Design and simulation of a thermal neutron beam for neutron capture studies at the Dalat research reactor
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https://doi.org/10.15625/0868-3166/17445Keywords:
MCNP, design, crystal Al2O3, sapphire, thermal neutron fluxAbstract
This paper presents the application of the MCNP5 code to conduct a modification design and simulation of a thermal neutron beam for neutron capture studies at the Dalat Nuclear Research Reactor (DNRR). The designed configuration of the horizontal neutron channel No.2 at the DNRR, which contains a conical collimator of 240.3 cm in length, and neutron filters of crystal Al2O3 and Bi with alternative thickness, were simulated. A pure thermal neutron beam can be obtained at the irradiation position when a composition of crystal filter of 20 cm Al2O3 and 6 cm Bi is employed. The thermal and epithermal neutron fluxes are 1.02´108 n/cm2/s (account for 97.92% of total neutron flux) and 0.22´107 n/cm2/s (account for 2.08% of total neutron flux), respectively.
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