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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 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 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 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.

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 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 paper summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) on a computational ITER benchmark and a series of 14 MeV neutron benchmark experiments. The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses. In general, FENDL-3 shows an improved performance for fusion neutronics applications. © 2014 Elsevier Inc.

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.

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.

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.

We carried out detailed measurement of the double-differential cross sections of fluorine for the emissions of protons, deuterons, tritons, and α-particles with 14.2MeV incident neutrons. An improved charged-particle spectrometer with a pencil DT-neutron beam furnished at the FNS facility of the Japan Atomic Energy Agency enabled us to obtain precise data with a fine energy resolution over a wide energy range and an angular range from 15 to 150°. The present experiment is the first simultaneous measurement of the four different kinds of charged particles and provides useful data to establish a nuclear reaction model of fluorine as well as to confirm previous experimental data. Angular-differential cross sections for several discrete peaks corresponding to excited states of residual nuclei were extracted to discuss the reaction mechanism of charged-particle emission. The obtained data suggest that the charged-particle emission reaction of fluorine has a complicated mechanism in which there are contribution...

6LiF small neutron detectors with an optical fiber have been used to measure 6Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of 6Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain 6Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li2O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the 6LiF detector and the ZnS detector is effective to separate the 6Li(n,α)T react...

We measured neutron spectra in beryllium and an ITER-TBM simulating assemblies by means of a Multi-Foil Activation Method (MFAM) at the FNS facility of JAEA to confirm applicability of MFAM. An unfolding code, NEUPAC, was used to determine neutron flux spectra. Initial guessed neutron spectra were calculated with the Monte Carlo code MCNP4C and a nuclear data library FENDL-2.1. JENDL Dosimetry File 99 was adopted as response data for reaction rates. We deduced neutron flux spectra in the simulated assemblies with the MFAM. From the experiments, it was found out that MFAM gave appropriate results, which indicated that MFAM was a prospect to the application of neutron spectrum measurement in ITER-TBM. However, it was also pointed out that some improvements for measurement of slow neutron below 10 eV were necessary.

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.

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.

A technique was developed to measure the angle-correlated neutron energy spectrum with the coincidence detection technique using a pencil-beam DT neutron source at the Fusion Neutronics Source (FNS) of the Japan Atomic Energy Agency (JAEA). Triple differential cross-sections of the 55 Mn(n, 2n) reaction were measured first, from which the total 55 Mn(n, 2n) reaction cross-section was deduced and compared with JENDL-3.3. The agreement was sufficient to prove the validity of the technique. With the present measuring technique, a complete measurement of a medium-heavy or heavy element such as manganese would become possible through several measurements, because the angle dependence of the emitted neutrons from it is normally weak. The technique will be a useful tool for measuring the angle-correlated neutron spectrum of light elements such as beryllium for investigation of their complicated nuclear reaction mechanism.

We carried out a detailed measurement of the neutron-induced α-particle emission double-differential cross section of carbon at 14.2MeV, for which there are few measured data in spite of its importance in many applications. In our measurement, a superior S/N ratio, high angular/energy resolutions and a wide detection energy range were realized with a pencil DT neutron beam and a countertelescope system. The obtained cross section for the 12C(n,α0)9Be(ground state) reaction agreed well with the results of previous experiments and evaluated nuclear data. The obtained angular-differential cross section of the 12C(n,n′+3α) reaction for α-particles showed a strong forward-peaked distribution that suggested a significant contribution of the direct reaction process to the 3α breakup. We attempted to calculate the emitted particle spectra by a Monte Carlo method and estimate the branching ratio of the channels that contribute to the 12C(n,n′+3α) reaction. As a result, it was found that the 12C(n,α)9Be* channels p...

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).

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.

The neutron elastic recoil detection analysis (NERDA) using a 14.2 MeV neutron beam has been developed to measure depth profiles of hydrogen isotopes up to several hundred micrometers under low background conditions. Results of a proof-of-principle experiment using polyethylene and deuterated polyethylene samples were presented. The detected energy spread due to the geometrical configuration was corrected to estimate depth profiles by an unfolding technique based on the iterative Bayes’ estimation. Hydrogen isotope depth profiles in the inner baffle plate of JT-60U were also measured as an application of NERDA. The hydrogen profile from the surface to 800 μm had three components, while a few counts of deuterium were detected. The detection limit of about 60 ppm was deduced for deuterium.

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.

Measurement of double-differential cross-sections of charged-particle emission induced by 14 MeV neutrons is quite important for estimating the nuclear heating and material damage of a fusion reactor. However, there are few experimental data available in spite of the importance of the nuclear reactions of light nuclei in fusion reactor development. In order to accurately measure differential cross-sections, especially of light nuclei, we propose a refined spectrometry system and a data processing method. The employment of a pencil-beam neutron source overcomes the fundamental difficulty of the measurement. As a result, a system with a counter telescope consisting of silicon surface barrier detectors satisfies high energy resolution and angular resolution, a wide detection energy range and adequate particle discrimination. To correct the energy loss of the emitted charged particles in the sample, a rigorous correction method by means of the unfolding procedure is applied. The validity and the capability of this spectrometry technique are confirmed from three basic measurements: the measurements of charged particles emitted from Teflon, α -particles emitted from the 27 Al( n , xα ) reaction and the protons recoiling from the 1 H( n , n ) reaction. This technique can be applied to the detailed DDXc measurements of various materials, especially of lighter materials, for fusion reactor modeling.

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.

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.

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 angle-correlated energy differential cross-section for 9 Be(n,2n) reaction has been measured with the coincidence detection technique and a pencil-beam DT neutron source of FNS, JAEA. Neutron energy spectra of two emitted neutrons were separately obtained as azimuthal- and axial-angle differential value, which is called TDX. From the measured TDX, double-angle differential cross sections, ADDX, angular differential cross section, ADX, and total cross section, TOX, were deduced. In the TDX, a characteristic energy spectrum for beryllium was seen. The azimuthal ADDX unexpectedly showed a 180deg. symmetrical spectrum, not a flat one. From the result of TOX, JENDL-3.3 and ENDF/B-VI showed a little underestimation by around 10%.

Detailed measurements of α-particle emission double differential cross section for beryllium and carbon with 14-MeV incident neutrons were carried out. In order to identify contributing reaction channels to the 9 Be(n,2n + 2α )a nd 12 C(n,n � + 3α) reactions, we attempted to reproduce energy distributions of emitted particles by Monte Carlo calculations according to reaction kinematics of a lot of channels which contribute to the reactions. As for the 9 Be(n,2n + 2α) reaction, a contribution of the 9 Be(n,α) 6 He* channels and the simultaneous breakup channel was suggested to account for the energy and angular distribution of emitted α-particles and neutrons. As for the 12 C(n,n'+3α) reaction, the 12 C(n,α) 9 Be* channels were largely contributed and well reproduced the experimental data.

Based on the successful result of fast heating of a shell target with a cone for heating beam injection at Osaka University in 2002 using the PW laser (Kodama et al 2002 Nature 418 933), the FIREX-1 project was started in 2004. Its goal is to demonstrate fuel heating up to 5 keV using an upgraded heating laser beam. For this purpose, the LFEX laser, which can deliver an energy up to10 kJ in a 0.5–20 ps pulse at its full spec, has been constructed in addition to the Gekko-XII laser system at the Institute of Laser Engineering, Osaka University. It has been activated and became operational since 2009. Following the previous experiment with the PW laser, upgraded integrated experiments of fast ignition have been started using the LFEX laser with an energy up to 1 kJ in 2009 and 2 kJ in 2010 in a 1–5 ps 1.053 µm pulse. Experimental results including implosion of the shell target by Gekko-XII, heating of the imploded fuel core by LFEX laser injection, and increase of the neutron yield due to fast heating compared with no heating have been achieved. Results in the 2009 experiment indicated that the heating efficiency was 3–5%, much lower than the 20–30% expected from the previous 2002 data. It was attributed to the very hot electrons generated in a long scale length plasma in the cone preformed with a prepulse in the LFEX beam. The prepulse level was significantly reduced in the 2010 experiment to improve the heating efficiency. Also we have improved the plasma diagnostics significantly which enabled us to observe the plasma even in the hard x-ray harsh environment. In the 2010 experiment, we have observed neutron enhancement up to 3.5 × 107 with total heating energy of 300 J on the target, which is higher than the yield obtained in the 2009 experiment and the previous data in 2002. We found the estimated heating efficiency to be at a level of 10–20%. 5 keV heating is expected at the full output of the LFEX laser by controlling the heating efficiency.

A neutron spectrometer system based on TOF (time of flight) measurement has been studied for determination of fuel density ratio in ITER. The effective collimator diameter was examined against both the saturation of the first detector and measurement efficiency as a spectrometer to achieve the sufficient statistics with required time resolution. Relation between collimator diameter and lower limit measurement efficiency was obtained. Measurement efficiency for every design of TOF spectrometer was calculated with MCNP. Those results indicate that two or three collimators of different size are needed to attain statistics errors within 10 %. An experiment to test the TOF spectrometer and to evaluate the system efficiency was conducted using accelerator neutron source. The result shows that both DD and DT neutrons are measured separately and simultaneously with a fast discrimination electronics system. The fast discrimination improves the counting capability by one order of magnitude.

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...

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...