Open AccessCalculation of the neutron flux distribution in the accelerator driven subcritical reactor with (Th-233U)O2 and (Th-235U)O2 mix fuel
Author(s) -
Tran Minh Tien,
Trần Quốc Dũng
Publication year - 2020
Publication title -
journal of physics. conference series
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.21
H-Index - 85
eISSN - 1742-6596
pISSN - 1742-6588
DOI - 10.1088/1742-6596/1451/1/012009
Subject(s) - neutron flux , neutron temperature , neutron , flux (metallurgy) , radius , nuclear physics , neutron cross section , neutron poison , physics , rod , neutron moderator , nuclear reactor core , materials science , nuclear engineering , medicine , computer security , alternative medicine , pathology , computer science , metallurgy , engineering
This paper presents results of calculating the neutron flux distribution in an accelerator driven subcritical reactor (ADSR) with (Th- 233 U)O 2 and (Th- 235 U)O 2 mixed fuel. An ADSR consists of 90 fuel rods, and 10 graphite reflector rods. All objects are placed in liquid lead. Thorium is replaced by mixture of (Th- 233 U)O 2 and (Th- 235 U)O 2 ; MCNP5 program has been used to calculate radial distribution of the neutron flux, axial distribution and energy distribution from (p, n) reaction.The calculation results show that the axial distribution of the thermal and fast neutron flux reduce from the center of core but reduction rate is different.The thermal neutron flux decreases gradually from 0 to 2.5cm; decreases rapidly from 25cm to 5cm. In comparison, the thermal neutron flux is smaller than fast neutron flux from 0 to 4cm along the radius but the thermal neutron flux is larger than fast neutron flux at distances greater than 5cm along the radius of the reactor.