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Abstract: An invessel neutron monitor using micro-fission chambers is one of the most important diagnostics to measure the fusion power in ITER. A single micro-fission chamber with 10mg of 235U to be installed in a gap between a shielding blanket module and a vacuum vessel can cover the range of the fusion power from 100kW to 1GW. For the low fusion power operation including D–D operation, a micro-fission chamber bundle consisting of 13 micro-fission chamber units has been designed. Nuclear heating of the micro-fission chamber for high power operations and the micro-fission chamber bundle for low power operation were analyzed by neutron transport calculations using the MCNPX code. Heat transport analyzes for the micro-fission chambers were carried out with a general purpose Finite Element Method code ANSYS 8.1, which indicated that the detector temperature might be kept less than the operational temperature limit by the thermal conduction to the vacuum vessel with the heat transfer coefficient of 100W/m2 K or larger. [Copyright &y& Elsevier]

For ion temperature measurement in the ITER high power operation phase, we propose a promising high-energy-resolution neutron spectrometer based on the associated particle detection using a proton recoil telescope and a time-of-flight spectrometer. To verify the operational principle and the basic performance of this system, we have set up a prototype system through Monte Carlo simulations and carried out a preliminary experiment with a deuterium-tritium (DT) neutron beam at the Fusion Neutronics Source, JAEA. The results have demonstrated that the energy resolution could be achieved around 3.3% (in full width at half maximum) for DT neutrons. In addition, the design consideration of this system for an application to ITER is given on the detection efficiency, the applicable range in the ITER operation, and the detector lifetime. [ABSTRACT FROM AUTHOR]

The possibility of measurement of fuel ratio from the DT/DD reaction ratio with a neutron spectrometer in the International Thermonuclear Experimental Reactor was studied. The results of neutron transport calculations using the Monte Carlo N-particle [MCNP, a general Monte Carlo N-particle transport code, version 4C, LA-13709, edited by J. F. Briesmeister, Los Alamos National Laboratory, 2000] code system indicated the possibility of DD spectrum separation from the scattered/energy-degraded neutrons derived from DT neutrons by selection of the measurement location and collimator design. A time-of-flight spectrometer was used in this study. An experiment using DD and DT neutrons from an accelerator was conducted at the Japan Atomic Energy Agency Fusion Neutronics Source facility. Using a sophisticated circuit with three discriminators for the first detector, separation of DD and DT neutron components of the spectrum in the megahertz region was confirmed experimentally. [ABSTRACT FROM AUTHOR]

For fusion neutron profile monitoring, we propose to use a new position sensitive fast neutron detector (PSFND). We have considered an applicability of the present PSFND to two types of fusion neutron camera; a pinhole camera and a double crystal time-of-flight camera. [ABSTRACT FROM AUTHOR]

A natural diamond detector (NDD) has been installed on the JT-60U tokamak to measure the flux and the energy distribution of charge exchange (CX) fast neutral particles. A NDD has many important advantages to be used as a CX neutral particle analyzer, for example very compact size, high energy resolution, and high radiation hardness etc., while the neutrons and γ rays are a large noise source in the deuterium plasma. The shield was set up around the NDD to reduce those noises. Time-resolved energy distribution of CX neutral particles corresponding to injected beam energy have been successfully obtained under high intensity neutron yield Yn>1015 n/s. Further enhanced neutral particle fluxes during sawtooth oscillation and Alfvén eigenmodes were observed with the NDD. The performance of the NDD as CX neutral particle spectrometer under high intensity neutron yield was demonstrated for the first time on JT-60U in this work. [ABSTRACT FROM AUTHOR]

Abstract: 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 55Mn(n,2n) reaction were measured first, from which the total 55Mn(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. [Copyright &y& Elsevier]

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

Abstract: 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). [Copyright &y& Elsevier]

Abstract: The possibility of tritium self-sufficiency in a low aspect ratio (low-A) tokamak reactor with the outboard blanket only has been studied to simplify the inboard structure located in a high magnetic field. The tritium breeding ratio (TBR) of the outboard blanket increases by applying a neutron multiplier such as lead and beryllium to the inboard reflector. The effect in local TBR of a 40cm thick inboard reflector of lead is about 40% that of the inboard breeding blanket. The local TBR of a reactor with both the lead inboard reflector and outboard blanket using a Li2O breeder and a beryllium multiplier can be larger than 1.35 for aspect ratios less than 2.9. The local TBR of 1.35 satisfies the net TBR requirement of 1.05 where there is 78% coverage of the breeding zone over the total plasma-facing surface area. The result indicates there is a design solution achieving tritium self-sufficiency even in a low-A tokamak reactor with the outboard breeding blanket only, using an appropriate inboard reflector. Be12Ti and Li2TiO3 are recommended from the viewpoint of safety. In the combination, the aspect ratio is required to be less than 1.35 to achieve local TBR of 1.35. [Copyright &y& Elsevier]

Abstract: Sequential charged particle reactions, that is, reactions induced by secondary charged particles generated due to neutron reactions, may cause non-negligible radioactivity in low activation material. A computer code ACT4, which had been developed to calculate radioactivity, decay heat and gamma ray source for fusion reactors was supplemented an ability to deal with the reactions. Experiments conducted with Fusion Neutronics Source in Japan Atomic Energy Research Institute for the radioactivity by sequential charged particle reactions were analyzed with the newly revised code of ACT4. As the result, the activations via sequential reaction in LiF showed reasonable agreement with experimental data, which means that the ACT4 program will work well. However, the other activations in vanadium and low activation ferritic steel did not. The disagreement must be caused by incorrectness in neutron cross sections, charged particle cross sections, charged particle emission spectra, etc., and they need to be fully reviewed for the detailed and reliable design analysis of fusion reactors. [Copyright &y& Elsevier]

Abstract: The neutronics study in support of the DEMO blanket concept has been performed using the DT neutrons at the FNS facility in JAERI on blanket mockups and a bulk beryllium assembly. From the results obtained in the blanket mockups, it has been clarified that the C/Es of the integrated tritium production are 1.01–1.04 and 1.11–1.13 in the experiments without and with the neutron reflector, respectively. The tritium production can be very accurately predicted in the experiment without the reflector. In the mockups with the 12.6 and 25.2mm thick tungsten armors, the integrated tritium productions are reduced by 3 and 6% relative to the case without the armor, respectively. Based on analysis of the integral experiment on the beryllium assembly, it is possible to conclude that the calculation uncertainty of the reaction rates in beryllium provides an acceptable agreement with the experimental data related to tritium breeding. Tritium techniques used for TPR measurements have been tested in the international benchmark experiment. There is a very good agreement within the estimated uncertainties between results obtained at JAERI and at TUD, though the results obtained at ENEA show an underestimation (about 15–20%) in comparison with FNS and TUD results. [Copyright &y& Elsevier]

Abstract: The process of neutron leaking from a 14MeV neutron source facility was analyzed by calculations and experiments. The experiments were performed at the Fusion Neutron Source (FNS) facility of the Japan Atomic Energy Institute, Tokai-mura, Japan, which has a port on the roof for skyshine experiments, and a 3He counter surrounded with a polyethylene moderator of different thicknesses was used to estimate the energy spectra and dose distributions. The 3He counter with a 3-cm-thick moderator was also used for dose measurements, and the doses evaluated by the counter counts and the calculated count-to-dose conversion factor agreed with the calculations to within ∼30%. The dose distribution was found to fit a simple analytical expression, and the parameters and are discussed. [Copyright &y& Elsevier]

Abstract: This paper presents an evaluation of accessibility of the Li loop piping considering activated corrosion product. International Fusion Materials Irradiation Facility (IFMIF) is a deuteron-lithium (Li) stripping reaction neutron source for fusion materials testing. Target assembly and back wall are designed as fully remote maintenance component. Accessibility around the Li loop piping will depend on activation level of the deposition materials due to the back wall erosion/corrosion process under liquid Li flow. Activation level of the corrosion products coming from the AISI 316LN back wall is calculated by the ACT-4 of the THIDA-2 code system. The total activities after 1 day, 1 week, 1 month and 1 year cooling are 3.1×1014, 2.8×1014, 2.3×1014 and 7.5×1013 Bq/kg, respectively. Radiation dose rate around the Li loop pipe is calculated by QAD-CGGP2R code. Activated area of the back wall is 100cm2. Corrosion rate is assumed 1μm/year. When 10% of the corrosion material is supposed to be deposited on the inner surface of the pipe, the dose rate is calculated to be less than a permissible level of 10μSv/h for hands-on maintenance, therefore, the maintenance work is assessed to be possible. [Copyright &y& Elsevier]

Abstract: Radiation detector was made of a high-quality CVD polycrystalline diamond composed of frost column like structure diamond grains, and induced charge distribution spectra and drift velocities were measured by using alpha particles. As a result, the CVD polycrystalline achieved maximum induced charge of 83% of HP/HT type IIa diamond. Moreover, the CVD crystal had lower charge loss on electrons compared with the HP/HT type IIa diamond. Drift velocities of electrons and holes were v e =7.7×104 and v h =7.3×104cm/s at an electric field of 20 kV/cm, respectively. In addition, response function measurement for 14 MeV neutrons was carried out. [Copyright &y& Elsevier]

Abstract: An integral breeding blanket experiment has been done at Fusion Neutronics Source of JAERI with an assembly of three layers of lithium titanate, beryllium, and low activation ferritic steel F82H. The experimental assembly and the neutron source were enclosed in a SS316 stainless steel can and irradiated with D–T neutrons. The tritium production rate was measured with small Li2CO3 pellet detectors. The tritium production from7Li was estimated by means of the reactions32S(n, p)32P and35Cl(n, )32P which have effective cross-sections similar to7Li(n, n)T but are more sensitive methods. The neutron flux in the assembly was checked with activation foils. The measurement results were analyzed with the three-dimensional Monte Carlo code MCNP-4C and three different beryllium cross-section evaluations taken from FENDL/MC-2.0, FENDL/E-2.0, and EFF-3.03. The library FENDL/MC-2.0 was also used for the other materials in the assembly. All three beryllium libraries model the fast flux in the assembly well. Differences were seen for indium foils. An overestimation was observed for the tritium production. [Copyright &y& Elsevier]