scholarly journals A MONTE CARLO CODE FOR RELATIVISTIC RADIATION TRANSPORT AROUND KERR BLACK HOLES

2013 ◽  
Vol 777 (1) ◽  
pp. 11 ◽  
Author(s):  
Jeremy D. Schnittman ◽  
Julian H. Krolik
Keyword(s):  
Black Holes ◽  
Monte Carlo ◽  
Radiation Transport ◽  
Monte Carlo Code ◽  
Kerr Black Holes

Cattle’s Thyroid Dose Estimation with Compartmental Model of Iodine Metabolism and Monte Carlo Transport Technique

2018 ◽  
pp. 48-54
Author(s):  
Ю. Кураченко ◽  
Yu. Kurachenko ◽  
Н. Санжарова ◽  
N. Sanzharova ◽  
Г. Козьмин ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Thyroid Gland ◽  
Compartmental Model ◽  
Radiation Transport ◽  
Absorbed Dose ◽  
Dose Calculation ◽  
Average Dose ◽  
Monte Carlo Code ◽  
Thyroid Dose ◽  
The Subject

Purpose: This work aims first to improve the reliability of absorbed dose calculation in critical organs of cattle during internal irradiation immediately after radiation accidents by a) improving the compartmental model of radionuclide metabolism in animal body; b) the use of precision computing technologies for modeling as the domain, and the actual radiation transport. In addition, the aim of the work is to determine the agreed values of the 131I critical dose in the cattle thyroid, leading to serious gland dysfunction and its follow-up destruction. Material and methods: To achieve aforecited goals, comprehensive studies were carried out to specify the parameters of the compartmental model, based on reliable experimental and theoretical data. Voxel technologies were applied for modeling the subject domain (thyroid gland and its environment). Finally, to solve the 131I radiation transport equation, the Monte Carlo code was applied, which takes into account the contribution of gamma and beta radiation source, and the contribution of the entire chain of secondary radiations in the dose calculation, up to the total energy dissipation. Results: As the main theoretical result, it is necessary to emphasize the conversion factor from the 131I activity, distributed uniformly in volume of the thyroid gland, to the average dose rate in the gland (Bq × Gy/s). This factor was calculated for both cows and calves in the selected domain configuration and thyroid morphology. The main practical result is a reliable estimation the lower bound of the absorbed dose in the thyroid, which in a short time leads to its destruction under internal 131I irradiation: ~300 Gy. Conclusion: Usage a compartmental model of the 131I metabolism with biokinetic parameters, received on the basis of reliable experimental data, and precise models of both the subject area and radiation transport for evaluation the dose in the cattle thyroid after the radiation accident allowed to obtain reliable values of the thyroid dose, adducting to its destruction at short notice.

Organ Dose Calculations Using ICRP Adult Voxel Phantoms and TRIPOLI-4 Monte Carlo Code

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Yi-Kang Lee
Keyword(s):  
Monte Carlo ◽  
Computational Models ◽  
Radiation Transport ◽  
Organ Dose ◽  
Radiation Shielding ◽  
Radiological Protection ◽  
Monte Carlo Code ◽  
Coupled Transport ◽  
Nuclear Data Library ◽  
Voxel Phantoms

Abstract The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP (International Commission on Radiological Protection) Reference Male and Reference Female. In 2018, the Working Group 6 (WG6) of European Radiation Dosimetry Group (EURADOS) organized a study on the usage of the ICRP voxel reference phantoms. Organ dose calculation tasks with radiation transport codes were proposed in occupational, environmental, and medical dosimetry. The TRIPOLI-4 Monte Carlo radiation transport code has been widely used in radiation shielding, criticality safety, and reactor physics fields for supporting French nuclear energy research and industrial applications. To enhance the application fields of TRIPOLI-4, the 2018 EURADOS-WG6 tasks are being taken into account by using different features of the TRIPOLI-4 code. In this work, the ICRP reference voxel phantoms were first adapted into TRIPOLI-4. More than 14 × 106 voxels were represented in a mixed lattice geometry including 140 organs-tissues and 52 tissue media. Diverse exposure scenarios were then investigated by using 60Co and 241Am gamma-ray sources, 16N beta source, and 10 keV neutron source. The TRIPOLI-4 standard nuclear data library was utilized on these neutron, photon, electron, and positron-coupled transport calculations. Energy deposition estimators for electron, positron, neutron, and photon coupled with mesh tally options were used to calculate the organ absorbed dose DT and the effective dose E. TRIPOLI-4 calculation methods and primary results for the EURADOS-WG6 voxel phantom exercise on organ dose study tasks are reported here.

scholarly journals BENCHMARKING OF THE SERPENT 2 MONTE CARLO CODE FOR FUSION NEUTRONICS APPLICATIONS

2021 ◽  
Vol 247 ◽  
pp. 04015
Author(s):  
A. Valentine ◽  
B. Colling ◽  
R. Worrall ◽  
J. Leppänen
Keyword(s):  
Monte Carlo ◽  
Radiation Transport ◽  
Nuclear Fission ◽  
Monte Carlo Code ◽  
Tritium Production ◽  
Radiation Fields ◽  
Remote Maintenance ◽  
Depth Analysis ◽  
Fusion Applications ◽  
Maintenance Requirements

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

Radiation transport calculations for 50 MV photon therapy beam using the Monte Carlo code GEANT4

2005 ◽  
Vol 115 (1-4) ◽  
pp. 503-507 ◽  
Author(s):  
Susanne Larsson ◽  
Roger Svensson ◽  
Irena Gudowska ◽  
Vladimir Ivanchenko ◽  
Anders Brahme
Keyword(s):  
Monte Carlo ◽  
Radiation Transport ◽  
Monte Carlo Code

scholarly journals Preliminary Validation of the Shift Monte Carlo Code for Fixed-Source Radiation Transport Problems

2019 ◽  
Vol 206 (1) ◽  
pp. 107-125
Author(s):  
Douglas E. Peplow ◽  
Kaushik Banerjee ◽  
Gregory G. Davidson ◽  
Ian R. Stewart ◽  
Mathew W. Swinney ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Radiation Transport ◽  
Monte Carlo Code ◽  
Source Radiation ◽  
Preliminary Validation ◽  
Transport Problems

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

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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