MONTE-CARLO SIMULATION OF QUASI-INFINITE DEPLETED URANIUM TARGET IRRADIATED BY 1…10 GeV DEUTERON AND PROTON BEAM

Simulation of a ~21 t depleted uranium target irradiated by 1…10 GeV proton and deuteron particles with the help of FLUKA simulation package was carried out. Neutron spectra and neutron flux in a target volume were obtained. Total number of 235U (n,f), 238U(n,f) reactions occurred in a target were determined. Beam particle power multiplication are calculated. The calculations were performed for the purpose of planning experiments on irradiation of a uranium target (22 tons of depleted uranium) at JINR (Dubna) within the framework of the international project “Energy and Transmutation of RAW”.

Establishment of neutron dose calibration fields based on a ²⁵²Cf source by Monte-Carlo method

This paper presents calculation results based on Monte-Carlo method to select an appropriate neutron moderator and design four configurations for a 252Cf irradiation system. These configurations provide six neutron spectra with the various average energies (1.04 MeV, 1.38 MeV, 1.69 MeV, 2.05 MeV, 2.46 MeV and 2.93 MeV) suitable for the calibration of neutron survey meters and personal dosimeters.

Calculation of critical core configurations of a research reactor using MTR, IRT-4M, VVR-KN fuel assemblies

This paper presents calculation results to determine critical core configurations and aminimum number of fuel assemblies (FAs) or uranium mass of a research reactor loaded with three types of FAs such as MTR, IRT-4M and VVR-KN. The MCNP5 code and ENDF/B7.1 library were applied to estimate characteristics parameters of the fuel types and the whole core. Infinitive multiplication factor kinf, neutron flux distribution and neutron spectra of the fuels were calculated. The reactor core configurations with three fuel types were modeled in 3-dimensions, and then the effective multiplication factors keff, relative radial power distribution of each configuration were also evaluated. From calculation results, twelve fuel loading schemes were chosen based on lowest uranium mass or smallest number of FAs loaded into the core. In addition, two full core configurations using VVR-KN and MTR FAs and consisting of beryllium reflectors, vertical irradiation facilities, horizontal neutron beam ports, etc. have been proposed for further consideration in thermal hydraulic calculations and safety analysis.

Serpent 2 Validation for Radiation Shielding Applications

This paper contributes to the validation of Serpent's photon transport and coupled neutron-photon transport routines. Two benhmarks presenting measurements of neutron and photon flux through different sized iron and lead spheres have been calculated using a development version of Serpent and MCNP6.2. The Serpent results were compared to the measurement results and the MCNP6.2 calculations. Additionally, the development version has been compared to the currently distributed Serpent version 2.1.31. In all cases, the Serpent calculated neutron and photon spectra followed the measured spectra fairly well. For the iron spheres, differences between Serpent and MCNP6.2 calculated neutron spectra were mostly below 2 % at neutron energies below 4 MeV. Differences between photon spectra through the iron spheres were mostly below 3 %. For the lead spheres, differences in the calculated neutron spectra were mostly below 1.5 % in the energy range 0.04-4.0 MeV. Differences between photon spectra through the 10 cm lead sphere were mostly below 5 % and for the larger spheres below 10 % except at higher photon energies above 6.5 MeV. Differences between the development version and the Serpent version 2.1.31 of the order of 3 % were observed in the photon spectra through the largest lead spheres with radius 20 and 30 cm when Gaussian Energy Broadening was not applied. These are probably related to the coupled neutron-photon transport routines in the different versions.