scholarly journals Algorithmic choices in WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

2015 ◽  
Vol 77 ◽  
pp. 176-193 ◽  
Author(s):  
Ryan M. Bergmann ◽  
Jasmina L. Vujić
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Continuous Energy

Continuous-energy Monte Carlo neutron transport on GPUs in the Shift code

2019 ◽  
Vol 128 ◽  
pp. 236-247 ◽  
Author(s):  
Steven P. Hamilton ◽  
Thomas M. Evans
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Continuous Energy

scholarly journals Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

2017 ◽  
Vol 103 ◽  
pp. 334-349 ◽  
Author(s):  
Ryan M. Bergmann ◽  
Kelly L. Rowland ◽  
Nikola Radnović ◽  
Rachel N. Slaybaugh ◽  
Jasmina L. Vujić
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Continuous Energy

Physics-oriented optimization strategy for the energy lookup algorithm in continuous energy Monte Carlo neutron transport simulation

2019 ◽  
Vol 234 ◽  
pp. 146-158
Author(s):  
Zhenping Chen ◽  
Jinsen Xie ◽  
Qian Guo ◽  
Qin Xie ◽  
Zijing Liu ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Optimization Strategy ◽  
Continuous Energy ◽  
Transport Simulation

scholarly journals DIAGNOSIS OF THE UNRESOLVED DOMAIN TREATMENT IN MONTE CARLO TRANSPORT CALCULATIONS THROUGH THE IDENTIFICATION AND MODELLING OF CRITICALITY SAFETY EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 10003
Author(s):  
N. García-Herranz ◽  
J. Rodríguez ◽  
A. Jiménez-Carrascosa ◽  
O. Cabellos
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Resonance Region ◽  
Nuclear Data ◽  
High Fidelity ◽  
Continuous Energy ◽  
Critical Facilities ◽  
Nuclear Systems ◽  
Monte Carlo Transport ◽  
Criticality Safety

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

scholarly journals A Continuous Energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the Neutron Energy Spectrum by Target Nuclei Tagging

2021 ◽  
Vol 8 (3B) ◽  
Author(s):  
Luiz Felipe Fracasso Chaves Barcellos ◽  
Bardo Ernst Josef Bodmann ◽  
Sergio De Queiroz Bogado leite ◽  
Marco Túlio Menna Barreto De Vilhena
Keyword(s):  
Monte Carlo ◽  
Energy Spectrum ◽  
Neutron Energy ◽  
Neutron Transport ◽  
Neutron Energy Spectrum ◽  
Continuous Energy ◽  
Energy Neutron ◽  
Monte Carlo Simulator

scholarly journals INVESTIGATION ON DETERMINISTIC TRUNCATION TO CONTINUOUS ENERGY MONTE CARLO NEUTRON TRANSPORT CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 04023
Author(s):  
Inhyung Kim ◽  
Yonghee Kim
Keyword(s):  
Monte Carlo ◽  
Figure Of Merit ◽  
Solution Method ◽  
Energy Transport ◽  
Neutron Transport ◽  
Computing Time ◽  
Computational Cost ◽  
Benchmark Problems ◽  
Continuous Energy ◽  
Transport Calculation

This paper presents the application and evaluation of a deterministic truncation of Monte Carlo (DTMC) solution method in a whole core reactor problem based on a continuous energy transport calculation. The DTMC method has been studied and developed as a systematic way to truncate the high-fidelity Monte Carlo (MC) solution to reduce the computational cost without compromising the essential reliability of the solution. Its fea-sibility and capability were preliminarily validated in several benchmark problems using a multi-group energy MC code. In this paper, further study has been conducted in the more practical application with the continuous-energy based MC calculation. The con-cept of the DTMC method is briefly described. Improvements to enhance the numerical stability and efficiency are specified in details. The DTMC method is applied to an SMR problem, in which reactor parameters are estimated to characterize the numerical per-formance and are compared to the standard MC method. Last, the computing time and corresponding figure-of-merit are evaluated.

A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification

2018 ◽  
Author(s):  
Wankui Yang ◽  
Baoxin Yuan ◽  
Songbao Zhang ◽  
Haibing Guo ◽  
Yaoguang Liu ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Research Reactor ◽  
Neutron Transport ◽  
Reactor Core ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Deterministic Theory ◽  
Continuous Energy ◽  
Transport Calculation ◽  
Good Agreement

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing keff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But keff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the keff difference between these two methods remains. Even though the keff results in this benchmark are less than keff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

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