Search

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 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 correlation between microstructure and mechanical property of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using 10.5 MeV Fe ions up to the nominal displacement damage of 20 dpa with the damage rate of 1 × 10−4 dpa/s, while the irradiation temperature was 300 °C. Oxide nanoparticles showed stable size distribution and mean size under ion irradiation up to 20 dpa. The irradiation microstructure examined by TEM revealed that the mean size and number densities of irradiation-induced defect clusters increased with the displacement damage. The correlation between irradiation microstructure and radiation hardening was theoretically calculated using the dispersed barrier hardening model. The results showed a good agreement with the experimentally measured hardness data up to irradiation at 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values under irradiation at 20 dpa. Radiation hardening in 12Cr-6Al ODS ferritic steel was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by α’ phase transformation.

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.

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 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.

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.

It is still indispensable to reduce defect densities in SiC to realize its potential superiority for power device applications. We have studied carrier recombination (CR) processes in p-type 4H-SiC substrate by our optical scheme of two wavelength excited photoluminescence (TWEPL) at 12K. The Donor acceptor pair (DAP) photo-luminescence (PL) decreased distinctly with the addition of the below gap excitation (BGE) light, which revealed the presence of CR levels whose energy location corresponds to the BGE photon energy of 0.98 eV. Its dependence on the BGE power density together with the difference obtained by the BGE energy of 1.17 eV revealed a distribution of CR levels inside the forbidden gap of SiC. The method is free from electrical contact and is applicable to whole wafer mapping.

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.

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.

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.

Lead is one of the most important candidate materials for nuclear fusion reactor blankets. We have carried out an integral benchmark experiment on lead at the DT neutron source facility of JAEA, FNS. A cubic lead assembly on a side of 45.3 cm was set up and was irradiated with the DT neutron source. Reaction rates of the Al(n,α)Na, Nb(n, 2n)Nb, Zr(n, 2n)Zr, and In(n, n′)115mIn reactions and neutron spectra above 2 MeV were measured inside the assembly with activation foil and a small NE213 spectrometer, respectively. A Monte Carlo code, MCNP5, was adopted to calculate the reaction rates and neutron spectra. The latest nuclear data libraries, JENDL-3.3 ENDF/B-VII.0, JEFF-3.1 and FENDL-2.1, were used in the calculation. The calculation results with the three libraries except for JENDL-3.3 agreed with the measured reaction rates and neutron spectra. On the other hand, that with JENDL-3.3 underestimated the measured ones with the depth. We found out that the inappropriate evaluation of the (n, 2n), elastic scattering and inelastic scattering reactions in the lead isotopes of JENDL-3.3 caused the disagreements

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.

Effects of residual stress on irradiation hardening were studied in advance for predicting irradiation assisted stress corrosion cracking. The specimens of SUS316 and SUS316L with several % plastic strains, which correspond to weld residual stress, were prepared by bending and keeping deformation under irradiation. Ion irradiations of 12 MeV Ni3+ were performed at 330, 400 and 550 oC to 45 dpa. No bended specimen was simultaneously irradiated with the bended specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nanoindenter. The residual stress did not relax even for the case of the higher temperature aging at 500 oC for the same time of irradiation. The residual stress after ion irradiation up to high dpa, however, relaxed at these experimental temperatures. The irradiation hardening of stressed specimen was obviously lower than that of un-stressed one in case of SUS316L irradiated at 300 oC to 12 dpa.

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).

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.

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.

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.

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.

The amplified optical parametric fluorescence (AOPF) from optical parametric chirped-pulse amplifier (OPCPA) was suppressed down to 5/spl times/10/sup 7/ with injecting a residual fundamental pulse of the seeded Q-switch Nd:YAG pumping laser as a quenching beam.

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.

High density compression of over 600 times solid density has been achieved by the laser implosion of a hollow shell pellet of deuterated and tritiated plastics (CDT). This is the highest density of material which has ever been realized in a laboratory and corresponds to the density four times higher than the center of the sun.

We demonstrated efficient fast heating of a solid wire target attached on a hollow entrant cone target with an ultra-intense laser. The heating temperatures were 1.2-2.1 keV and the heated temperatures were independent of the wire length, implying strong collimation of high energy density electrons produced in an interaction of the ultra-intense laser light with the cone-wire target along the wire. The high energy density electrons propagated along the wire target with maintaining the high energy density of electrons.

We have developed large size diffraction gratings and flat mirrors for large size pulse compressor in LFEX laser system. The required size of grating is 1.8 m, and we use 2 segment system with 91 cm element gratings. The substrates of mirrors and gratings are made of newly developed fused synthetic silica glass for low thermal expansion. The coating on this fused silica substrates are made by the ion beam assisted deposition (IAD) process for low stress coating. The size of element grating is 91 x 42 x 9 cm, and the largest mirror is 86 x 43 x 9 cm. The grating was lured by a novel scanning beam interference lithography system (NanoRuler II) in Plymouth Grating Laboratory in USA. The diffraction efficiency of the grating is better than 95 %, and the wave-front error is better than λ/3 at 632 nm.

The laser-driven equation-of-state (EOS) experiments for polyimide are presented. The experiments were performed with emission measurements from the rear sides of shocked targets at up to a laser intensity of 10 1 4 W/cm 2 or higher with 351 nm wavelength and 2.5 ns duration. Polyimide Hugoniot data were obtained up to 0.6 TPa with good accuracy. Applying low-density foam ablator to the EOS unknown material, we also obtained the data at a highest pressure of 5.8 TPa in the nonmetal materials. Those data were in agreement with the theoretical curves.

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.

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...