Resonance Self-Shielding Treatment for Fully Ceramic Micro-Encapsulated Fuels With the Embedded Self-Shielding Method

2017 ◽  
Author(s):  
Jikui Li ◽  
Liangzhi Cao ◽  
Tiejun Zu ◽  
HongChun Wu ◽  
Qingming He
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Method Of Characteristics ◽  
Benchmark Problems ◽  
Fuel Particle ◽  
Monte Carlo Code ◽  
Carbide Matrix ◽  
New Methodologies ◽  
Coated Fuel Particle ◽  
Double Heterogeneity

Fully ceramic micro-encapsulated (FCM) fuels generate double heterogeneity (DH) challenging greatly for classical resonance self-shielding calculation method. New methodologies have been proposed and verified in this research. The target of this study is to provide homogeneous multi-group cross sections reflecting the effect of DH. Embedded Self-Shielding Method (ESSM) [1] was selected to perform resonance self-shielding calculation. Therefore, Monte Carlo code MVP [2] which is capable of well modeling the stochastic dispersed tri-structural isotropic (TRISO) coated fuel particle throughout carbide matrix and method of characteristics (MOC) were chosen to develop the heterogeneous resonance integral (RI) tables for DH problems. Benchmark problems from reference [3] were provided to verify the new methodologies. The results show that ESSM with RI tables from MVP and MOC could well address the resonance calculation for DH problems.

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 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.

scholarly journals Moving from organ dose to microdosimetry: contribution of the Monte Carlo simulations

2005 ◽  
Vol 48 (spe2) ◽  
pp. 191-199 ◽  
Author(s):  
Christophe Champion
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Dose Calculation ◽  
Organ Dose ◽  
Monte Carlo Event ◽  
Monte Carlo Code ◽  
Geometrical Modelling ◽  
Electrons And Positrons ◽  
Wave Methods ◽  
Event Simulations

When living cells are irradiated by charged particles, a wide variety of interactions occurs that leads to a deep modification of the biological material. To understand the fine structure of the microscopic distribution of the energy deposits, Monte Carlo event-by-event simulations are particularly suitable. However, the development of these track structure codes needs accurate interaction cross sections for all the electronic processes: ionization, excitation, Positronium formation (for incident positrons) and even elastic scattering. Under these conditions, we have recently developed a Monte Carlo code for electrons and positrons in water, this latter being commonly used to simulate the biological medium. All the processes are studied in detail via theoretical differential and total cross sections calculated by using partial wave methods. Comparisons with existing theoretical and experimental data show very good agreements. Moreover, this kind of detailed description allows one access to a useful microdosimetry, which can be coupled to a geometrical modelling of the target organ and then provide a detailed dose calculation at the nanometric scale.

scholarly journals Burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006

2021 ◽  
Vol 8 (4) ◽  
pp. 10-19
Author(s):  
Tiep Nguyen Huu ◽  
Dung Nguyen Thi ◽  
Phu Tran Viet ◽  
Thanh Tran Vinh ◽  
Ha Pham Nhu Viet
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Wide Temperature Range ◽  
Collision Probability ◽  
Shielding Effect ◽  
Probability Method ◽  
Monte Carlo Code ◽  
Benchmark Data ◽  
Fuel Assemblies ◽  
Burnup Calculation

The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (lowenriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006. The new depletion capability of MCNP6 was applied in the burnup calculation of the VVER-1000 LEU benchmark assembly. The OTF (on-the-fly) methodology of MCNP6, which involves high precision fitting of Doppler broadened cross sections over a wide temperature range, was utilized to handle temperature variation for heavy isotopes. The collision probability method based PIJ module of SRAC2006 was also used in this burnup calculation. The reactivity of the fuel assembly, the isotopic concentrations and the shielding effect due to the presenceof the gadolinium isotopes were determined with burnup using MCNP6 and SRAC2006 incomparison with the available published benchmark data. This study is therefore expected to reveal the capabilities of MCNP6 and SRAC2006 in burnup calculation of VVER-1000 fuel assemblies.

scholarly journals Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte Carlo code

2019 ◽  
Vol 9 (3) ◽  
pp. 21-29
Author(s):  
Cuong Nguyen Kien ◽  
Dung Nguyen Thi ◽  
Phu Tran Viet ◽  
Tiep Nguyen Huu ◽  
Ha Pham Nhu Viet
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Research Reactor ◽  
Model Development ◽  
Nuclear Research ◽  
Monte Carlo Code ◽  
Reactor Kinetics ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Comparative Results

This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code. The DNRR Serpent 2 model was verified through comparing with the MCNP6 criticality calculations under different reactor conditions. The parameters to be compared include the effective neutron multiplication factor, radial and axial powerdistributions, and thermal neutron flux distributions. The comparative results generally show a good agreement between Serpent 2 and MCNP6 and thus indicate that the DNRR Serpent 2 model can be used for further calculations of the DNRR.

Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the MCU Monte-Carlo Code

2021 ◽  
Author(s):  
Artem S. Bikeev ◽  
Yulia S. Daichenkova ◽  
Mikhail A. Kalugin ◽  
Denis Shkarovsky ◽  
Vladislav V. Shkityr
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Cross Sections ◽  
Optimal Number ◽  
Physical Characteristics ◽  
Monte Carlo Code ◽  
Special Version ◽  
The Monte Carlo Method ◽  
Scattering Anisotropy ◽  
Neutron Cross Sections

Abstract The main purpose of this work is to study the possibility of using the few-group approximation for calculation of some neutron-physical characteristics of VVER-1000 core by means of special version of MCU code. The Monte-Carlo method for VVER-1000 core neutron-physical characteristics calculation using the few-group approximation with an estimate of neutron cross sections “by location“ was provided and tested in this research. The reduction of calculation time due to the transition from a pointwise model of representation of cross sections to the few-group approximation and methodical error of this approach were evaluated. Optimal number of energy groups was determined. It was found that consideration of the scattering anisotropy leads to a significant decrease in methodical error. Ways of further reduction of methodical error were worked out.

Neutralization of Ion Beams for Reduction of Charging Damage in Plasma Etching

2005 ◽  
Vol 494 ◽  
pp. 297-302 ◽  
Author(s):  
Aleksandra Nina ◽  
M. Radmilović-Radjenović ◽  
Z.Lj. Petrović
Keyword(s):  
Monte Carlo ◽  
Gas Phase ◽  
Plasma Etching ◽  
Cross Sections ◽  
Ion Beams ◽  
Monte Carlo Code ◽  
Motion Simulation ◽  
Ion Energy ◽  
Ion Motion ◽  
Neutral Beams

Neutral beams were proposed for plasma etching in order to reduce damage due to charging. We analyzed the efficiency of neutralization in a system proposed recently where a beam of ions was neutralized in the gas phase and in a set of narrow tubes. We studied surface neutralization as well as collisions of ions in the gas. In the case of collisions of ions with aperture walls, efficiency of neutralization was assumed to be 100%. We also investigated the influence of various parameters such as aperture length, tube diameter and initial ion energy on the efficiency of neutralization using a Monte Carlo code for ion motion simulation. In our calculations we used a well-established set of cross sections for argon.

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