scholarly journals MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS

2021 ◽  
Vol 247 ◽  
pp. 02007
Author(s):  
Tung Dong Cao Nguyen ◽  
Hyunsuk Lee ◽  
Xianan Du ◽  
Vutheam Dos ◽  
Tuan Quoc Tran ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Fast Reactor ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Metallic Fuel ◽  
Sodium Fast Reactor ◽  
The Difference ◽  
Nodal Diffusion ◽  
Core Problem

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

The Mean Projected Range and Range Straggling of Nd Ions Implanted in Silicon Carbide

2011 ◽  
Vol 474-476 ◽  
pp. 565-569
Author(s):  
Xi Feng Qin ◽  
Shuang Li ◽  
Feng Xiang Wang ◽  
Yi Liang
Keyword(s):  
Silicon Carbide ◽  
Monte Carlo ◽  
Spectrum Analysis ◽  
Monte Carlo Code ◽  
Experiment Data ◽  
Lateral Deviation ◽  
Projected Range ◽  
The Mean ◽  
The Difference ◽  
Good Agreement

In view of the influence of the projected range, the range straggling, and the lateral deviation of ions in materials on the property of device in the fabrication of photoelectric integration devices by ion implantation, the mean projected ranges and range straggling for energetic 200 – 500 keV Nd ions implanted in 6H-SiC were measured by means of Rutherford backscattering followed by spectrum analysis. The measured values are compared with Monte Carlo code (SRIM2006) calculations. It has been found that the measured values of the mean projected range Rp are good agreement with the SRIM calculated values; for the range straggling △Rp, the difference between the experiment data and the calculated results is much higher than that of Rp

The Generation of Few-Group Constants for Fast Reactor Analysis

2016 ◽  
Author(s):  
Xianan Du ◽  
Liangzhi Cao ◽  
Youqi Zheng
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Fast Reactor ◽  
Monte Carlo Calculation ◽  
Anisotropic Scattering ◽  
Slowing Down ◽  
Nodal Method ◽  
Reactor Calculation ◽  
Broad Energy

A way to generate the few-group cross sections for fast reactor calculation is presented in this paper. It is based on the three steps computational scheme. In the first step, the ultrafine method is used to solve the slowing down equation based on the ultrafine group cross section generated by NJOY. Optional 0D or 1D calculation is used to collapse energy group into broad energy groups. In the second step, the 2D RZ calculation using SN method is performed to obtain the space dependent neutron spectra to collapse broad energy groups into few groups. The anisotropic scattering is well handled by the direct SN calculation. Finally, the full core calculation is performed by using the 3D SN nodal method. The results are compared with continuous energy Monte-Carlo calculation. Both the cross section generated in the first step and the final keff in the last step are compared. The results match well between the three steps calculation and Monte-Carlo calculation.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

scholarly journals ANALYSIS OF THE OECD/NEA SFR BENCHMARK WITH ANTS REDUCED-ORDER NODAL DIFFUSION SOLVER AND THE SERPENT MONTE CARLO CODE

2021 ◽  
Vol 247 ◽  
pp. 04021
Author(s):  
Marton Szogradi
Keyword(s):  
Monte Carlo ◽  
Lattice Models ◽  
Working Party ◽  
Research Centre ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Reduced Order ◽  
State Present ◽  
Nodal Diffusion ◽  
Full Core

In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland. The neutronic solver of the framework consists of two calculation chains, providing full core solutions by the Serpent high fidelity code (1) and the AFEN/FENM-based reduced-order diffusion solver called Ants (2) capable of handling square and hexagonal geometries in steady-state. Present work introduces the simulation of a large 3600 MWth Sodium-cooled Fast Reactor (SFR) described within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of OECD. Full-core 3D results were obtained by Serpent for carbide- and oxide-fuel cores, moreover group constants were generated for Ants utilizing 2D super-cell and single assembly infinite lattice models of Serpent. The continuous-energy Monte Carlo method provided the reference results for the verification of the reduced-order method. Implementing the spatially homogenized properties, 3D solutions were obtained by Ants as well for both core configurations. Comparison was made between the various core designs and codes based on reactivity feedbacks (Doppler constant, sodium voiding, control rod worth) considering power distributions. Regarding reactivity sensitivity on geometry, axial fuel- and radial core expansion coefficients were evaluated as well.

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.

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