TY - JOUR
T1 - Calculation of thermal neutron scattering data of MgF2 and its effect on beam shaping assembly for BNCT
AU - Hu, Jiaqi
AU - Qiao, Zhaopeng
AU - Fan, Lunhe
AU - Tang, Yongqiang
AU - Cao, Liangzhi
AU - Zu, Tiejun
AU - He, Qingming
AU - Li, Zhifeng
AU - Wang, Sheng
N1 - Publisher Copyright:
© 2022 Korean Nuclear Society
PY - 2023/4
Y1 - 2023/4
N2 - MgF2 as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF2 were calculated, based on the phonon expansion model. The structural properties of MgF2 were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF2. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF2 were reduced by 27.7% and 8.2%, respectively.
AB - MgF2 as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF2 were calculated, based on the phonon expansion model. The structural properties of MgF2 were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF2. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF2 were reduced by 27.7% and 8.2%, respectively.
KW - Beam shaping assembly
KW - Monte-Carlo method
KW - Phonon density of states
KW - Thermal neutron scattering data of MgF
UR - https://www.scopus.com/pages/publications/85150075780
U2 - 10.1016/j.net.2022.12.027
DO - 10.1016/j.net.2022.12.027
M3 - 文章
AN - SCOPUS:85150075780
SN - 1738-5733
VL - 55
SP - 1280
EP - 1286
JO - Nuclear Engineering and Technology
JF - Nuclear Engineering and Technology
IS - 4
ER -