Fusion Neutronics Source Facility Research Articles

Analysis of postoperation radiation hazards in inertial electrostatic confinement fusion neutron source facility at center of plasma physics under institute for plasma research

Center of Plasma Physics-Institute for Plasma Research, Nazirakhat, Assam, has the neutron source based on Inertial Electrostatic Confinement of Fusion. In order to scale up the source, it is planned to build a neutron source facility. The facility needs the construction approval from AERB India for occupational radiation safety during the operation and after the operation. It will also assess the short-term and long-term radiological hazards due to the neutron source. In order to evaluate the post irradiation radiation hazards due the neutron source, the radioactivation analysis of laboratory is carried out and reported here. The radioactivity in the laboratory building and source components after short-term and long-term operation has been assessed. The tritium activity in the soil is also evaluated. The calculation is also done for the contact dose rate estimations after operation to assess the maintenance applicability inside the laboratory. The inhalation dose inside the laboratory after the operation has been also calculated to avoid any hazards after operation maintenance activities. The article provides the complete details of post operation hazards analysis for the Inertial Electrostatic Confinement Fusion neutron source facility.

Integral test of International Reactor Dosimetry and Fusion File on graphite assembly with DT neutron at JAEA/FNS

We perform an integral test of a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), which has been released from the International Atomic Energy Agency recently. A pseudo-cylindrical graphite assembly is used with a DT neutron source at the fusion neutronics source facility of Japan Atomic Energy Agency. A lot of foils for measurements of dosimetry reaction rates are placed in the small space along the central axis in the assembly. After the DT neutron irradiation, forty dosimetry reaction rates are measured and are compared with calculated values, which are obtained by using a Monte Carlo neutron transport code MCNP5-1.40 and a nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.03 as the response functions for the dosimetry reactions. The calculated results generally show good agreements with the experimental ones. The cross section data of the 68Zn(n,np)67Cu and 68Zn(n,d)67Cu reactions and the 113In(n,γ)114mIn reaction, which are not included in IRDFF-v1.03, are needed for the analysis of the 67Zn(n,p)67Cu reaction and the 115In(n,2n)114mIn reaction as the competitive reactions, respectively.

Isotope production for medical usage using fast neutron reactions

Abstract 99Mo was produced by the 100Mo(n,2n)99Mo reaction using 14-MeV neutrons from the D(3H,n)4He reaction at Fusion Neutronics Source Facility (FNS) at JAEA. A target of titanium oxide gel with molybdenum oxide was irradiated and used as the 99mTc/99Mo generator. The growth curve of 99mTc in the titanium gel target and the elution curve of 99mTc from the 99Mo/99mTc generator were obtained.

Analysis of neutron propagation from the skyshine port of a fusion neutron source facility

The process of neutron leaking from a 14 MeV 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, D ( r ) = Q D exp ( - r / λ D ) r and the parameters Q D and λ D are discussed.

Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

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.

Radioactivity production around the surface of a cooling water pipe in a D-T fusion reactor by sequential charged particle reactions

Around the surface of a cooling pipe in a D-T fusion reactor, it is expected that the radioactivity production via what is known as ‘Sequential Charged Particle Reaction (SCPR)’ would be enhanced by recoiled proton from hydrogen in cooling water. In order to simulate the circumstances, several sheets of foil with a thickness of 50–250 μm were laminated on a polyethylene board for six fusion materials (Fe, Cu, V, Ti, W, Pb). The laminated samples were irradiated with intense D-T neutrons at the fusion neutronics source facility in JAERI. After irradiation, the decay gamma rays emitted from the sequential reaction products ( 56Co, 65Zn, 51Cr, 48V, 184Re, 206Bi) were measured and the effective cross-sections for producing those were obtained at several positions. The present results indicated that the sequential reaction rate increases prominently as the location becomes closer to hydrogen compounds.

Microstructure evolution in D–T neutron irradiated silver

Irradiation of high purity silver with 14 MeV D–T neutrons was carried out at the fusion neutron source facility in Japan Atomic Energy Research Institute. The range of neutron fluence was 6.1×10 17 to 1.1×10 21 n/m 2, which is lower than in earliest D–T neutron irradiations. Thin foil and bulk specimens were irradiated at 288, 423 and 573 K, and observed using transmission electron microscopy. For irradiation at 288 K, the fraction of interstitial clusters in bulk is higher than that in thin foil. In irradiation experiments at 288 and 432 K, the number density of defect clusters is proportional to the neutron fluence to the power of 1.3. In irradiation experiments at 573 K, the defects are mostly stacking fault tetrahedrons (SFTs) and their number density is proportional to the neutron fluence. During isochronal annealing of specimens irradiated at 288 K, disappearance and coalescence of defect clusters were observed, and SFTs were mobile.

Neutronics experiments for ITER at JAERI/FNS

A series of fusion neutronics experiments has been performed at the Fusion Neutronics Source facility at Japan Atomic Energy Research Institute as ITER EDA R&D tasks to find ways of dealing with various nuclear problems originating from 14 MeV neutrons in ITER. Recently three groups of experiments were carried out: (1) straight duct streaming experiments, (2) decay heat experiments, and (3) development of a fusion power monitor utilizing activation of water. The straight duct streaming experiments suggest that the calculation accuracy for straight duct streaming analyses in ITER nuclear designs is ±40%. The decay heat experiments show that the accuracy of the decay heat calculation is within 10% for copper and type 316 stainless steel, while it is ∼30% for tungsten. It is demonstrated that a fusion power monitor utilizing activation of water is applicable to ITER.

Investigation of x-ray-energy responses of semiconductor detectors under deuterium–tritium fusion-produced neutron irradiation

For the purpose of investigating fusion-produced neutron effects on semiconductor x-ray detectors, detection characteristics of x-ray tomography detectors used before and after deuterium–tritium (DT) and/or DD fusion-plasma experiments in the Joint European Torus (JET) tokamak are studied using synchrotron radiation from a 2.5 GeV positron storage ring at the Photon Factory. Degradations in the responses after neutron exposure into the detectors are found to have functional dependence on x-ray energy. Changes in the depletion thicknesses of the detectors are investigated by means of impedance analyses. The Fusion Neutronics Source (FNS) facility of the Japan Atomic Energy Research Institute is also employed for well calibrated DT fusion-produced neutron irradiation onto these semiconductor x-ray detectors. Recovery of the response degradation is found due to a method for supplying the operational bias to the degraded detector. Our theory is applied to interpret these detector characteristics under the irradiation of DT fusion-produced neutrons in the JET tokamak and the FNS facility.

Point Defects Observed in D-T neutron irradiated Copper, Silver and Gold at 288 K with a Rotating Target in FNS_JAERI

ABSTRACTA 14 MeV D-T(fusion) neutron irradiation was carried out at fusion neutron source facility (FNS) in Japan Atomic Energy Research Institute (JAERI). Specimen temperature was controlled to 288 K. Fluence was 6.1 × 1017to 1.1 × 1021 n/m2. Both TEM thin foil and bulk specimens were irradiated at the same position. At 1018 n/m2, defects observed were single isolated dot defects. With increasing fluence, dot defects changed to complicate structure and made groupings. In a dot group, interstitial clusters and vacancy clusters were observed together. The present result was explained by the modeling that point defects in a nascent damage cascade move in crystal at 288 K and form their defect groupings.

An Alternative, ‘Non-Electrical’ Pathway for Fusion Energy Development

The physics and technology that is being developed for and that will be demonstrated in ITER [1] will be sufficient to make a very good neutron source, there are a number of potential ‘national missions’ for a good neutron source, and the further technology advances beyond ITER that would be required for a neutron source facility are essentially the same as the advances that would be required for an electrical energy producing fusion demonstration reactor. Some preliminary considerations are presented for an alternative pathway for fusion energy development, proceeding from the present through an international test reactor (ITER) stage to a fusion neutron source facility (or non-electrical applications) stage and finally to the deployment of fusion electrical power reactors. Recent studies of two types of fusion neutron source facilities for ‘national missions’ are reviewed as representative examples.

Nuclear heating measurements for SS-316, copper, graphite, tungsten, chromium, beryllium in a copper centered assembly bombarded with 14MeV neutrons and analysis

Under ITER/EDA R&D Task T-218, an experiment on nuclear heating was conducted at the Fusion Neutronics Source Facility (FNS) in JAERI in order to provide experimental data of nuclear heating for structural materials and to validate the data and methods relevant to ITER design calculations. Probe materials, SS-316, Cu, W, Zr, Be, Cr and graphite were tested in a copper centered experimental assembly bombarded with 14 MeV neutrons. A calorimetric method and TLDs were employed for total heating and γ-heating measurements, respectively. The measurements were carried out along a central axis of a copper centered experimental assembly. The total heating rate in the probe material was derived from a temperature rise measured by a thermistor attached to the probe. MCNP4A with JENDL-3.2, JENDL-Fusion File and FENDL-1 libraries were used for experimental analyses. Nuclear heating was derived with the use of KERMA factors based on JENDL-3.2 and FENDL-1. The adequacy of the KERMA data is discussed based on nuclear heating ratios of calculation to experiment (C/E).

Measurements of Tritium and 14C Production Cross Sections for 14.7-MeV Neutrons on 17O and 18O

Samples of water isotopically enriched in 17O, 18O, and 2H along with natural water were bombarded by neutrons from the intense deuterium-tritium source provided by the Fusion Neutron Source facility. After irradiation, the accumulated concentrations of 3H and 14C activities were determined by the liquid scintillation method. Special attention was paid to 14C losses in the gas phase during irradiation and preparation of scintillation counting samples. Cross sections for the 17O(n,α)14C, 18O(n,n’α)14C, 17O(n,t)15N, and 18O(n,t)16N reactions at 14.7 MeV have been measured for the first time. The following values have been obtained for these reactions: 18.0 ± 3.3, 35.4 ± 6.5, 0.82 ± 0.15, and 26.8 ± 4.9 mb, respectively, relative to the 93Nb(n,2n)92mNb standard reaction cross section of 460 mb. A study of the systematics of (n,t) reactions at 14.7 MeV on light nuclei (atomic number Z < 10) has been carried out. The experimental cross-section values are also compared with data in the comprehensive activation libraries.

A Summary of Benchmark Experiments for Simulation of Fusion Reactors Using an Annular Blanket with a Line Deuterium-Tritium Source

The Japan Atomic Energy Research lnstitute (JAERI)/U.S. Department of Energy collaborative program was performed using the Fusion Neutronics Source facility at JAERI. In Phase III of this program, tritium breeding measurements were conducted in prototypical blankets driven by a simulated deuterium-tritium neutron line source. This phase differed from the earlier two phases in respect to the spatial distribution of the source as the earlier experiments were done with a point neutron source. This series basically consisted of an annular test blanket and a pseudoline source to investigate the effect of source spread on the neutronic performance. A concise description is on the outlines of the simulated line source, the test blanket systems for Phases-IIIA, -IIIB, and -IIIC, measured items, experimental results, and their analyses. 23 refs., 8 figs., 3 tabs.

Japan Atomic Energy Research Institute/United States Integral Neutronics Experiments and Analyses for Tritium Breeding, Nuclear Heating, and Induced Radioactivity

A large number of integral experiments for fusion blanket neutronics were performed using deuterium-tritium (D-T) neutrons at the Fusion Neutronics Source facility as part of a 10-yr collaborative program between the Japan Atomic Energy Research Institute and the United States. A series of experiments was conducted using blanket assemblies that contained Li 2 O, beryllium, steel, and water-coolant channels with a point neutron source in a closed geometry that simulated well the neutron spectra in fusion systems. Another series of experiments was conducted using a novel approach in which the point source simulated a pseudo-line source inside a movable annular blanket test assembly, thus providing a better simulation of the angular flux distribution of the 14-MeV neutrons incident on the first wall of a tokamak system. A number of measurement techniques were developed for tritium production, induced radioactivity, and nuclear heating. Transport calculations were performed using three-dimensional Monte Carlo and two-dimensional discrete ordinates codes and the latest nuclear data libraries in Japan and the United States. Significant differences among measurement techniques and calculation methods were found. To assure a 90% confidence level for tritium breeding calculations not to exceed measurements, designers should use a safety factor >1.1 to 1.2, depending on the calculation method. Such a safety factor may not be affordable with most candidate blanket designs. Therefore, demonstration of tritium self-sufficiency is recommended as a high priority for testing in near-term fusion facilities such as the International Thermonuclear Experimental Reactor (ITER). The radioactivity measurements were performed for >20 materials with the focus on gamma emitters with half-lives <5 yr. The ratio of the calculated-to-experimental (C/E) values ranged between 0.5 and 1.5, but it deviated greatly from unity for some materials with some cases exceeding 5 and others falling below 0.1. Most discrepancies were attributed directly to deficiencies in the activation libraries, particularly errors in cross sections for certain reactions. A microcalorimetric technique was vastly improved, and it allowed measurements of the total nuclear heating with a temperature rise as low as 1 μK/s. The C/E ratio for nuclear heating deviated from 1 by as much as 70% for some materials but by only a few percent for others.

Direct Nuclear Heating Measurements and Analyses for Plasma-Facing Materials

Experimental measurement of nuclear heating rates was carried out in a simulated D-T fusion neutron environment from 1989 through 1992 under the U.S. DOE/JAERI collaborative program at the Fusion Neutronics Source Facility. Small probes of materials were irradiated in close vicinity of a rotating target. A sophisticated microcalorimetric technique was developed for on-line measurements of local nuclear heating in a mixed neutron plus photon field. Measurements with probes of graphite, titanium, copper, zirconium, niobium, molybdenum, tin, tungsten, and lead are presented. These measurements have been analyzed using the three-dimensional Monte Carlo code MCNP and various heating number/kerma factor libraries. The ratio of calculated to experimental (C/E) heating rates shows a large deviation from 1 for all the materials except tungsten. For example, C/E`s for graphite range from 1.14 ({delta} = 10%) to 1.36 (10%) for various kerma factor libraries. Uncertainty estimates on total nuclear heating using a sensitivity approach are presented. Interestingly, C/E data for all libraries and materials can be consolidated to obtain a probability density distribution of C/E`s that very much resembles a Gaussian distribution centered at 1.04. The concept of `quality factor` is defined and elaborated so as to take cognizance of observed uncertainties on prediction ofmore » nuclear heating for all the nine materials. 45 refs., 69 figs., 9 tabs.« less

Radiation shielding analysis of a large helical device

Abstract Radiation shielding analysis of a large helical device (LHD) has been performed to provide information on a radiation environment needed for equipment design and operation planning. First, the applicability of the JENDL-3 library which was applied to the shielding calculations of the LHD was tested by means of a benchmark experiment on the shielding efficiency of ordinary concrete carried out at the fusion neutronics source facility at JAERI. Benchmark analysis was carried out with the MCNP continuous energy Monte Carlo code to test the JENDL-3 library itself. The same benchmark calculation was carried out with the DOT3.5 code to examine the groupwise cross-section library of fusion-40 processed from JENDL-3. Secondly, the evaluation of the following items with the DOT3.5 code was revised, using fusion-40 instead of the formerly used GICX-40 library: (1) dose distribution inside and outside the LHD building; (2) dose distribution inside the cellar of the LHD building resulting from steamed radiations; (3) Environmental dose distribution resulting from sky-shine effect; (4) activity and dose rate levels of vacuum vessel and superconducting magnet.

Direct nuclear heating measurements in fusion neutron environment and analysis

Experimental measurement of nuclear heating rates has been carried out in a simulated D-T fusion neutron environment from 1989 through 1990 under the USDOE/JAERI collaborative program at the Fusion Neutronics Source Facility. The microcalorimetric technique has been employed for online measurements. Small probes of materials have been irradiated in close vicinity of a rotating target. A typical probe contains a core measuring 2 cm in diameter by 2 cm in length. Probes of leading candidates, for different applications, have been investigated: molybdenum, tungsten, titanium, graphite (plasma facing components), copper (magnet coils), iron, stainless steel 304, nickel (structural material components) and aluminum. The measured temperature-change rates range from 30 μK/s (iron) to 330 μK/s (graphite). The corresponding nuclear heating rates range from ∼ 35 μW/g (tungsten) to ∼ 225 μW/g (graphite). These measurements have been analyzed using three dimensional Monte Carlo code MCNP and various heating number/kerma factor libraries. The ratio of computed to measured heating rates shows large deviation from 1 for all the materials. In addition, there is a large spread for different libraries; for example, this ratio varies from 1.03 to 1.81 for aluminum. Also, there have been three experiments with each having a host medium of iron, graphite or copper, that measures 85 mm in diameter by 100 mm in length. Small single probes of the host medium graphite and tungsten are placed inside to measure the spatial profile of heat deposition. Analysis of the measurements shows that the ration of computed to measured rates varies widely, e.g., in iron host, it goes from 0.5 to 1.1. Further effort is to be invested to locate the sources of this discrepancy.

Analyses and Intercomparison for Phase I Fusion Integral Experiments at the FNS Facility

Phase I integral experiments of U.S./JAERI Collaborative Program on Fusion Breeder Neutronics which are carried out at the Fusion Neutronics Source (FNS) facility at JAERI ranged from D-T neutron source characterization experiments, tritium production rate (TPR) measurements in a reference Li/sub 2/0 assembly, first wall experiments with and without coolant simulation and beryllium neutron multiplier experiments in various configurations. Both U.S. and Japan have independently analyzed these experiments using their own data base and codes. Analytical predictions obtained by both countries are compared to measured values. Results of this intercomparison is presented in this paper.