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 componen...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.展开更多
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...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.展开更多
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 mus...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.展开更多
文摘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.
基金Supported by Innovation Fund for Small Technology-based Firms (99C26212210085)
文摘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.
基金Supported by National Natural Science Foundation of China (10105003)
文摘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.
