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.

scholarly journals Experimental and Simulated Spectral Gamma-Ray Response of a NaI(Tl) Scintillation Detector used in Airborne Gamma-Ray Spectrometry

2021 ◽  
Author(s):  
David Breitenmoser
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
Radiation Transport ◽  
Natural Uranium ◽  
Detector Response ◽  
Gamma Ray Spectrometry ◽  
Response Analysis ◽  
Maximum Relative Error ◽  
Potassium Sources ◽  
Significant Difference

<p>The objective of this work is to simulate the spectral gamma-ray response of NaI(Tl) scintillation detectors for airborne gamma-ray spectrometry (AGRS) using Monte Carlo radiation transport codes. The study is based on a commercial airborne gamma-ray spectrometry detector system with four individual NaI(Tl) scintillation crystals and a total volume of 16.8 l. Monte Carlo source-detector simulations were performed in an event-by-event mode with the commercial multi-purpose transport codes MCNP6.2 and FLUKA. Validation measurements were conducted using <sup>241</sup>Am, <sup>133</sup>Ba, <sup>60</sup>Co, <sup>137</sup>Cs and <sup>152</sup>Eu radiation sources with known activities and source-detector geometries. Energy resolution functions were derived from these measurements combined with additional measurements of natural Uranium, Thorium and Potassium sources. The simulation results are in good agreement with the experimental data with a maximum relative error in the full-energy peak counts of 10%. In addition, no significant difference between the two Monte Carlo radiation transport codes was found with respect to a 95% confidence level. The validated detector model presented herein can be adopted for angular detector response analysis and calibration computations relating radionuclide activity concentrations with spectral detector counts.</p>

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

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

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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