium target assembly. Due to the nuclear heat generation, all the Test Modules inside the TC will be actively cooled. Other components like supporting structures, pipelines, cables etc., will be passively cooled by natural convection. The heat will be removed from the steel liners surrounding the TC by active water cooling. This paper concerns the thermo-hydraulic simulations of the Test Cell using Ansys-CFX. The current simulation model includes the natural convection inside the TC, several forced convective water flows in the pipelines attached on the steel liners and the helium-cooled HFTM (High Flux Test Module). The simulations provide the only means for validating the design before the construction and operation.

The latest design updates of the IFMIF-EVEDA reference test cell (TC) are described with emphasis on the following key components: active cooling pipes for concrete biological shielding walls and stainless steel liner, TC gas leak tight boundary, and piping and cabling inside TC and between TC and the access cell (AC). Water cooling is adopted for concrete shielding walls and the liner. Buried pipes are selected for active cooling of the TC surrounding shielding walls; directly welded pipes on the liner are used to remove nuclear heat of the liner. Technical features and layout of the cooling pipes are preliminary defined. The TC vacuum boundary, which includes the TC liner, an independent TC cover plate, a rubber based sealing gasket, and welding seams between interface shielding plugs and TC liner, is described. Engineering design of the piping and cabling plugs as well as the arrangement of pipes and cables under the TC covering plate and the AC floor are updated. Pipes and cable tunnels inside the shielding plugs are arranged with several bends for minimizing neutron streaming from inside to outside of the TC. Pipes, cables, and the corresponding penetrations between the TC and the AC are carefully arranged for convenient access and maintenances.

The IFMIF test cell (TC) design has been further developed and optimized in the EVEDA phase, and finally the reference TC design has been proposed. In order to carry out the detailed neutronic analysis for the reference TC design, a very detailed geometrical model for Monte Carlo neutronic calculations has been prepared directly from engineering CAD data by utilizing the McCad conversion software developed at KIT. The geometrical model includes the detailed descriptions of the lithium target system proposed by Japan, all test modules based on the EVEDA phase design, and the 3-dimesional arrangement of the biological shielding. The Monte Carlo code McDeLicious, which is an enhancement to MCNP5, has been utilized in order to adequately simulate the neutron and photon productions from the 6,7Li(d,xn) reactions in the lithium target. The present analysis is focusing on the nuclear heating distribution inside the biological shielding, the nuclear property of the TC liner, and the biological dose distribution around TC during operation. Some countermeasures for reducing the He production in the liner are discussed.

As part of the ongoing Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF), an experimental device suitable for the irradiation and online tritium release measurements of solid breeder ceramics and beryllium is investigated. This experimental device is called the Tritium Release Test Module (TRTM). In the preliminary design phase, the possible thermal conditions, the tritium diffusion budgets, and the mechanical loads have been studied by analytical calculations and numerical codes. The most important results concern the tritium production and nuclear heating induced in the structures, the temperature distribution in the specimen region and the structure, and the diffusion of tritium through the safety barriers.

The McCad geometry conversion tool has been developed at KIT to enable the automatic conversion of CAD models into the semi-algebraic geometry representation as utilized in Monte Carlo particle transport simulations. McCad is entirely based on open source software, it is running under the Linux operating system and utilizes the Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). The converted geometry models can be output in the syntax of the Monte Carlo codes MCNP and TRIPOLI. Related visualization capabilities are based on the coupling of McCad with the ParaView software and allow to overlay mesh tally distributions to the CAD geometry. This enables perspective 3D representations or animations on the CAD geometry. The paper presents the current status of the McCad approach and its implementation, and discusses its capabilities, limitations as well as future development needs. The use of McCad for fusion neutronics applications is illustrated on the examples of the MCNP model generation for ITER Test Blanket Modules (TBM) and the test cell facility of the IFMIF neutron source including Monte Carlo shielding calculations using the converted models.

A re-evaluation of irradiation conditions in IFMIF utilizing the new EVEDA geometry model has been carried out. The focus is on the re-assessment of the nuclear responses in the High Flux Test Module (HFTM) as well as other Test Modules. The aim of the present analysis is to provide a reliable and realistic estimate based on the latest design and knowledge. To this end, we have utilized state-of-the-art nuclear data and an updated beam footprint as prepared on the basis of deuteron beam dynamics simulations combined with the Monte Carlo code McDeLicious, an enhancement to MCNP5 developed ad hoc for IFMIF neutronics calculations. The present analysis proves that (1) the averaged DPA rate of 26 dpa/y can be achieved in HFTM with 500 cm 3 volume, (2) the He/DPA value for HFTM center rigs is similar as the value expected for DEMO first wall, (3) the neutron flux gradient in HFTM center rigs is less than 10%/cm.

The IFMIF (International Fusion Material Irradiation Facility) test cell design has been further developed and optimized based on the existing modular test cell concept. Key features of the current test cell include actively cooled surrounding shielding walls with coverage of internal surfaces with stainless steel liner, independent two layer top shielding plugs for protecting the access cell from neutron and gamma radiation from the test cell, optimized piping and cabling plugs for accommodating pipe and cable penetrations and for minimizing neutron streaming, rearranged lithium quench tank to outside of the test cell, etc. According to preliminary neutronic calculation results, limited access to the quench tank area for maintenance after beam shut-off can be expected with the current arrangement. Maintenance of the lithium inlet and outlet pipes as well as the two beam ducts are also possible by introducing removable shielding plugs which can be removed and replaced in case of failure.

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.

This work presents neutronic analyses to support the IFMIF target and test cell (TTC) design in the framework of the Broader Approach activities. A very detailed Monte Carlo geometry model of IFMIF TTC based on the modular TTC concept was prepared directly from a CAD model by using the McCad conversion software which has been developed at KIT. The Monte Carlo code McDeLicious, which is an enhancement to MCNP5, was utilized and nuclear heating, displacement damage and gas production rates in the TTC vessel wall were calculated. The calculation result shows that there are two prominent peaks; downstream of the test modules due to the high energy neutron contribution for gas productions and upstream due to neutron back-streaming along the beam ducts. The result suggests it is very important in the neutronic analysis to consider the detailed configuration of TTC and test modules. The dose rate distribution during operation has been assessed for the rooms adjacent to TTC across thick surrounding walls. The necessary thickness for the shielding walls has been examined. The result demonstrates the substantial improvement in the shielding capability for the top access cell with the present TTC design.

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.

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.

Our previous analysis of the integral experiments (in situ and TOF experiments) on beryllium with DT neutrons at JAEA/FNS pointed out two problems by using MCNP4C and the latest nuclear data libraries; one was a strange larger neutron peak around 12 MeV appearing in the TOF experiment analysis with JEFF-3.1 and the other was an overestimation on law energy neutrons in the in situ experiment analyses with all the nuclear data libraries. We investigated reasons for these problems in detail. It was found out that the official ACE file MCJEFF3.1 of JEFF-3.1 had an inconsistency with the original JEFF-3.1, which caused the strange larger neutron peak around 12 MeV in the TOF experiment analysis. We also found out that the calculated thermal neutron peak was probably too large in the in situ experiment. On trial we examined influence of the thermal neutron scattering law data of beryllium metal in ENDF/B-VI. The result pointed out that the coherent elastic scattering cross-section data in the thermal neutron scattering law data of beryllium metal were probably too large.

In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron cross-section and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R&D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

In the previous blanket neutronics experiments conducted at the FNS facility of Japan Atomic Energy Agency, the following disagreements between experiments and analyses have been pointed out: (1) in the experiment with a 6 Li-enriched Li 2 TiO 3 layer and a beryllium layer, the calculation was found to overestimate the tritium production rate (TPR) by approximately 10% when a neutron source reflector composed of SS316 was attached. (2) In the experiment with natural Li 2 O pebbles sandwiched by beryllium layers, TPR near the rear beryllium layer was overestimated by up to 10%. In order to identify the cause of the above problems clearly, a new blanket neutronics experiment with DT neutrons using natural Li 2 TiO 3 layer and beryllium layers was conducted at FNS. TPR distributions inside the natural Li 2 TiO 3 layer in assemblies with and without the source reflector were measured with Li 2 CO 3 pellets. The measured TPR agreed well with the calculations within 6% in both assemblies. The influence of the reflector and beryllium on TPR prediction accuracy was not seen in the present experiment.

Lithium zirconate, Li 2 ZrO 3 , is known as a candidate blanket material in a fusion reactor. Various neutronics benchmark experiments for zirconium have thus been carried out so far. According to the independent benchmark studies by two parties, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. However, the reason has not yet been made clear up to now. The author group expects it would be due to a problem of evaluation for the nat Zr(n,2n) reaction cross-section, because the cross-section measurement is basically not possible with the foil activation method for zirconium isotopes except for 90 Zr. In the present study, two neutrons emitted from nat Zr(n,2n) reaction have been measured directly to investigate the reason for the above overestimation. The measurement was done with our own special technique of detecting angle-correlated neutrons by the coincidence detection technique and the pencil-beam DT neutron source of FNS, JAEA. Angle-correlated energy differential cross-sections for nat Zr(n,2n) reaction were successfully measured. The obtained total cross-section above the emitted neutron energy of 800keV was fairly larger than the one evaluated in JENDL-3.3. The total cross-section of nat Zr(n,2n) reaction was estimated by extrapolating the spectrum down to zero energy taking into account the nuclear temperature. The estimated cross-section value with the nuclear temperature of 1 MeV, which is larger than the one adopted in JENDL-3.3, was in acceptable agreement with JENDL-3.3. It is suggested from the result that the disagreement pointed out in the previous benchmark studies may be due to inappropriate nuclear temperature used in the evaluation. Further investigation of the nuclear temperature employed in the nuclear data evaluation should thus be carried out once again.

Integral benchmark experiments with DT neutrons are not always sufficient for nuclear data benchmarking in the MeV region, below 10 MeV. A neutron spectrum shifter, which will be placed between a sample and a DT neutron source, is effective to moderate DT neutrons incident to the sample. In order to estimate effects of the spectrum shifter, the ratio of the contribution of 14 MeV neutrons in the leakage neutron and gamma-ray spectra was calculated with MCNP-4C for an experimental configuration at FNS of JAEA, Japan. The calculations were carried out for a Li 2 TiO 3 sample with a Be, D 2 O, or 7 LiD spectrum shifter. It was found out that the Be shifter was superior to others and the Be shifter was effective to decrease the contribution of 14 MeV neutrons especially for secondary gamma-ray spectrum measurements.

In an irradiation experiment using a LiAl/Pb assembly, we found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10–16 MeV region by around 30% and that in the 0.5–5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5–10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: the cross-sections of the (n,2n) reaction in JENDL-3.3 for lead isotopes are too large and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for 208Pb in JENDL-3.3 reproduce previous experimental double-differential cross-section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5–10 MeV with JEFF-3.1 is still unclear.

In the author’s group, a fusion–fission (FF) hybrid energy system has been analyzed using our own burnup calculation system consisting of Monte Carlo transport code MCNP-4C and point burnup code ORIGEN2.1. Since the neutron energy spectrum changes along with progress of burnup in a subcritical system, it is necessary to update one-group cross-section library in each burnup step. The one-group cross-sections are normally updated by collapsing the evaluated nuclear data such as JENDL and ENDF using a neutron flux calculated by an appropriate transport code such as MCNP. The collapsed cross-sections are handed over to ORIGEN, and the reaction rates for burnup of elements are thereafter estimated accurately. As well known, MCNP generates track-length (TL) data in the neutron transport calculation, which are base data to estimate the neutron flux. We thus use the track-length data directly instead of the calculated neutron flux, in order to evaluate the reaction rate as accurately as possible. However, the number of TLs becomes extremely large and thus it takes a longer computation time. We therefore reduce the number of TLs used in the cross-section collapsing process as far as the accuracy is conserved. However, in some energy region the number of TLs is inversely too small to conserve the original cross-section accuracy of the evaluated nuclear data files, because the number of TL data per unit energy is smaller than that of the nuclear data. In the present study, the weight-window (WW) technique of MCNP was applied to our burnup calculation system in order to control the number of TLs in such an energy region artificially and to complete the collapsing process accurately in the whole energy region. As a result, the variance of the calculated neutron flux thus deteriorates slightly, but the number of TLs could be successfully adjusted to conserve the accuracy of the nuclear data file in the whole energy region. And the accurate reaction rate estimation for burnup with MCNP was finally realized and simultaneously the computation time could be saved reasonably.

Irradiation experiments with a LiAlPb assembly have been performed using the DT neutron generator of Technical University of Dresden. The main goal was to qualify LiF thermoluminescence detectors (TLDs) for tritium breeding experiments and to test a recently improved NE-213 spectrometer, both for application in a neutronics mock-up experiment of the European HCLL-TBM at ENEA Frascati and Technical University of Dresden (TUD) planned for 2008. The LiF-TLD were set up inside the LiAlPb assembly and irradiated with the DT neutron generator of TUD in continuous mode. A good agreement with TPR calculated with MCNP was found. Fast neutron spectra inside the assembly have been obtained with the NE-213 detector. The flux of the thermal and epithermal neutrons inside the assembly, which is important for the tritium production rate from Li 6 , was measured with a He 3 counter and the DT neutron generator operated in pulsed mode.

Double-differential cross-section (DDX) for emitted charged particles is necessary to estimate material damage, gas production and nuclear heating in a fusion reactor. Detailed measurements of the cross-sections for beryllium, carbon and fluorine, which are among the composition materials of expected fusion blankets and first walls, were carried out with a charged-particle spectrometer using a pencil-beam DT neutron source. As verification of the cross-sections evaluated in three nuclear libraries (JENDL-3.3, ENDF/B-VI and JEFF-3.1), our measured data were compared with the data evaluated in the libraries. From the comparison, the following problems were pointed out: Beryllium: Remarkable differences in energy and angular distribution for α-particles were observed between the measured data and the libraries. The estimated total cross-section for α-particle production well agreed with the libraries. Carbon: There was a discrepancy of about 20% between JENDL-3.3 and ENDF/B-VI (JEFF-3.1) for α-particle production cross-section, and no DDX for α-particles is given in the libraries. Our obtained total cross-section for α-particle production was rather consistent with ENDF/B-VI (JEFF-3.1), and the value evaluated in JENDL-3.3 seemed too large. Fluorine: The remarkable differences for DDX of protons and α-particles were observed between the obtained result and JENDL-3.3, although detailed DDX was stored only in JENDL. The obtained total cross-sections mostly supported the evaluation of ENDF/B-VI (JEFF-3.1).

The effects of ELM-like heat pulses on tungsten surface morphology were studied by using a fundamental wave (wavelength of 1.06 μm) of a YAG laser, which was operated in a multi-pulsed mode with an effective pulse length of about 100 μm, typical pulse length of ELMs. Tungsten base temperature was 640 °C, heated by a sheath heater and laser pulse heating. The energy absorption rate of the laser light was about 30% for mirror finished surface. Under non-melting conditions (laser pulse energy fluence: ≤0.1 kJ/cm2), roughening and cracking formation within the laser spot appeared. As the energy fluence decreased, the larger shot number was necessary for the surface morphology change to appear. Even at the highest temperature of around 1240 °C, surface roughening and fine cracking appeared at the large shot number (200,000 shots).

Fusion–fission hybrid reactor is a new energy system based on a fusion reactor. Recently, it is thought to be able to apply to nuclear waste incineration. For the neutronics analysis, the author's group developed their own neutronics and burnup calculation system, which consists of a general-purpose Monte Carlo code MCNP-4B and a point burnup code ORIGEN2. Since the fusion–fission hybrid reactor is operated in a sub-critical state, the cross-section library attached in ORIGEN2 cannot be utilized as well known. One-group cross-sections for burnup calculation had to, thus, be evaluated by a tally function of MCNP in our calculation system. However, all the necessary nuclides could not be dealt with appropriately by the above procedure because of several limitations of MCNP. In the present study, a new procedure has thus been developed to collapse cross-sections by directly using tracking information of neutron accumulated in MCNP calculation. From the test calculation, it was found that the new collapsing procedure successfully removed approximation in the previous evaluation method of one-group cross-sections, which was employed in the calculation system the authors developed previously. Also, ongoing feasibility studies for speedup of the system were briefly discussed, because this kind of cross-section evaluation procedure requires a long computation time at present. The present procedure would be a general method to evaluate collapsed cross-section for burnup with a transport calculation result of continuous energy Monte Carlo code.

There are very thin and long slits at boundaries between vacuum vessel and port plugs in ITER. The slits are 0.5–2 cm in width and their length amount about 200 cm in depth. Therefore, it is a key issue to evaluate neutron-shielding effects of assembly with such thin and long slits. In order to verify them, our group has started to conduct the D-T neutronics experiment with an experimental iron assembly at the Fusion Neutronics Source (FNS) in Japan Atomic Energy Agency (JAEA) under the ITER/ITA Task 73–10. Fission rates among a slit in the assembly were measured with U-238 and U-235 micro-fission chambers. From our experiment, the following facts were found: (1) the C/E value of U-238 fission rate is within 0.9–1.1 at the most measured points and (2) the C/Es of the U-235 fission rate were almost closed to the unity within 0.9–1.2. However, they tended to decrease for the depth slightly.

A fusion–fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In our group, a calculation system for analysis of fusion–fission hybrid reactor has been developed and various transport and burnup calculations were carried out for hybrid systems with U–Pu fuel cycle and ITER model so far. It was confirmed that such system is feasible and even beneficial. However, there is difficulty to hold down the production rate of minor actinides. Thus, the study of fusion–fission hybrid energy system with Th–U fuel cycle was performed. The calculations for hybrid system with U–Pu fuel cycle were carried out again because of development of the calculation system. As a result of analysis, a fusion–fission hybrid system with thorium fuel cycle meets requirements which include self-sufficiency of tritium, subcriticality of the system and acceptable cooling capability. In comparison with U–Pu cycle, the production rate of minor actinides of Th–U fuel cycle is generally lower than that of U–Pu cycle as expected. In addition, we investigated two different coolants of helium gas and water. The water-cooled blanket is suitable for energy generation. On the other hand, gas-cooled blanket has an advantage in FP transmutation.

Two nuclear microprobe techniques; deuteron-induced nuclear reaction analysis (d-NRA) and neutron elastic recoil detection analysis (NERDA), are presented to obtain hydrogen isotope depth profiles of fusion reactor materials in a shallow and deep region, respectively. Tritium and deuterium depth profiles in a bumper limiter of TFTR used in the D–T experiments were measured as an application of d-NRA. The T(d, α)n and D(d, t)H reactions were used to determine the depth profiles. The tritium concentration had a peak, whereas deuterium showed a broad distribution in the profile. Also, the application of NERDA is demonstrated. A proof-of-principle experiment was performed using a standard sample of deuterated polyethylene. The depth profile and resolution for NERDA are discussed.

Double-differential charged-particle emission cross section (DDXc) is a fundamental value to estimate nuclear heating and material damages in fusion reactor development. In the present study, measurements of DDXc for several light elements have been carried out with a newly developed spectrometer. In the spectrometer, a good signal-to-noise ratio and very low minimum measurable energy were realized by using a pencil-beam DT neutron source of FNS in JAERI and a counter-telescope system with very thin silicon surface barrier detectors of appropriate thicknesses. The detailed DDXc data in a wide emission energy range were obtained for 27Al, 9Be, natC and 19F. For 27Al(n,xα) reactions, the measured total cross section agreed well with the evaluated data. For 9Be(n,xα) reactions, there were differences in the α-particle emission DDX between measurement and the evaluated data in the detail of energy structure. For natC(n,xα) reaction, a large discrepancy between the present result and the previous experimental data appeared. For 19F(n, charged-particle) reactions, some peaks corresponding to excited states of the residual nuclei were observed in the energy spectrum of emitted proton, deuteron, and triton particles.

For the evaluation of the damages, nuclear heating, and also tritium inventory in the fusion materials, the precise information of the energetic charged particles produced in the materials by nuclear reactions is important. Fusion neutronics source in Japan Atomic Engineering research Association, JAERI/FNS, has started the measurements of the double differential cross-sections (DDXs) and the angular differential cross-sections (ADXs) for the candidate materials of fusion reactors using the collimated D–T neutron beam system and the precise results of the DDXs and ADXs for the 9 Be( n , αx ) and 9 Be( n , t ) reactions have been obtained. Also, from the comparison of the evaluated nuclear data for the beryllium, it was found that the calculated DDX of 9 Be( n , α ) above about 3 MeV was underestimated and the total 9 Be( n , t ) cross-section obtained with the measurement of the DDX and ADX was in good agreement with the evaluated nuclear data. In the following sections, the details of the measurement methods for the charged particles and the obtained DDX and ADX for the beryllium are described.

Neutron calculations have been carried out for a fusion–fission (FF) hybrid reactor system, focused on the Japanese situation, to be realized as soon as possible. The ITER facility configuration has been taken as basis for the calculations. In the blanket region reprocessed plutonium and natural uranium oxide were taken as nuclear fuel for neutron/energy multiplication and lithium oxide with beryllium for tritium breeding. The plasma conditions were derived from the JT-60 experience. The neutron analysis has been done with a three-dimensional Monte Carlo code and a point burnup code. As a result, it was found that the proposed FF concept can be operated for at least 5 years with acceptable good performance from the point of view of tritium breeding, sub-criticality, energy multiplication and used for transmutation of long-lived fission products without nuclear fuel exchange and shuffling.

For diagnostics of confined and escaping a particles in burning plasma, we have proposed to apply the measurements of the gamma rays from D(α,γ) 6 Li and 19 F(α,n) 22 Na nuclear reactions and verified the validity to the diagnostics by accelerator experiments. From our experiments, we obtained the properties of the emitted gamma rays and some prospects for the diagnostics on ITER-like plasma based on their measurements. This paper discusses the spectrum properties of the emitted gamma rays and their applications to a particles diagnostics.

Time integrated X-ray emission images at an energy range of 1.8–3.6 keV were obtained to evaluate laser irradiation nonuniformity on fusion pellets. The experimental results were analyzed with a simple illumination calculation. A positive correlation was found between the nonuniformity obtained from the measured X-ray emission intensity and that from the calculated laser irradiation intensity, although the absolute values of the nonuniformities are discrepant.

In the previous studies, the calculated TPRs were overestimated by more than 10 % compared with the measured values in the experiment with a neutron source reflector. In order to confirm that these overestimation are found on other reactions, reaction rate distributions are measured on 197 Au(n,γ) 198 Au and 93 Nb(n,2n) 92m Nb in the ITER TBM mockups with and without a reflector by the activation foil method with DT neutron irradiation experiments. Analyses are performed with MCNP-4C and FENDL-2.1. The ratios of the calculation results to the experimental ones with a reflector are slightly larger than those without a reflector on the reaction rate of 197 Au(n,γ) 198 Au.

McDeLicious is an extension to the MCNP Monte Carlo code with the ability to simulate the generation of source neutrons based on D-Li interaction processes. This paper presents an overview of the M...

The safety analysis for the design of the IFMIF accelerator facility is an ongoing task within the IFMIF project, and the radioprotection analysis is one of the main sections of this task. The high energy beam transport line is a very sensitive section since it contains intense radiation sources. This work presents the radiological analysis of its current layout, and the design of local radiation shields, in order to meet project dose limits. These limits are established to protect workers in neighboring rooms during operation and to ensure unrestricted access for maintenance operations.

This work presents neutronic analyses to support the layout of the high energy beam transport (HEBT) section of the IFMIF neutron source in the framework of the Broader Approach (BA) EVEDA activities. In the HEBT section, neutron back streaming from the lithium target can cause significant damage to accelerator components and result in their activation. In order to estimate the resulting radiation doses, detailed neutron and photon flux distributions inside the Target Interface Room (TIR) and the Radiation Isolation Room (RIR) during operation are evaluated by using the Monte Carlo code McDeLicious, which is an enhancement to MCNP5. The obtained results show that the major contribution to the TIR dose during operation will come from neutrons streaming from the target through the beam ducts and fromsecondaryphotonsproducedintheseparts.Itseems to be impossible to use any semiconductor devices inside TIR, while for mechanical devices there should be no problem. The dose after shutdown due to decay gammas waspreliminarilyestimatedforthebeamductatthemost activated place in TIR. In order to reduce the shutdown dose rate, the use of a low-Mn-content aluminium alloy is proposed.

To examine the accuracy in the neutronics calculations for the Li/V-alloy blanket system without Be neutron multiplier, a fusion neutronics experiment on a Li/V-alloy assembly has been performed wi...