An efficient hybrid multi-level CMFD in space and energy for accelerating the high-fidelity neutron transport calculation

2021 ◽  
Vol 161 ◽  
pp. 108446
Author(s):  
Chen Hao ◽  
Lixun Liu ◽  
Le Kang ◽  
Yunlin Xu
Keyword(s):  
Neutron Transport ◽  
High Fidelity ◽  
Transport Calculation ◽  
Multi Level

Development of high-fidelity neutron transport code STREAM

2021 ◽  
pp. 107915
Author(s):  
Sooyoung Choi ◽  
Wonkyeong Kim ◽  
Jiwon Choe ◽  
Woonghee Lee ◽  
Hanjoo Kim ◽  
...  
Keyword(s):  
Neutron Transport ◽  
High Fidelity ◽  
Transport Code

A low noise, high fidelity cross phase modulation in multi-level atomic medium

2021 ◽  
Author(s):  
Liangwei Wang ◽  
Jia Guan ◽  
Chengjie Zhu ◽  
Runbing Li ◽  
Jing Shi
Keyword(s):  
Phase Modulation ◽  
Low Noise ◽  
High Fidelity ◽  
Cross Phase Modulation ◽  
Atomic Medium ◽  
Multi Level

A multi-level parallelization concept for high-fidelity multi-block solvers

1997 ◽  
Author(s):  
Ferhat F. Hatay ◽  
Dennis C. Jespersen ◽  
Guru P. Guruswamy ◽  
Yehia M. Rizk ◽  
Chansup Byun ◽  
...  
Keyword(s):  
High Fidelity ◽  
Multi Level

scholarly journals Validation of the code package MCPV adapted for neutron transport calculation within WWER-440 reactor near-vessel space

2019 ◽  
Vol 20 (2) ◽  
pp. 153-158
Author(s):  
O.M. Pugach ◽  
◽  
S.M. Pugach ◽  
V.L. Diemokhin ◽  
V.N. Bukanov ◽  
...  
Keyword(s):  
Neutron Transport ◽  
Transport Calculation

scholarly journals A HIGH-FIDELITY SIMULATION OF THE C5G7 BENCHMARK BY USING THE PARALLEL ENTER CODE

2021 ◽  
Vol 247 ◽  
pp. 06023
Author(s):  
Zhenglin Ruan ◽  
Haibing Guo
Keyword(s):  
Large Scale ◽  
High Efficiency ◽  
Nuclear Reactors ◽  
Neutron Transport ◽  
Reaction Cross ◽  
High Fidelity ◽  
Control Rod ◽  
Section Data ◽  
Transport Code ◽  
Deterministic Transport

In simulation of advanced nuclear reactors, requirements like high precision, high efficiency and convenient to multi-physics coupling are putting forward. The deterministic transport method has the advantage of high efficiency, capable of obtaining detailed flux distribution and efficient in multi-physics coupling, but its accuracy is limited by the homogenized reaction cross-section data and core modelling exactness. The traditional two-steps homogenization strategy may introduce substantial deviation during the assembly calculation. It is possible to conduct a whole core deterministic transport simulation pin-by-pin to achieve higher accuracy, which eliminates the assembly homogenization process. The C5G7 benchmarks were proposed to test the ability of a modern deterministic transport code in analyzing whole core reactor problems without spatial homogenization. Different deterministic code that developed by different methods were applied to the benchmark simulation and some of them solved the benchmark accurately. However, there still exist some drawbacks in the given calculation processes which carried out by some other deterministic transport codes and we could find that the fuel pin cell in the assembly were not exactly geometrically modelled owing to the limit of the code. Consequently, the calculation precision could be improved by utilizing a high-fidelity geometry modelling. In this paper, the C5G7 benchmarks with different control rod position and different configuration were calculated by the finite element SN neutron transport code ENTER [1], and the results were presented after massively parallel computation on TIANHE-II supercomputer. By introducing a large scale high-fidelity unstructured meshes, high fidelity distributions of power and neutron flux were gained and compared with the results from other codes, excellent consistency were observed. To sum up, the ENTER code can meet those new requirements in simulation of advanced nuclear reactors and more works and researches will be implemented for a further improvement.

Development of a Two-Dimensional Modularity Characteristics Code for Neutron Transport Calculation

2013 ◽  
Author(s):  
Liang Liang ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Youqi Zheng
Keyword(s):  
Large Scale ◽  
Method Of Characteristics ◽  
Neutron Transport ◽  
Engineering Application ◽  
Polar Angle ◽  
Two Dimensional ◽  
Coarse Mesh ◽  
Characteristics Method ◽  
Large Scale Problems ◽  
Transport Calculation

The method of characteristics (MOC) has been widely used in lattice code for its high precision and easy complement. However, the long characteristics method needs large quantity of PC memory when dealing with large scale problems. The modularity MOC method could significantly reduce the PC memory when calculating the problem which contains lots of repeatedly geometries, like the fuel assembly in the reactor. In this method, only typical geometric cells are selected to trace the rays, and then the geometry information of these cells is stored. So, the modularity MOC method is feasible to perform well in the calculation with large scale. When tracing the rays, the technique of mesh ray generating and the corresponding azimuthal quadrature set are both applied. The techniques make sure that each ray has the reflected ray in the boundary so it is convenient to describe the boundary condition. The optimal polar angle and the Guass quadrature set are selected as the polar quadrature set. Furthermore, the coarse mesh finite difference (CMFD) is employed to accelerate the calculation. A pin cell is chosen as the coarse mesh. The CMFD solution provides the MOC with much faster converged fission and scattering source distributions. The LOTUS code is developed and the numerical results show that the code is precise for engineering application and the CMFD acceleration is effective.

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