J. Hori

Development of in-vessel neutron monitor using micro-fission chambers for ITER

A neutron monitor using micro-fission chambers to be installed inside the vacuum vessel has been designed for the International Thermonuclear Experimental Reactor (ITER). The monitoring system needs to be insensitive to the changes of the plasma position and the profile, and the locations behind upper and lower outboard blankets were selected as appropriate based on the neutron transport calculations with the Monte Carlo code for neutron and photon transport (MCNP). Employing both pulse counting and Campbelling modes in the electronics, the ITER requirement of 107 dynamic range with 1 ms temporal resolution will be accomplished. The system meets the 10% accuracy required for the fusion power monitor. A set of a 235U micro-fission chamber with 12 mg UO2 and a fissile-material-free “blank” detector to eliminate noise issues arising from γ rays, etc. were fabricated based on the design. The vacuum leak rate of the chamber with the mineral insulated (MI) cable, resistances between the central conductor and ou...

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

Measurement of radiation skyshine with D-T neutron source

Abstract The D–T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×1011 n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9×0.9 m2 was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D–T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9×10−22 Sv/(source neutron) at a distance of 30 m from the D–T target point and the dose rate was attenuated to 4×10−24 Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m.

Measurement of keV-neutron capture cross sections and capture gamma-ray spectra of 143,145,146Nd

The neutron capture cross sections and capture γ-ray spectra of 143,145,146 Nd were measured in the neutron energy region of 10 to 90keV and at 550keV. A neutron time-of-flight method was adopted with a 1.5-ns pulsed neutron source by the 7 Li(p, n) 7 Be reaction and with a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections were obtained with the error of about 5% by using the standard capture cross sections of 197 Au. The evaluated values of JENDL-3.2 and previous measurements were compared with the present results. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. An anomalous shoulder was observed around 2 MeV in the γ-ray spectra of 145,146 Nd, and the energy position of the shoulder was consistent with the systematics obtained in our Drevious work.

Neutron Tranamission Experiment of Boron-doped Resin for the JT-60SC Neutron Shield using 2.45 Mev Neutron Source

The neutron shielding design has been carried out for the JT-60 superconducting modification using the boron-doped resin. For the boron doped resin with 0.8 and 2 wt % boron, the neutron transmission experiment with 2.45 MeV neutron source has been performed. The shielding performance of the boron-doped resin for the fast neutron flux does not depend on the boron density in the range of 0.8-2 wt% boron. On the other hand, for the thermal neutron flux the shielding performance of 2 wt% boron is superior to that of 0.8 wt% boron. Most of the reaction rates by DOT3.5 and MCNP calculations agree with the experiment results within 20 %.

Neutronics Experiment of 6Li-enriched Breeding Blanket with Li2TiO3/Be/F82H Assembly Using D-T Neutrons

To improve the nuclear performance of fusion breeding blankets, the application of thermal type blankets with 6Li-enriched ceramics and low-activation ferritic steel is suggested. However, appropriate neutronics experiments to investigate the nuclear performance of such blankets have never been carried out. Therefore, we have done blanket neutronics experiments with stratified 95-% enriched 6Li2TiO3, stainless steel F82H and beryllium blocks using JAERI’s Fusion Neutronics Source (FNS) and verified the prediction accuracy of the tritium production rate and the production rate of selected isotopes in F82H in with the Monte Carlo code MCNP-4B and JENDL Fusion File, JENDL-3.2 and ENDF/B-VI data. Calculated tritium and 51Cr production rates were in good agreement with the measured values within the experimental error. However, discrepancies were found for 187W and 56Mn production rates.

Shutdown dose evaluation experiment for ITER

A shutdown dose experiment was conducted by using the FNS at JAERI under the ITER RD (2) the shutdown dose rate could be evaluated by the new direct one-step Monte Carlo method within the experimental error, and the accuracy of the method is sufficient for shielding design of the International Thermonuclear Experimental Reactor (ITER).

Measurements of Deuteron‐Induced Activation Cross Sections for IFMIF Accelerator Structural Materials

Activation cross sections for deuteron‐induced reactions on aluminum, copper, and tungsten were measured by using a stacked‐foil method. The stacked foils were irradiated with deuteron beam at the AVF cyclotron in the TIARA facility, JAERI. We obtained the activation cross sections for 27Al(d,2p)27Mg, 27Al(d,x)24Na, natCu(d,x)62,63Zn, 61,64Cu, and natW(d,x)181–184,186Re, 187W in the 22–40 MeV region. These cross sections were compared with other experimental ones and the data in the ACSELAM library calculated by the ALICE‐F code.

Tritium measurements for 6Li-enriched Li2TiO3 breeding blanket experiments with D-T neutrons

ABSTRACT The JAERI Fusion Neutronics Source (FNS) group has carried out experiments with breeding blanket mock-ups composed of layers of beryllium, ferritic steel F82H and 6Li enriched lithium titanate ceramics, Li2TiO3. Pellets of enriched Li2TiO3 with a diameter of 12 mm and a thickness of 2 mm were used as detectors inside the tritium breeding layer. After irradiation, the pellets were dissolved and the tritium activity in the sample solution was measured by liquid scintillation counting. The experimentally obtained tritium production profile in the lithium titanate layer agreed well with MCNP calculations within the estimated error range of the experimental values (10%). Tritium loss from the pellet during storage time at room temperature, a few days, was experimentally found to be negligible.

Irradiation tests of radiation resistance optical fibers for fusion diagnostic application

To promote development of radiation-resistant core optical fibers, the ITER-EDA (International Thermonuclear Experimental Reactor-Engineering Design Activity) recommended carrying out international round-robin irradiation tests of optical fibers to establish a reliable database for their applications in the ITER plasma diagnostics. Ten developed optical fibers were irradiation-tested in a Co-60 gamma cell, a Japan Materials Testing Reactor (JMTR). Also, some of them were irradiation tested in a fast neutron irradiation facility of FNS (Fast Neutron Source), especially to study temperature dependence of neutron-associated irradiation effects. Included were several Japanese fluorine doped fibers and one Japanese standard fiber (purified and undoped silica core), as well as seven Russian fibers. Some of Russian fibers were drawn by Japanese manufactures from Russian made pre-form rods to study effects of manufacturing processes to radiation resistant properties. The present paper will describe behaviors of growth of radiation-induced optical transmission loss in the wavelength range of 350-1750nm. Results indicate that role of displacement damages by fast neutrons are very important in introducing permanent optical transmission loss. Spectra of optical transmission loss in visible range will depend on irradiation temperatures and material parameters of optical fibers.