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

Plasma Science and Technology 2023年第12期25卷 115-121页

作者： H L SWAmI S VALA m RAJPUT m abhangi Ratnesh KUmAR A SAXENA Rajesh KUmAR Institute for Plasma Research Gandhinagar 382428India Homi Bhabha National Institute Training School ComplexMumbai 400094India

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.

关键词： neutron generator neutronics accelerator physics design nuclear fusion

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

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Plasma Science and Technology 2019年第6期21卷 147-152页

作者： H L SWAmI m abhangi Sanchit SHARmA S TIWARI A N mISTRY V VASAVA V mEHTA S VALA C DANANI V CHAUDHARI P CHAUDHURI Institute for Plasma Research Gandhinagar Gujarat 382428 India Homi Bhabha National Institute(HBNI) Anushaktinagar Mumbai 400094 India Pandit Deendayal Petroleum University Gandhinagar Gujarat 382007 India

A shield module is associated with an Indian Test Blanket module (TBm) in ITER to limit the radiation doses in port inter-space *** shield module is made of stainless steel plates and water *** is identified as an important component for radiation protection because of its radiation exposure control *** radiation protection classification leads to more assurance of the component *** order to validate and verify the design of the shield module,a neutronic laboratory-scale experiment is designed and *** experiment is planned by considering the irradiation under a neutron source of 14 meV and yields of 1010 n *** 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 *** 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),*** neutronic design simulation is assessed using mCNP5 and FENDL 2.1 crosssection *** paper covers neutronic design,irradiation and the outcome of the experiment in detail.

关键词： TBm shield module neutronic experiment ITER mCNP neutron attenuation

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