Analysis of concrete labyrinth shielding and radiation dose for APF plasma focus neutron source by FLUKA Monte Carlo code

The research aims to measure the radiation dose rate over the radiation shielding which is made of paraffin and aluminium and to determine the best shield material for the safety of radiation workers. The examination used MCNP (Monte Carlo N-Particle) simulator to model the BNCT neutron source and the shield. The shield should reduce radiation to less than the dose limit of 10.42 µSv/h, which is assumed to be the most conservative limit when the duration of workers is 1920 h. The first design resulted in a radiation dose rate which was still greater than the limit. Therefore, optimization was done by adding the lead on the outer part of the shield. After optimization by adding the lead with certain layers, the radiation dose rate decreased, with the largest dose being 57.60 µSv/h. Some locations over the limit could be overcome by other radiation protection aspects such as distance and time. The paraffin blocks were covered by aluminium to keep the shield structure. The lead was used to absorb the gamma ray which resulted from the interaction between the neutrons and aluminium.

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Evaluation of absorbed radiation dose on pregnant and fetus: Monte Carlo code

The Annals of Clinical and Analytical Medicine ◽

10.4328/acam.20224 ◽

2020 ◽

Vol 11 (suppl 3) ◽

Keyword(s):

Monte Carlo ◽

Radiation Dose ◽

Monte Carlo Code ◽

Absorbed Radiation Dose

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Assessment of ocular beta radiation dose distribution due to 106Ru/106Rh brachytherapy applicators using MCNPX Monte Carlo code

International Journal of Cancer Therapy and Oncology ◽

10.14319/ijcto.0203.8 ◽

2014 ◽

Vol 2 (3) ◽

pp. 02038 ◽

Cited By ~ 7

Author(s):

Nilseia Barbosa ◽

Luiz da Rosa ◽

Artur Menezes ◽

Juraci Reis ◽

Alessandro Facure ◽

...

Keyword(s):

Monte Carlo ◽

Radiation Dose ◽

Dose Distribution ◽

Beta Radiation ◽

Monte Carlo Code

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Simulating neutron transport in long beamlines at a spallation neutron source using Geant4

Journal of Neutron Research ◽

10.3233/jnr-190134 ◽

2020 ◽

Vol 22 (2-3) ◽

pp. 183-189

Author(s):

Douglas D. DiJulio ◽

Isak Svensson ◽

Xiao Xiao Cai ◽

Joakim Cederkall ◽

Phillip M. Bentley

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Variance Reduction ◽

Neutron Transport ◽

High Energy ◽

Low Energy ◽

Deep Penetration ◽

Monte Carlo Code ◽

Spallation Neutron Source ◽

Unique Challenge

The transport of neutrons in long beamlines at spallation neutron sources presents a unique challenge for Monte-Carlo transport calculations. This is due to the need to accurately model the deep-penetration of high-energy neutrons through meters of thick dense shields close to the source and at the same time to model the transport of low- energy neutrons across distances up to around 150 m in length. Typically, such types of calculations may be carried out with MCNP-based codes or alternatively PHITS. However, in recent years there has been an increased interest in the suitability of Geant4 for such types of calculations. Therefore, we have implemented supermirror physics, a neutron chopper module and the duct-source variance reduction technique for low- energy neutron transport from the PHITS Monte-Carlo code into Geant4. In the current work, we present a series of benchmarks of these extensions with the PHITS software, which demonstrates the suitability of Geant4 for simulating long neutron beamlines at a spallation neutron source, such as the European Spallation Source, currently under construction in Lund, Sweden.

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Technical feasibility study of BAEC TRIGA reactor (BTRR) as a neutron source for BNCT using OpenMC Monte Carlo code

Progress in Nuclear Energy ◽

10.1016/j.pnucene.2020.103418 ◽

2020 ◽

Vol 126 ◽

pp. 103418

Author(s):

Sharif Abu Darda ◽

Abdelfattah Y. Soliman ◽

Mohammed S. Aljohani ◽

Ned Xoubi

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Feasibility Study ◽

Monte Carlo Code ◽

Technical Feasibility ◽

Triga Reactor

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Neutron Radiography System Collimator Design via Monte Carlo Calculation

18th International Conference on Nuclear Engineering: Volume 3 ◽

10.1115/icone18-29745 ◽

2010 ◽

Author(s):

Hamid Jafari ◽

Majid Shahriari

Keyword(s):

Monte Carlo ◽

Neutron Flux ◽

Neutron Source ◽

Gamma Ray ◽

Neutron Radiography ◽

Image Plane ◽

Monte Carlo Code ◽

Radiography System ◽

Neutron Filter ◽

Non Destructive

Neutron radiography uses the unique interaction probabilities of neutrons to create images of materials. This imaging technique is non-destructive. MCNP Monte Carlo Code has been used to design an optimized neutron radiography system that utilizes 241Am-Be neutron source. Many different arrangements have been simulated to obtain a neutron flux with higher amplitude and more uniform distribution in the collimator outlet, next to image plane. In the final arrangement the specifications of neutron filter, Gamma-ray shield and beam collimator has been determined. Simulations has been Carried out for a 5Ci 241Am-Be neutron source. In this case 43.8 n/cm2s thermal neutron flux has been achieved at a distance of 35cm from neutron source.

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Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

Journal of Physics Conference Series ◽

10.1088/1742-6596/956/1/012010 ◽

2018 ◽

Vol 956 ◽

pp. 012010 ◽

Cited By ~ 1

Author(s):

H Basiri ◽

H Tavakoli-Anbaran

Keyword(s):

Monte Carlo ◽

Neutron Flux ◽

Thermal Neutron ◽

Neutron Source ◽

Thermal Neutron Flux ◽

Monte Carlo Code

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EFFECT OF FISSION SOURCE SPECTRUM ON MONTE CARLO CALCULATION OF EX-CORE QUANTITIES

EPJ Web of Conferences ◽

10.1051/epjconf/202124702027 ◽

2021 ◽

Vol 247 ◽

pp. 02027

Author(s):

Eva E. Davidson ◽

Tara M. Pandya ◽

Katherine E. Royston ◽

Thomas M. Evans ◽

Andrew T. Godfrey ◽

...

Keyword(s):

Monte Carlo ◽

Neutron Source ◽

Radiation Transport ◽

Source Spectrum ◽

Detector Response ◽

Monte Carlo Code ◽

Sampling Schemes ◽

Transport Calculation ◽

Water Reactors ◽

Quantities Of Interest

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a235U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a235U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for235U,238U,239Pu, and241Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to235U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.

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