Search

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.

International audience; Design, development, manufacturing, and test activities of the RF power system (RFPS) for Linear IFMIF Prototype Accelerator (LIPAc) were completed in Europe. Installation and commissioning activities were carried out at the International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) site in Rokkasho, Japan. Challenging IFMIF requirements led to a number of innovations during design and development. Commissioning required a major effort on calibration and fine setting and led to development of new functionalities. The RFPS was validated under full power and 125-mA deuteron beam loading conditions, in the radio frequency quadrupole (RFQ), demonstrating good performance. This article is an overview about how technical challenges impacted the prototype RFPS design and its functional evolution during commissioning, cavity conditioning, and beam operation of the RFQ.

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 construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in order to demonstrate the feasibility of the low energy section of an IFMIF deuteron accelerator up to 9 MeV with a beam current of 125 mA in CW. The next milestone of the project is the so-called Phase-B beam commissioning, and one of the missions is to demonstrate the acceleration of 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% through RFQ. Most of the components and subsystems necessary for Phase-B were delivered by 2016 under the responsibility of Fusion for Energy (F4E) as in-kind contributions of several European institutes, namely CEA (France), CIEMAT (Spain), INFN (Italy), and SCK-CEN (Belgium), and QST is in charge of the installation of the delivered equipment. The installation and check-out of the RFQ subsystem and the RF power system was completed by July 2017, and the RF conditioning of the RFQ cavity started. Also, the installation and the check-out of the important sub-systems for Phase-B, namely MEBT and beam diagnostics, have been completed. Keywords: IFMIF, LIPAc, RFQ, Accelerator, Deuteron, Neutron source

The Radio Frequency Quadrupole (RFQ) linac and 1.6 MW RF power system of the Linear IFMIF Prototype Accelerator (LIPAc) facility in the International Fusion Energy Research Center (IFERC) in Rokkasho (Japan) has been installed and conditioned. During the assembly and tuning process, the RFQ cavity was protected with a temporary tent from the potential deterioration of performance caused by dust. The vacuum in the cavity was improved through the 100 °C baking process of the cavity. The high power test of the 175 MHz RF systems up to 200 kW in CW for each of the eight RF chains was performed for checking its stable output reproducibility in Japan, before connecting 9–3/16 inch coaxial transmission lines from the RF chains to the RF input couplers of the cavity. It was confirmed that the eight RF chains provided the balanced RF power to the single RFQ cavity in-phase using a feedback loop and a synchronization system. The peak power in the cavity achieved 150 kW in the pulsed mode, which corresponds approximately to the required electric field to accelerate proton beam. Such RF conditioning process is ongoing to achieve 600 kW approximately required for deuteron beam commissioning planned in 2018.

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.

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.

During the Engineering Validation and Engineering Design Activities (EVEDA) phase (2007-2014) of the International Fusion Materials Irradiation Facility (IFMIF), an advanced engineering design of the High Flux Test Module (HFTM) has been developed with the objective to facilitate the controlled irradiation of steel samples in the high flux area directly behind the IFMIF neutron source. The development process addressed included manufacturing techniques, CAD, neutronic, thermal-hydraulic and mechanical analyses complemented by a series of validation activities. Validation included manufacturing of 1:1 parts and mockups, test of prototypes in the FLEX and HELOKA-LP helium loops of KIT for verification of the thermal and mechanical properties, and irradiation of specimen filled capsule prototypes in the BR2 test reactor. The prototyping activities were backed by several R&D studies addressing focused issues like handling of liquid NaK (as filling medium) and insertion of Small Specimen Test Technique (SSTT) specimens into the irradiation capsules. This paper provides an up-todate design description of the HFTM irradiation device, and reports on the achieved performance criteria related to the requirements. Results of the validation activities are accounted for and the most important issues for further development are identified.

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.

This work presents neutronic analyses to support the design of the test facility of the International Fusion Materials Irradiation Facility (IFMIF) in the framework of the Broader Approach activities. In order to evaluate the necessary biological shielding thickness and the reliability of remote handling tools for activated components inside the Test Cell, a detailed activation analysis has been carried out to obtain a detailed residual dose map resulting from the entire High Flux Test Module and Test Cell by utilizing the McDeLicious-11 Monte Carlo code and the mesh-tally based rigorous 2-Step approach (R2Smesh) developed at KIT. The absorbed dose rate expected at the working place of remote handling tools inside the Test Cell is approximately 100 Gy/h. A wall thickness of 150 cm for heavy concrete can be used as the design basis for hot cells in the test facility.

*Recently IAEA has been conducting a coordinated research project to update the FENDL-2.1 nuclear data library to FENDL-3. In the present study, FENDL-3, Starter Library, Release 4 (FENDL-3/SLIB4) has been extensively tested through analysis of many fusion relevant benchmark problems. FENDL-3/SLIB4 gives better results for the FNG ITER shielding benchmark and the tungsten benchmark experiments than the previous FENDL-2.1. As for the shielding benchmark in the neutron energy range above 20 MeV, the TIARA transmission experiments through iron and concrete using p-Li quasi-monoenergetic neutron source have been analyzed. FENDL-3/SLIB4 provides satisfactory results for the TIARA shielding benchmark through iron. For concrete, however, some discrepancy was found. The present analysis suggests importance of 16 O data in

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.

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 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.

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.

For the International Thermonuclear Experimental Reactor (ITER), the European Union intends to introduce a test blanket module using liquid lithium lead. In the present study, a preliminary experim...

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.

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.

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.

For the first time, tritium production rates in the water cooled pebble bed blanket are experimentally examined by using DT neutrons with two partial mockups; multi-layered mockup with water and pebble bed mockup. Tritium production rates (TPRs) are calculated by numerical analyses using Monte Carlo code MCNP-4C with nuclear data libraries FENDL-2.1 and JENDL-3.3. For experimental analysis on the pebble bed mockup, precise modelling method is proposed using the hexagonal close-packed heterogeneous geometry. Prediction uncertainties of TPRs are clarified through the experimental analyses. Ratios of the calculation results to the experimental results on local TPRs are 0.89–1.10 in the multi-layered mockup, and 0.95–1.13 in the pebble bed mockup. The calculation results on integrated tritium productions agree well with the experimental ones in both mockups. It is clarified that integrated tritium productions can be accurately evaluated.

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.

In order to experimentally evaluate the tungsten armour impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed using DT neutrons at the Fusion Neutron Source facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are made of a set of layers consisting of 0–25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li2TiO3 and 100–200 mm thick beryllium with a cross-section of 660 × 660 mm in maximum. Pellets of Li2CO3 are embedded in the Li2TiO3 layers to measure the tritium production rate. By installing the 5 mm, 12.6 mm and 25.2 mm thick tungsten armours, the sum of the integrated tritium productions at the pellets are reduced by about 2.1%, 2.5% and 6.1% relative to the case without the armour, respectively. Numerical calculations have been conducted using the Monte Carlo code. In the case of the mockups with the tungsten armour, calculation results for the sum of the integrated tritium productions agree well with the experimental data within 4% and 19% in the experiments without and with a neutron reflector, respectively.

A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author’s group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation.A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, whi...

The HELIOS supercomputer system at International Fusion Energy Research Centre, Aomori, Japan has been extensively utilized in radiation safety studies for the International Fusion Materials Irradiation Facility (IFMIF). This paper is focusing on the neutronic analysis to support the layout of the high energy beam transport (HEBT) section of IFMIF. The McDeLicious-11 Monte Carlo code, which is an enhancement to MCNP5, was utilized in order to simulate the neutron generation in the IFMIF lithium target through d-Li(d,xn) reactions and the R2Smesh approach was utilized to evaluate the ambient dose distribution after shutdown. The necessary thickness of the biological shielding and additional local shielding for the HEBT section has been evaluated. The accessibility of HEBT rooms and a necessary cooling time is discussed based on the result of shutdown dose analysis.

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...