Point Neutron Source Research Articles

Absolute calibration of the JT-60U neutron monitor using a 252Cf neutron source (abstract)a)

Absolutely calibrated measurements of the neutron yield are important for the evaluation of the plasma performance such as the fusion gain Q in D-D operating tokamaks. The time-resolved neutron yield is measured with 235U and 238U fission chambers and 3He proportional counters in the JT-60U tokamak. The in situ calibration was performed by moving the 252Cf neutron source toroidally through the JT-60 vacuum vessel. Detection efficiencies of three 235U and two 3He detectors were measured for 92 locations of the neutron point source in toroidal scans at two different major radii. The total detection efficiency for the torus neutron source was obtained by the averaging the point efficiencies over the whole toroidal angle. The uncertainty of the resulting detection efficiency for the plasma neutrons is estimated to be ±11%.

Activation Analysis for the Cavity of the PROMETHEUS ICF Design

AbstractActivation analysis for the cavity of the PROMETHEUS ICF design, which uses a wetted first wall protection scheme, has been performed. It has been found that the PROMETHEUS cavity produces about 0.9 Ci/W of thermal power at shutdown after the full 30 years operation, which is about the same amount of radioactivity of other ICF and MCF fusion designs with low activation materials. It was possible, however, to reduce the radioactivity inventory level in the shield by one to three orders of magnitude by introducing a new shield design that uses B4C, Pb, SiC and water instead of using the conventional concrete shield. Furthermore, the effect of using the spherical and cylindrical modeling on the prediction of radioactivity in the first wall has been studied. It has been found that the cylindrical model with a point neutron source at the center of the cylinder reduces the radioactivity of the short half-life products to about 80% of the values that would be obtained by using purely spherical modeling. ...

Neutron Yield Measurement in Tokamaks

Absolutely calibrated measurements of the neutron yield are important for the evaluation of the plasma performance such as the fusion gain Q in D-D operating tokamaks. Time-resolved neutron yield is measured with neutron detectors using fission chamber in large tokamaks. The neutron detectors are calibrated by moving the 252Cf neutron source toroidal ly through the vacuum vessel. In JT-60U, the detection efficiencies were measured by the toroidal scan of the neutron point source resulting the uncertainty of ±11%. The maximum neutron yield of 1.3×1016n/s and the fusion gain QDT of 0.2were obtained in the initial D-D experiment of JT-60U.

Overview of the latest experiments under the JAERI/USDOE collaborative program on fusion neutronics

Abstract The JAERI/USDOE collaborative program is going on using the FNS facility at JAERI. The phase-III is a new series on tritium breeding measurements with a totally different experimental arrangement from the previous ones that used a point neutron source geometry. The new series basically consists of an annular test blanket and a pseudo-line source to investigate the effect source spread on the neutronic performance. A concise description is made on the features of the simulated line source, the test blanket systems for phase-IIIA and -IIIB, measured items, experimental results, and their analysis. Within the same program, nuclear heating rate, gamma-ray activities and resultant energy releases have been measured to provide experimental data for safety related issues. A comparative study was also performed among the phase-I, phase-II and phase-III.

Dense Plasma Focus as a Neutron Source for Fusion Research

The properties of dense plasma focus, a point source of fusion neutrons with high flux density, are described and its applications to problems in fusion technology are discussed, including assessme...

In situ calibration of TFTR neutron detectors

We report results of the TFTR fission detector calibration performed in December 1988. A NBS-traceable, remotely controlled 252Cf neutron source was moved toroidally through the TFTR vacuum vessel. Detection efficiencies for two 235U detectors were measured for 930 locations of the neutron point source in toroidal scans at 16 different major radii and vertical heights. These scans effectively simulated the volume-distributed plasma neutron source and the volume-integrated detection efficiency was found to be insensitive to plasma position. The Campbell mode is useful due to its large overlap with the count rate mode and large dynamic range. The resulting absolute plasma neutron source calibration has an uncertainty of ±13%.

Review of Scattering Corrections for Calibration of Neutron Instruments

In this paper we consider scattering corrections that must be made when neutron personnel instruments are calibrated with a point source of neutrons near the centre of a calibration room. We consider three basic types of scatter: (1) scattering from a plane slab, (2) scattering from walls of an enclosed room, and (3) air scatter. In each case we proceed from an idealised situation, where the correction can be predicted analytically, to an actual calibration room, where predictions must be checked by Monte Carlo calculations or experiment. Calculations show that the postulation of an image source provides a good model for predicting the fluence of neutrons scattered from aluminium and concrete slabs. Calculations also show how the fluence of neutrons scattered from the walls of an enclosed room relates to the case of neutron fluence in a spherical cavity and how the neutron fluence is increased by scattering in the air around source and detector. Finally, calculations are compared with experiments on the response of a neutron dose equivalent monitor.

A Comparative Study of Thermal Neutron Fusion Blanket Arrangements

Four basic configurations, block, homogeneous, multilayered, and heterogeneous systems, for tritium breeding in natural lithium water-reflected fusion blankets are compared. A 14-MeV point neutron source is used with a lead multiplier. All the calculations are done in rectangular geometry with the Monte Carlo code MORSE-E and Los Alamos National Laboratory's 30-group neutron cross-section set CLAW-IV in P/sub 3/ scattering approximation. The homogeneous system gives the best and the block concept gives the lowest tritium breeding, but neither of these concepts is practical for use in fusion blankets. Among the practical blanket arrangements, for equal amounts of natural lithium and water, the heterogeneous and multilayered arrangements give almost identical tritium breeding.

Measurements of the Neutron and Gamma-Ray Fluences in the TFTR Test Cell Due to a Point Source Simulating D-T Fusion Plasma Neutron Production

Spectral distributions of high-energy neutrons (0.9 less than or equal to E/sub n/ less than or equal to 14.5 MeV) and of high-energy gamma rays (0.4 less than or equal to E/sub ..gamma../ less than or equal to 9.4 MeV) due to a deuterium-tritium (D-T) neutron point source simulating the extended fusion plasma neutron source in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory are reported. A D-T neutron generator was positioned inside the vacuum vessel at ten different locations around the torus. Neutrons and gamma rays were detected by a liquid-scintillator-based detector (4.65-cm diam X 4.22 cm high) with electronic pulse-shape discrimination to differentiate between events in the detector due to incident neutrons and those due to incident gamma rays. The detector was placed on the median plane of the reactor at 8.85 m from the geometric center of the TFTR. Two spectral distributions, one for neutrons and the other for gamma rays, were obtained for each of 18 measurements. The neutron data exhibit a high-energy peak dominated by uncollided primary-energy neutrons and a low-energy contribution from the scattered neutrons. The gamma-ray data exhibit a high-energy contribution due to neutron capture gamma rays and amore » low-energy contribution due to gamma rays following neutron inelastic scattering reactions.« less

Measurements and calculations on the shielding effect of small slabs for 14 MeV neutrons

Abstract The shielding effect of small slab assemblies containing graphite, lead, lithium fluoride and uranium has been tested for 14 MeV source neutrons. Emergent neutron spectra were measured with an NE213 scintillation counter and gamma ray events were rejected by pulse shape discrimination. Calculations have also been made using the one-dimensional transport code ANISN and the ZZFEWG1/31-1F cross section data set, which is in P3 form, over the energy range 14 to 1.8 MeV. The small slabs with a point neutron source were simulated by using a spherical calculation with a nonreflecting inner boundary. Agreement between measurement and calculation indicates that at least for the materials studied the cross section set and calculation model are suitable for studying the behaviour of thin shields, such as sections of a fusion reactor blanket.

Blanket material and engineering issues, and requirements for experiments and facilities

Fusion nuclear technology components face a new and tougher environment than experienced by any previous nuclear application. The development of feasible and attractive fusion nuclear components will require a broad research effort including experiments, model development and innovative designs. In order to provide an efficient strategy, the important issues must be identified and characterized, and required experiments and facilities must be defined. This paper considers the experiments needed to solve the uncertainties associated with liquid breeder and solid breeder blankets. Many of these uncertainties are strongly material-related and involve a number of areas ranging from basic material properties under irradiation to highly interactive phenomena involving several materials, components and/or environmental conditions. They can be addressed in a number of progressively more integrated experiments. Test facilities to investigate these uncertainties range from simple bench-top experiments, to large non-nuclear tests, fission reactors and point neutron sources, and fusion test facilities. Possible experiments and facilities have been analyzed in a timeframe consistent with the present US goal of supporting an assessment of the feasibility and attractiveness of fusion around the year 2000.

Overview of the TFTR Lithium Blanket Module Program

The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests.

Summary of Recent Results from the Jaeri/U.S. Fusion Neutronics Phase I Experiments

The tritium breeding ratio (TBR) in a fusion blanket is one of the most important parameters to be considered in the conceptual design of fuel-self-feeding deuterium-tritium (D-T) reactors. There are many factors, both physics and engineering, that affect the TBR value in a design. Efforts in the neutronics area are to provide data needed for identifying and selecting materials and configurations that achieve an optimal TBR, and to improve the accuracy of neutronics predictions. A program is in progress on integral neutronics experiments and their analyses in the collaboration between the Japan Atomic Energy Research Institute (JAERI) and the US Dept. of Energy. In the phase I experiments, a simplified blanket test module of Li/sub 2/O is set at a portion of an enclosure with a 14-MeV point neutron source at its center. The test module is composed of unit size blocks and is able to mock up the complexity of radial configurations of various blanket concepts in full size. Neutronic quantities are measured along the central axis of the module. Main concerns are the tritium production rate (TPR) and the neutron spectrum. New measuring techniques have been developed to meet the requirements of this program in both on-line typemore » and time-integrated type measurements.« less

Tritium Assay of Li2O in the LBM/LOTUS Experiments

The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM. PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results form the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system.

The LBM Program at the EPFL/LOTUS Facility

An experimental program of neutron transport studies of the Lithium Blanket Module (LBM) is being carried out with the LOTUS point-neutron source facility at École Polytechnique Fédérale de Lausanne (EPFL), Switzerland. Preliminary experiments use passive neutron dosimetry within the fuel rods in the LBM central zone, as well as, both thermal extraction and dissolution methods to assay tritium bred in Li2O diagnostic wafers and LBM pellets. These measurements are being compared and reconciled with each other and with the predictions of two-dimensional discrete-ordinates and continuous-energy Monte-Carlo analyses of the LOTUS/LBM system.

Validated Deep-Penetration, Air-Over-Ground, Neutron/Gamma-Ray Transport

The results of a program are reported whose objective has been to establish the reliability and accuracy of tissue kerma estimates near the ground, out to deep penetration ranges, from a point neutron source in an air-over-ground geometry. The results take the form of expected error in calculated neutron and secondary gamma-ray kerma out to 2-km range for any neutron source height and energy spectrum. In the first of two approaches, experimental data permitting absolute evaluation in one dimension is used in conjunction with an evaluated calculational procedure for two dimensions to obtain overall error estimates. In the second approach, errors obtained from comparisons of measurement and calculation in air-over-ground geometry are evaluated to obtain overall error estimates. When the results of these two approaches are averaged, it can be concluded with confidence that kerma to 2 km will probably be calculated to be 10 to 15% lower than measured values for neutrons and 20 to 25% lower for gamma rays when this cross-section set and recommended calculational procedure or equivalents are used.

Application of the Image Method to Thermal Neutron Leakage Problems

The applicability and parameters of the image method for solving the thermal neutron leakage distribution on the plane surface of the semi-infinite hydrogen-containing medium in the case of the neutron field originating from a point source of fast neutrons is investigated.The research performed has shown that for the use of the image method it is necessary to introduce not only the concept of the effective extrapolated boundary but also the concept of an effective origin of the point source of fast neutrons.The values of effective extrapolated length and effective location of the point source were calculated and examined by experiment for pure water.

Tokamak Fusion Generators for Nuclear Radiation Effects Testing

The potential utilization of near-term tokamak test reactors for nuclear radiation damage evaluation and hardening programs is examined. Physical layouts of the irradiation modules and methods of neutron spectrum tailoring are described. Tokamak fusion generators provide much larger test volume and can give more faithful spectral simulations of thermonuclear threat environments than do accelerator-driven point neutron sources or pulsed fast-fission reactors. However, the tokamak is basically a long-pulse source producing relatively low neutron dose rates (~105 rads/s) that limit its application to damage programs with devices and subsystems that are sensitive to integrated radiation dose rather than dose rate.

The Attenuation of 14-Mev Neutrons from a Point Source Embedded in Concrete Shields

Measurements have been made of the thermal neutron fluxes at various distances from a point source of 14-MeV neutrons embedded in three large shields. The first shield consisted of a 24-m3 assembly of solid concrete building blocks of density 2.2 g/cm3 and gave an attenuation length of 14.1 ± 0.7 cm at a distance of 150 cm from the source. The second shield was a layered assembly of wood and concrete blocks with a mean density of 1.92 g/cm3 and gave an attenuation length of 15.7 ± 0.7 cm at the same distance. The third assembly consisted of a cube of side 61 cm of steel surrounded by concrete blocks; at a distance of 120 cm from the source it gave an attenuation length of 9.3 ± 0.4 cm.

Radiation Transport in Earth for Neutron and Gamma-Ray Point Sources above an Air-Ground Interface

Two-dimensional discrete ordinates methods were used to calculate the instantaneous dose rate in silicon and neutron and gamma ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 meters) above an air--ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth--concrete model containing 0.9 meters of borated concrete beginning 0.5 meters below the earth's surface to obtain fluence distributions to a depth of 3.0 meters. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon and, in all cases, the secondary gamma ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths greater than 0.6 meter. 4 figures, 4 tables.