scholarly journals Collisions of Nucleons with Atoms: Calculated Cross Sections and Monte Carlo Simulation

2021 ◽  
Vol 9 ◽  
Author(s):  
Francesc Salvat ◽  
José Manuel Quesada
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Impact Ionization ◽  
Neutron Transport ◽  
Emission Spectra ◽  
Monte Carlo Code ◽  
Dose Distributions ◽  
Elastic Collisions ◽  
Proton Impact

After a summary description of the theory of elastic collisions of nucleons with atoms, we present the calculation of a generic database of differential and integrated cross sections for the simulation of multiple elastic collisions of protons and neutrons with kinetic energies larger than 100 keV. The relativistic plane-wave Born approximation, with binding and Coulomb-deflection corrections, has been used to calculate a database of proton-impact ionization of K-shell and L-, M-, and N-subshells of neutral atoms These databases cover the whole energy range of interest for all the elements in the periodic system, from hydrogen to einsteinium (Z = 1–99); they are provided as part of the penh distribution package. The Monte Carlo code system penh for the simulation of coupled electron-photon-proton transport is extended to account for the effect of the transport of neutrons (released in proton-induced nuclear reactions) in calculations of dose distributions from proton beams. A simplified description of neutron transport, in which neutron-induced nuclear reactions are described as a fractionally absorbing process, is shown to give simulated depth-dose distributions in good agreement with those generated by the Geant4 code. The proton-impact ionization database, combined with the description of atomic relaxation data and electron transport in penelope, allows the simulation of proton-induced x-ray emission spectra from targets with complex geometries.

scholarly journals METHODOLOGY FOR HIGH-FIDELITY DETERMINISTIC MODELLING OF SWISS LWR FUEL ASSEMBLIES

2021 ◽  
Vol 247 ◽  
pp. 06011
Author(s):  
A. Bernal ◽  
M. Pecchia ◽  
D. Rochman ◽  
A. Vasiliev ◽  
H. Ferroukhi
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Power Distribution ◽  
Neutron Transport ◽  
3D Models ◽  
Experimental Information ◽  
Monte Carlo Code ◽  
Deterministic Modelling ◽  
Fuel Assemblies ◽  
Paul Scherrer

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated against a database of experimental information. Moreover, PSI has improved the pin power calculations by developing models of Swiss fuel assemblies for the Monte Carlo code MCNP, with the isotopic compositions obtained from the In-Core Fuel Management data of the Studsvik suite of codes, by using the SNF code. A step forward is to use a neutron code based on fast deterministic neutron transport methods. The method used in this work is based on a planar Method of Characteristics in which the axial coupling is solved by 1D SP3 method. The neutron code used is nTRACER. Thus, the methodology of this work develops nTRACER models of Swiss PWR fuel assemblies, in which the fuel of each pin and axial level is modelled with the isotopic composition obtained from SNF. This methodology was applied to 2D and 3D calculations of a Swiss PWR fuel assembly. However, this method has two main limitations. First, the cross sections libraries of nTRACER lack some of the isotopes obtained by SNF. Fortunately, this work proves that the missing isotopes do not have a strong effect on keff and the power distribution. Second, the 3D models require high computational memory resources, that is, more than 260 Gb. Thus, the nTRACER code was modified, so now it uses only 8 Gb, without any loss of accuracy. Finally, the keff and power results are compared with Monte Carlo calculations obtained by Serpent.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

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.

scholarly journals Numerical simulation of emission spectra from ion beam-heated aluminum plasmas

1992 ◽  
Vol 10 (2) ◽  
pp. 349-363 ◽  
Author(s):  
J. J. MacFarlane ◽  
P. Wang
Keyword(s):  
Cross Sections ◽  
Impact Ionization ◽  
Ion Beam ◽  
Emission Spectra ◽  
Particle Beam ◽  
Rate Equations ◽  
Hartree Fock ◽  
Non Local ◽  
Temperature Diagnostic ◽  
Ionization Cross Sections

Non-local thermodynamic equilibrium radiative transfer calculations have been performed to predict emission spectra from plasma heated by intense proton beams. Multilevel, steady-state atomic rate equations were solved self-consistently with the radiation field to determine excitation and ionization populations. Ion beam effects were, included in the rate equations. Proton-impact ionization cross sections were calculated using a plane wave Born approximation model with Hartree-Fock wave functions for the electrons. We examined the dependence of emission spectra on the temperature and thickness of the plasma. In addition, Kα satellite line spectra were computed to assess its potential as a temperature diagnostic. Calculated Kα spectral results are compared with recent Particle Beam Fusion Accelerator II experimental data.

scholarly journals Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

2013 ◽  
Vol 2013 ◽  
pp. 1-9 ◽  
Author(s):  
Mario Matijević ◽  
Dubravko Pevec ◽  
Krešimir Trontl
Keyword(s):  
Cross Sections ◽  
Neutron Transport ◽  
National Laboratory ◽  
General Purpose ◽  
Monte Carlo Code ◽  
Oak Ridge ◽  
Benchmark Data ◽  
Selection Of ◽  
Good Agreement ◽  
Reactor Dosimetry

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

scholarly journals Verifying the accuracy of 3D-CRT dose distributions calculated by the Prowess Panther treatment planning system (TPS) with Monte Carlo (MC) simulation for head-and-neck (H&N) patients

2019 ◽  
Vol 3 (2) ◽  
pp. 90-99
Author(s):  
Luong Thi Oanh ◽  
Duong Thanh Tai ◽  
Hoang Duc Tuan ◽  
Truong Thi Hong Loan
Keyword(s):  
Monte Carlo ◽  
Head And Neck ◽  
Treatment Planning ◽  
Planning System ◽  
Treatment Planning System ◽  
Average Dose ◽  
Monte Carlo Code ◽  
Dose Distributions ◽  
Mc Simulation ◽  
3D Crt

The purpose of this study is to verify and compare the three Dimensional Conformal Radiation Therapy (3D-CRT) dose distributions calculated by the Prowess Panther treatment planning system (TPS) with Monte Carlo (MC) simulation for head-and-neck (H&N) patients. In this study, we used the EGSnrc Monte Carlo code which includes BEAMnrc and DOSXYZnrc programs. Firstly, the clinical 6 MV photon beams form Siemens Primus linear accelerator at Dong Nai General Hospital were simulated using the BEAMnrc. Secondly, the absorbed dose to patients treated by 3D-CRT was computed using the DOSXYZnrc. Finally, the simulated dose distributions were then compared with the ones calculated by the Fast Photon Effective algorithm on the TPS, using the relative dose error comparison and the gamma index using global methods implemented in PTW-VeriSoft with 3%/3 mm. There is a good agreement between the MC and TPS dose. The average gamma passing rates were 92.8% based on the 3%/3 mm. The average dose in the PTV agreed well between the TPS with 0.97% error. MC predict dose was higher than the mean dose to the parotid glands and spinal cord compared to TPS. We have implemented the EGSnrc-based Monte Carlo simulation to verify the 3D-CRT plans generated by Prowess Panther TPS. Our results showed that the TPS agreed with the one of MC.  

scholarly journals Processing and application of nuclear data for low temperature criticality assessment

2020 ◽  
Vol 239 ◽  
pp. 14006
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Tara Hanlon ◽  
Richard Hiles ◽  
Malcolm Lingard ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Low Temperature ◽  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Monte Carlo Code ◽  
Temperature Range ◽  
Nuclear Data Library ◽  
Criticality Assessment ◽  
Good Agreement

Until recently, criticality safety assessment codes had a minimum temperature at which calculations can be performed. Where criticality assessment has been required for lower temperatures, indirect methods, including reasoned argument or extrapolation, have been required to assess reactivity changes associated with these temperatures. The ANSWERS Software Service MONK® version 10B Monte Carlo criticality code, is capable of performing criticality calculations at any temperature, within the temperature limits of the underlying nuclear data in the BINGO continuous energy library. The temperature range of the nuclear data has been extended below the traditional lower limit of 293.6 K to 193 K in a prototype BINGO library, primarily based on JEFF-3.1.2 data. The temperature range of the thermal bound scattering data of the key moderator materials was extended by reprocessing the NJOY LEAPR inputs used to produce bound data for JEFF-3.1.2 and ENDF/B-VIII.0. To give confidence in the low temperature nuclear data, a series of MONK and MCBEND calculations have been performed and results compared against external data sources. MCBEND is a Monte Carlo code for shielding and dosimetry and shares commonalities to its sister code MONK including the BINGO nuclear data library. Good agreement has been achieved between calculated and experimental cross sections for ice, k-effective results for low temperature criticality benchmarks and calculated and experimentally determined eigenvalues for thermal neutron diffusion in ice. To quantify the differences between ice and water bound scattering data a number of MONK criticality calculations were performed for nuclear fuel transport flask configurations. The results obtained demonstrate good agreement with extrapolation methods. There is a discernible difference in the use of ice and water data.

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