Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasmafacing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 10^(12)n s^(-1). The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 10^(12)n s^(-1)with a 110 keV D^(+) ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 10^(12)n s^(-1). Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.

^(235)U enrichment detection system for nuclear fuel rod based on compact D–D neutron generator

A reliable ^(235)U enrichment uniformity detection system based on a compact D–D neutron generator is developed to detect the ^(235)U enrichment uniformity of different fuel elements in the same nuclear fuel rod. The high-yield compact D–D neutron generator provides 2.45 MeV D–D neutrons, decelerated by a moderator to thermal neutrons or epithermal neutrons, thereby inducing ^(235)U fission to produce highly excited state fission fragments that undergo de-excitation via γ-ray emission. The system detects the ^(235)U enrichment uniformity of a nuclear fuel rod by measuring γ-rays and establishing a relationship between the γ-ray count rate and ^(235)U enrichment in nuclear fuel. The proposed system yields a confidence probability of 99.99% for a relative ^(235)U enrichment deviation of 10% and a neutron yield of 5 × 10^(8) n/s, and the detection accuracy increases with increasing neutron yield. Furthermore, the developed system can satisfy quality control requirements for nuclear fuel production to promote the safe development of nuclear power.

Fluorine Assessment of Igneous Rocks by Neutron Activation Analysis

The main goal of the current study was to determine the fluorine in the rock samples coal (SARM-18, SARM-19, and SARM-20), opal glass NBS91 and phosphate rock NBS694 using neutron activation analysis. Neutrons energy of 14 MeV used for irradiation was produced by bombardment of a water-cooled titanium tri-tide target with a beam of deuterons accelerated to a potential of 350 KV to develop a neutron flux (108 n⋅cm-2⋅s-1) on the sample at the neutron generator in the ECN (Netherlands Energy Research Foundation) Petten. This new approach contributes to the existing knowledge of fluorine measurement by the coincidence channels investigation of the positron energy with respect to decay time for each radionuclide element. The present study was designed to determine the fluorine by fast neutron through the reaction of F19 (n, 2n) F18. Interference was treated by irradiating the standard of these elements and reasonable selection of the decay time between the end of irradiation and beginning of counting time. The results of this method indicate that the concentration of fluorine is agreed fairly with literatures.

In-situ Calibration Techniques and Preliminary Assessment of Accuracy for NFM on ITER

Absolutely calibrated measurements of the neutron yields which need to cover both D-D and D-T phase of the international thermal-nuclear experimental reactor （ITER） are important for the evaluation of fusion power and fusion gain Q in D-D and D-T operations. This paper describes the in-situ calibration techniques and methods, the neutron sources including ^252Cf and neutron generator for calibration, the preliminary accuracy assessment and the error analyses. In addition, some difficult problems regarding the in situ calibration for the neutron flux monitor （NFM） on ITER are presented and discussed.

Improvement of the determination of hydrogen content in a multicomponent sample by D-T generator

If a D T generator is used as a neutron source to simultaneously measure the content of carbon, hydrogen and oxygen in a multicomponent sample by NIPGA （Neutron Induced Prompt Gamma-ray Analysis）, the 14 MeV neutron flux can be regarded as a constant value. The relationship between the production of the hydrogen characteristic gamma-rays and its content is nonlinear. In this paper, we use MCNP （Monte Carlo N-Particle Transport code） to simulate the relationship and analyze it. In practical measurement of the characteristic gamma-ray, it＇s impossible to get the net count. Therefore, we use the experiment to obtain the relationship between the hydrogen content and the total count of its characteristic gamma-rays. If we use the relationship combined with the simulation result to calculate the hydrogen content, the metrical precision can be much increased. The deviation of hydrogen content between NIPGA and chemical analysis is less than 0.25%, which meets the requirement of coal industry.

Designing of the 14 MeV neutron moderator for BNCT

In boron neutron capture therapy （BNCT）, the ratio of the fast neutron flux to the neutron flux in the tumor （RFNT） must be less than 3% If a D-T neutron generator is used in BNCT, the 14 MeV neutron moderator must be optimized to reduce the RFNT. Based on the neutron moderation theory and the simulation results, tungsten, lead and diamond were used to moderate the 14 MeV neutrons. Satisfying RFNT of less than 3%, the maximum neutron flux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer, a 14 cm thick lead layer and a 21 cm thick diamond layer.