Detail publikačního výsledku

Monte Carlo analysis of HDPE using PHITS and MCNP for neutron shielding applications

UNAGAR, V.; MAKWANA, R.; BARALA, S.; MEENA, D.; GUPTA, S.; KAVUN, Y.; MEHTA, M.; VASHI, V.; SINGH, R.; CHAUHAN, R.; MUKHERJEE, S.; SINGH, N.; KATOVSKÝ, K.

Originální název

Monte Carlo analysis of HDPE using PHITS and MCNP for neutron shielding applications

Anglický název

Monte Carlo analysis of HDPE using PHITS and MCNP for neutron shielding applications

Druh

Článek WoS

Originální abstrakt

The nuclear radiation shielding materials are having a great importance in reactor shielding, nuclear laboratories, radiation sources storage, as well as in space radiation shielding. The experimental methods are time consuming, facility dependent, as well as costly affair in order to manufacture the shielding materials. The alternative approach is to use nuclear transport codes such as GEANT, MCNP, PHITS, FLUKA for nuclear radiation shielding application. The present work is planned to study the prediction capabilities of transport code for calculation of shielding parameters using two different radiation transport codes and comparison with experimental results. The simulation measurements were carried out by Monte Carlo analysis, 241Am-Be neutron source & neutron dosimeter, and High Density Polyethylene (HDPE) were modeled using MCNP 6 and PHITS 3.22 codes. Subsequently, experimental neutron dose rates were measured after different thickness of HDPE samples using live 241Am-Be neutron source and NRF31 dosimeter. The study shows a good agreement between the experimental and simulated results which suggest that the simulation method can be used for the optimization of new shielding composites.

Anglický abstrakt

The nuclear radiation shielding materials are having a great importance in reactor shielding, nuclear laboratories, radiation sources storage, as well as in space radiation shielding. The experimental methods are time consuming, facility dependent, as well as costly affair in order to manufacture the shielding materials. The alternative approach is to use nuclear transport codes such as GEANT, MCNP, PHITS, FLUKA for nuclear radiation shielding application. The present work is planned to study the prediction capabilities of transport code for calculation of shielding parameters using two different radiation transport codes and comparison with experimental results. The simulation measurements were carried out by Monte Carlo analysis, 241Am-Be neutron source & neutron dosimeter, and High Density Polyethylene (HDPE) were modeled using MCNP 6 and PHITS 3.22 codes. Subsequently, experimental neutron dose rates were measured after different thickness of HDPE samples using live 241Am-Be neutron source and NRF31 dosimeter. The study shows a good agreement between the experimental and simulated results which suggest that the simulation method can be used for the optimization of new shielding composites.

Klíčová slova

Neutron dose; Neutron shielding; PHITS; MCNP; Dose rate

Klíčová slova v angličtině

Neutron dose; Neutron shielding; PHITS; MCNP; Dose rate

Autoři

UNAGAR, V.; MAKWANA, R.; BARALA, S.; MEENA, D.; GUPTA, S.; KAVUN, Y.; MEHTA, M.; VASHI, V.; SINGH, R.; CHAUHAN, R.; MUKHERJEE, S.; SINGH, N.; KATOVSKÝ, K.

Rok RIV

2026

Vydáno

01.11.2024

Nakladatel

Springer Nature

Periodikum

Journal of Radioanalytical and Nuclear Chemistry

Svazek

333

Číslo

11

Stát

Maďarsko

Strany od

5457

Strany do

5464

Strany počet

8

URL

https://link.springer.com/article/10.1007/s10967-024-09577-5

BibTex

@article{BUT201555,
  author="{} and  {} and  {} and  {} and  {} and  {} and  {} and  {} and  {} and  {} and  {} and  {} and Karel {Katovský}",
  title="Monte Carlo analysis of HDPE using PHITS and MCNP for neutron shielding applications",
  journal="Journal of Radioanalytical and Nuclear Chemistry",
  year="2024",
  volume="333",
  number="11",
  pages="5457--5464",
  doi="10.1007/s10967-024-09577-5",
  issn="0236-5731",
  url="https://link.springer.com/article/10.1007/s10967-024-09577-5"
}