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.

STOPPING CROSS SECTIONS IN CARBON FOR LOW-ENERGY ATOMS WITH

1963 ◽  
Vol 41 (9) ◽  
pp. 1424-1442 ◽  
Author(s):  
J. H. Ormrod ◽  
H. E. Duckworth
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Atomic Number ◽  
Cross Sections ◽  
Monte Carlo Calculation ◽  
Low Energy ◽  
Energy Interval ◽  
Nuclear Scattering

The electronic stopping cross sections in carbon for atomic projectiles with [Formula: see text] have been determined in the energy interval from 10 to 140 kev. In doing so a Monte Carlo calculation was used to subtract from each experimentally observed cross section the contribution which arises from nuclear scattering. The trend of the results thus obtained agrees well with theory. In addition, however, a periodic dependence of Sε on the atomic number of the projectile is observed.

scholarly journals IMPROVEMENT OF SCALE-XSProc MULTIGROUP CROSS SECTION PROCESSING BASED ON THE CENTRM POINTWISE SLOWING DOWN CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 02011
Author(s):  
Seog Kim Kang ◽  
Andrew M. Holcomb ◽  
Friederike Bostelmann ◽  
Dorothea Wiarda ◽  
William Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Reaction Rate ◽  
Neutron Transport ◽  
Energy Structure ◽  
Reactor Physics ◽  
Slowing Down ◽  
Monte Carlo Calculations ◽  
Scattering Matrices

The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations. This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.

Neutronic simulation of China experimental fast reactor start-up tests- part II: MCS code Monte Carlo calculation

2020 ◽  
Vol 148 ◽  
pp. 107710
Author(s):  
Tuan Quoc Tran ◽  
Jiwon Choe ◽  
Xianan Du ◽  
Hyunsuk Lee ◽  
Deokjung Lee
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Monte Carlo Calculation ◽  
Start Up ◽  
China Experimental Fast Reactor

scholarly journals Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction

2014 ◽  
Vol 2014 ◽  
pp. 1-18 ◽  
Author(s):  
G. G. Kulikov ◽  
A. N. Shmelev ◽  
V. A. Apse
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Reactor Core ◽  
Large Contribution ◽  
Delayed Neutrons ◽  
Slowing Down ◽  
Radiogenic Lead ◽  
Radiative Neutron ◽  
Long Lifetime ◽  
Radiative Neutron Capture

Light materials with small atomic mass (light or heavy water, graphite, and so on) are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb) for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable of transforming some part of prompt neutrons leaked from the core into the reflected neutrons with properties similar to those of delayed neutrons, that is, sufficiently large contribution to reactivity at the level of effective fraction of delayed neutrons and relatively long lifetime, comparable with lifetimes of radionuclides-emitters of delayed neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in the fast reactor. This can improve the fast reactor safety and reduce some requirements to the technologies used to fabricate fuel for the fast reactor.

MONTE CARLO CALCULATION OF THE ENERGY DISTRIBUTION OF BACKSCATTERED ELECTRONS

2002 ◽  
Vol 16 (28n29) ◽  
pp. 4405-4412 ◽  
Author(s):  
Z. J. DING ◽  
X. D. TANG ◽  
H. M. LI
Keyword(s):  
Monte Carlo ◽  
Elastic Scattering ◽  
Energy Distribution ◽  
Cross Section ◽  
Monte Carlo Calculation ◽  
Simulation Method ◽  
Elastic Peak ◽  
Secondary Electrons ◽  
Backscattered Electrons ◽  
Cylindrical Mirror

The full energy distribution of backscattered electrons from elastic peak down to true secondary electron peak has been calculated by a Monte-Carlo simulation method by including cascade secondary electrons production. The simulation method is based on the use of a dielectric function for describing electron inelastic scattering and secondary excitation, and the use of Mott cross section for electron elastic scattering. This calculation reproduces well the backscattering background observed in the direct mode of AES. The calculated absolute electron yields have been compared with the available experimental data. The simulation has indicated that, due to the effect of the elastic scattering differential cross section and detection solid angles, the shape of the energy distribution measured with a cylindrical mirror analyzer may differ from the overall energy spectrum of emitted electrons.

Evaluation Technique With MCNP for Temperature Dependent Problems

2010 ◽  
Author(s):  
Hiroyuki Fujime ◽  
Shinji Abe ◽  
Kazuya Yamaji ◽  
Daisuke Sato ◽  
Hideki Matsumoto
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Monte Carlo Calculation ◽  
Design Codes ◽  
Design Code ◽  
Temperature Dependent ◽  
Validation And Verification ◽  
The World ◽  
Profile Changes ◽  
Complex Temperature

Monte Carlo calculation has come to be used as reference solutions instead of experiments in nuclear design code validation and verification (V&V), although comparisons with measurements are still indispensable for V&V in nuclear design. MCNP [1] is one of the most famous Monte Carlo codes widely used in the world. Many reference results are given for the analyses of critical experiments. When using the use MCNP calculations for validations of commercial design codes, we will face to a problem of lacking temperature dependent cross-sections. The cross-sections can be generated by the NJOY code [2]. However, if the model has complex temperature distribution, many NJOY calculations are necessary. Besides, if the temperature profile changes with fuel power and so on, many NJOY calculations have to be performed again and again. These back and forth procedures make us give up using MCNP for commercial LWR calculations. In order to solve this problem, we propose an easy approximation to solve the temperature problems using MCNP. Note that our technique does not require any code modifications.

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