A simulation on desired neutron flux for the boron neutron capture therapy (BNCT) purpose by using Monte Carlo N-Particle (MCNPX)

Abstract

Boron Neutron Capture Therapy (BNCT) is a nuclear application to cure cancer by using the interaction between the boron compound and the slow (thermal and epithermal) neutron beam obtained from the neutron source. The previous feasibility study shows that the thermal column can generate higher thermal neutrons for BNCT application at the TRIGA MARK II reactor. Currently, the BNCT facility is planned to be developed at the thermal column. Thus, to fabricate the nuclear application such as BNCT required several required such as the study of design, material, neutron desired and safety for the application. Therefore, the use of simulation software such as Monte Carlo N-Particle (MCNP) was extensively essential to identify and determine the most convenient design of the BNCT facility with proper safety requirement standard. Besides that, the simulation also useful in terms of identifying good material and best geometry and the uses of simulation can guide to make sure this research achieves the best results. Generally, MCNP function to track the particle over its energy range using the irradiation transport and the neutron flux and the gamma flux produced from the simulation using MCNPX code was consider right if the relative errors contributed was lower than 5%. This research simulates that the uncovered collimator provides high neutron flux and high collimator performance compared to the fully covered collimator. This research is useful in the simulate on the desired neutron from the BNCT facility design at the TRIGA MARK II reactor.

Shielding Design and Dose Evaluation for HTR-PM Fuel Transport Pipelines by QAD-CGA Program

“…e Monte Carlo method can deal with complex geometric structures, material composition, and source terms, by simulating the real particle transport process, such as scattering, transmission, and absorption. Widely used software based on the Monte Carlo method includes MCNP [7][8][9], FLUKA (http://www.fluka.org/fluka.php; [10], GEANT4 [11,12], and so on. However, the Monte Carlo method usually requires a longer calculation time.…”

Monte Carlo Simulation and Experimental Validation for Radiation Protection with Multiple Complex Source Terms and Deep Penetration for a Radioactive Liquid Waste Cementation Facility

“…Traditional theoretical formula prediction methods, such as the QAD program [10][11][12][13] which is based on pointkernel integration technology [14][15][16], can quickly perform the shielding calculation, but usually, the calculation results are too conservative. Currently, there are many Monte Carlo programs widely used, such as MCNP [17][18][19], FLUKA (http://www.fluka.org/fluka.php) [20], and GEANT4 [21,22], which have been verified, can provide very accurate simulation results [23].…”