The Japan Atomic Energy Research Institute (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.

The neutron source characteristics of the Japan Atomic Energy Research Institute (JAERI)/U.S. Department of Energy collaborative program on fusion neutronics Phase-IIA and -IIB experiments are determined by measuring neutron spectra and various activation rates in the cavity and on the inner surface of the enclosure and the test regions. The analyses are performed by both JAERI and the United Stares using individual nuclear data and transport codes. The neutron spectra are generally well predicted by both Monte Carlo and S-n calculations in the energy range of 15 MeV to a few kilo-electron-volts, except for energies 10 to 1 MeV. The discrepancies between the measured and the calculated activation rates are within +/-10% when recently evaluated nuclear data are used. Through the present investigation, the characteristics of incident neutrons in the test region can be satisfactorily predicted.

The experiments performed in the Japan Atomic Energy Research Institute/U.S. Department of Energy collaborative program on fusion blanket neutronics are designed with consideration of geometrical and material configurations. The general guide that is used to design the engineering-oriented neutronics experiment, which uses an accelerator-based 14-MeV neutron source, is discussed and compared with neutronics characteristics of the reactor models. Preparation of the experimental assembly, blanket materials, and the neutron source is described. A variety of techniques for measuring the nuclear parameters such as the tritium production rate are developed or introduced through the collaboration as a basis of the neutronics experiments. The features of these techniques are discussed with the experimental error and compared with each other.

Neutronics experiments for two types of heterogeneous blankets are performed in the Phase-IIC experiments of the Japan Atomic Energy Research Institute/U.S. Department of Energy collaborative program on fusion blanket neutronics. The experimental system uses the same geometry as the previous Phase-IIA series, which was a closed geometry that used a neutron source enclosure of lithium carbonate. The heterogeneities selected for testing are the beryllium edge-on and the water coolant channel assemblies that appear in typical blankets. In the former, the beryllium and the lithium-oxide (Li2O) layers are piled up alternately in the front part of the test blanket. In the latter, the two simulated water cooling channels are emplaced vertically in the Li2O blanket. These channels produce a steep gradient of neutron flux and a significant spectrum change around the material boundary. The calculation accuracy and measurement method for these transient regions are key areas of interest in the experiments. The measurements are performed for the tritium production rate and the other nuclear parameters as well as the previous experiments. The void effect is found to not be negligible around the heterogeneous region for the detector with a low-energy response. At the same time, enhancements of tritium production are seen near the beryllium and hydrogenous material. However, the current Monte Carlo calculation shows good agreement with the experiment even in such a boundary.

A variety of techniques to measure the nuclear parameters, such as tritium production rate, neutron spectrum, reaction rate and gamma-ray heating, in a simulated fusion blanket assembly have been developed or introduced through the JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics. The features of those techniques are summarized and discussed with the experimental error. The present measurement techniques provided data with error ranges of 3–5% for tritium production, 5–10% for the neutron spectrum, 3–6% for the activation reaction and 10–20% for the gamma-ray heating rate.

Fusion neutronics experiments are performed on a full-coverage blanket with various configurations of a beryllium neutron multiplier. In the basic experimental system, a lithium carbonate enclosure contains a lithium oxide test zone and a deuterium-tritium neutron source to simulate a neutron spectrum in a fusion reactor. Five beryllium configurations are adopted to examine the effects of neutron multiplication and reflection by beryllium. The measurements are carried out along the central line in the test Zone. Various measurement techniques are applied to obtain the tritium production rate distribution, which is one of the most important parameters for assessing the fetal tritium breeding ratio in a fusion blanket. In addition, the reaction rates and the neutron spectrum are measured to provide test data for confirmation of calculation results. These data are compared among six different configurations of the experimental system. Consistency between the different techniques for each measured parameter is also tested among different experimental systems. The experimental results are compared with the calculations by DOT3.5 using JENDL9/PRI and /PR2. The calculation differs from the experimental data by < 10%, except for the beryllium zone.

An innovative liquid-metal reactor, the Integral Fast Reactor (IFR), is being developed at Argonne National Laboratory. One characteristic of the IFR is the fuel cycle closure. Fissile material bred and fissionable material produced in the reactor are recycled back into the reactor. Waste generated during fuel reprocessing will be packaged into special waste canisters and will be shipped to a repository for final disposal. Prior to its removal from the facility, a measurement of the fissile content will be necessary as a part of an overall fissile material inventory accountability system. A particular form of nondestructive assay called delayed neutron nondestructive assay (DNNDA) is being developed to assist in the establishment of an accountability system. A preliminary neutronics investigation for the current DNNDA has been made to assist and verify the characteristics of the design from a neutronic aspect. A 10(11) n/s, 14-MeV neutron source would provide adequate counting statistics for fissile material at the milligram to gram level.

A pseudo-line D–T neutron source has been developed with new experimental techniques. This line source was applied in sophisticated neutronics experiments for an annular blanket arrangement simulating the tokamak geometry, as a new series in the JAERI/USDOE collaborative experimental program on fusion neutronics. The source characteristics of the present line source and the measurements for an annular assembly are described. The discussion on the experimental results focuses on the tritium production rate measured in an annular blanket and comparisons were made with the previous point source experiment, and also between the annular blankets with and without an armor reflector of graphite.

The experimental measurement of low energy neutron spectra in a large iron assembly is essential to examine nuclear data and neutron transport codes for nuclear heating estimations in the superconducting magnet of a fusion reactor as well as shielding design. In-situ neutron spectra between a few keV and 1 MeV in a 0.95 m thick cylindrical iron assembly, bombarded with D-T neutrons, have been measured, using a proton recoil gas proportional counter (PRC). A newly developed data acquisition system for PRC was adopted, where high voltage varies continuously in ramped shape during acquisition. Experimental analyses were performed by the DOT3.5 and MCNP codes with cross section sets based on the JENDL-3 nuclear data library. From the comparison between the measured and calculated neutron spectra, the following facts were obtained. The calculation by DOT3.5 overestimated the experiment by a factor of about 2 for the flux below 500 keV in the front parts of the iron assembly and underestimated by more than a factor of 2 for the flux above 500 keV in the rear parts. The calculation by MCNP gave a general agreement with the measurement, but the discrepancy above 100 keV increased as the detector position moved deeper into the assembly. The calculation above 500 keV underestimated the measured spectrum by more than 30% at the 0.81 m depth.

Gamma heating measurements have been made in a low-Z assembly irradiated with 14-MeV neutrons and (n,n') gammas produced by a Texas Nuclear Model 9400 neutron generator. The assembly is composed of 144 magnesium sleeves (5cm × 5cm × 60cm × 2 mm thick) filled with graphite to simulate a fusion blanket test module. Heating measurements were made in the mid-line of the assembly using a proportional counter operating in the Continuously-varied Bias-voltage Acquisition (CBA) mode. The neutron induced atomic recoil signal was rejected by observing the signal rise-time differences inherent to radiations of different LET. The experiment was modelled using the one-dimensional radiation transport code ANISN/PC. The operating limits of this technique were identified by comparing measurements made at different positions in the assembly and then comparing these measurements to the calculated flux.

A new data acquisition technique (the continuously-varied bias-voltage acquisition mode) has been developed and tested for the low-flux broad-energy regime characteristic of existing fusion blanket mockups. This method of analysis allows for the acquisition of data spanning several orders of magnitude in energy with a single proportional counter. Utilizing this method, the gamma energy deposition in a mixed neutron and gamma field was measured.

As the first period of the Phase II series of the JAERI/USDOE collaborative program, neutronic parameters have been measured for a simulated Li2O/Be breeder blanket in closed geometry. The experimental system consists of a lithium-oxide test zone and a lithium carbonate enclosure containing a DT neutron source at the Fusion Neutronics Source (FNS) facility at JAERI. Tested blankets were of three 5 cm thick configurations of beryllium neutron multiplier zone. The experiments were performed to examine spatial distributions of reaction rates and the neutron spectrum in the source cavity, and relative profiles of the tritium production rate (TPR), reaction rates, and neutron spectra between the beryllium configurations. A zonal TPR measuring technique, suitable especially for direct comparison with a Monte Carlo method, was applied to a steep gradient distribution. The experimental results of TPRs showed that the beryllium sandwiched system provided the most effective TBR gain (integrated TPR) of about 20% compared with the non-multiplier system. The reaction rate distributions and neutron energy spectra were also provided to test a calculational code system for nuclear design.

As a part of the Phase-II experimental series of JAERI/USDOE collaborative program on fusion blanket neutronics, the phase-IIB experiment has been performed. The experiment provides information of neutron multiplication and reflection by the inner berryllium layer in a full-coverage blanket geometry. The measurements were carried out at the positions in the test zone on tritium production rate (TPR) using various methods, on reaction rate using foil activation technique and on neutron energy spectrum using NE213 and gas proportional counters. The experimental results showed that the effect of the full-coverage beryllium was a 10% increase for T7 (TPR for 7Li) and a factor of 2–5 increase for T6 (TPR for 6Li). The increase of the integrated TPR for natural lithium (Tn) in the test zone due to the inner beryllium layer was above 60% compared to the non-beryllium system in the Phase-II geometry.

ABSTRACTNeutron parameters were measured in Phase-I experiments of JAERI/USDOE collaborative program. A brief description of the experimental systems and measuring techniques is given. The range of measured source neutron spectrum is from a few keV to 16 MeV by the use of proton-recoil counters and a small NE213 spectrometer. Tritium production rates were measured by on-line and integral type techniques. The in-system scalar spectra were measured by the NE213 spectrometer. Foil activation method was applied to obtain the reaction rate distributions.

A brief description of the experimental tools available for fusion neutronics experiments is given. Attention is paid to error estimates mainly for the measurement of tritium breeding ratio in simulated blankets using various techniques.

Measurements of degraded fission-neutron spectra using recoil proportional counters are done routinely for studies involving fast reactor mockups. The same techniques are applicable to measurements of neutron spectra required for personnel dosimetry in fast neutron environments. A brief discussion of current applications of these methods together with the results of a measurement made on the LITTLE BOY assembly at Los Alamos are here described.