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.

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas.The shield module is made of stainless steel plates and water channels.It is identified as an important component for radiation protection because of its radiation exposure control functionality.The radiation protection classification leads to more assurance of the component design.In order to validate and verify the design of the shield module,a neutronic laboratory-scale experiment is designed and executed.The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s-1.The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials.The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene),respectively.The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 crosssection data.The paper covers neutronic design,irradiation and the outcome of the experiment in detail.