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

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.

scholarly journals Optimizing the RMC Code Using the Decay Chain Method for Large-Scale Decay Calculations

2021 ◽  
Vol 9 ◽  
Author(s):  
Hao Li ◽  
Ganglin Yu ◽  
Shanfang Huang ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Large Scale ◽  
Trajectory Analysis ◽  
Linear Complexity ◽  
Neutron Transport ◽  
Decay Chain ◽  
High Fidelity ◽  
Reactor Physics ◽  
Transport Code ◽  
Full Core

Decay calculations play an important role in reactor physics simulations. In particular, the isotopic decay of the burned fuel during refueling is important for predicting the startup reactivity of the following burn-up cycle. In addition, there is a growing interest in high-fidelity simulations where the mesh in the burn-up region can involve millions of regions. However, existing models repeatedly solve the same Bateman equations for each region, which is a waste of calculational resources. RMC is a Monte Carlo neutron transport code developed for advanced reactor physics analysis including criticality calculations and burn-up calculations. This paper presents a decay calculation method named the Decay Chain Method (DCM) to optimize the RMC code for large-scale decay calculations. Unlike traditional methods, the Decay Chain Method solves the Bateman equations one decay chain at a time rather than one region at a time. The decay calculation in the burn-up mode then treats the decay steps as zero power burn-up steps with some optimized calculational methods to further reduce the calculational time. These methods were evaluated for a single pin example and for a Virtual Environment for Reactor Applications (VERA) full-core example. The calculations for the single pin example verify the accuracy of the decay step treatment in the burn-up mode and show the improved efficiency. The single pin is divided into 1–1,000,000 decay regions to study the efficiency differences between the Transmutation Trajectory Analysis (TTA) and DCM methods. Both methods have a linear complexity with respect to the number of regions but DCM costs just one-sixtieth of the TTA time. In the simplified VERA full core example, the DCM method reduces the decay calculation time to 0.32 min from 75.26 min while the accuracy remains unchanged.

Consistent Transport Transient Solvers of the High-Fidelity Transport Code PROTEUS-MOC

2020 ◽  
Vol 194 (7) ◽  
pp. 508-540
Author(s):  
Albert Hsieh ◽  
Guangchun Zhang ◽  
Won Sik Yang
Keyword(s):  
High Fidelity ◽  
Transport Code ◽  
Transient Solvers

Cosmic-ray fast-neutron flux measurements in the atmosphere at various latitudes

1968 ◽  
Vol 46 (10) ◽  
pp. S1023-S1026 ◽  
Author(s):  
S. A. Korff ◽  
R. B. Mendell ◽  
M. Merker ◽  
W. Sandie
Keyword(s):  
Solar Cycle ◽  
Neutron Transport ◽  
Full Range ◽  
Cosmic Ray ◽  
Leakage Flux ◽  
Transport Code ◽  
The World ◽  
Flux Measurements ◽  
High Elevations ◽  
The Relationship

We have extended in time our series of balloon flights, made at several latitudes between Hyderabad, India, and Ft. Churchill, Manitoba, to altitudes close to the top of the atmosphere. In these flights the neutrons generated by the cosmic radiation in the energy interval between 1 and 10 MeV are measured. The first set of measurements was made during the period of the minimum of solar activity, and the more recent flights carry the work into the start of the next solar cycle. A decrease in intensity at high elevations with the onset of the present solar cycle has been noted. Further data were also obtained on an airplane flight around the world over both poles, thus covering the full range of latitudes at two opposite longitudes. The relationship between the observed neutron spectrum and that derived by the use of a neutron transport code will be discussed. We shall also discuss other factors emerging from this analysis, including the numbers for radiocarbon production and the leakage flux.

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

2006 ◽  
Author(s):  
Tatjana Jevremovic ◽  
Mathieu Hursin ◽  
Nader Satvat ◽  
John Hopkins ◽  
Shanjie Xiao ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Three Dimensional ◽  
Open Architecture ◽  
Coarse Mesh ◽  
Research Reactors ◽  
Acceleration Techniques ◽  
Performance Accuracy ◽  
Transport Code ◽  
Neutron Propagation

The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Rebalancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined.

scholarly journals Parallelization of Neutron Transport Code ATES3 on BARC's Parallel System

2011 ◽  
Author(s):  
Kislay Bhatt ◽  
Vibhuti Duggal ◽  
Rajesh Kalmady ◽  
Anurag Gupta
Keyword(s):  
Neutron Transport ◽  
Parallel System ◽  
Transport Code

scholarly journals IMPLEMENTATION OF QUADRATIC AXIAL TRIAL FUNCTIONS IN THE HIGH-FIDELITY TRANSPORT CODE PROTEUS-MOC

2021 ◽  
Vol 247 ◽  
pp. 03015
Author(s):  
Guangchun Zhang ◽  
Won Sik Yang
Keyword(s):  
Linear Approximation ◽  
Quadratic Approximation ◽  
Linear Functions ◽  
Computational Time ◽  
High Fidelity ◽  
Test Results ◽  
Memory Requirement ◽  
Transport Code ◽  
Axial Variation ◽  
Memory Reduction

PROTEUS-MOC is a pin-resolved high-fidelity transport code, in which the axial variation of angular flux is represented in terms of orthogonal polynomials. Currently, PROTEUS-MOC employs linear functions and requires relatively fine axial meshes to achieve high accuracy, which increases the number of axial meshes and hence the memory requirement. In this study, aiming to reduce the memory requirement and potentially the computational time by allowing larger axial meshes, we have extended the PROTEUS-MOC transport solution method to quadratic trial functions. Preliminary tests for the performance of quadratic trial functions have been performed using the 3-D C5G7 benchmark problem. Test results showed that for the same axial mesh configuration with relatively large sizes, the quadratic approximation yields about 2 to 5 times more accurate pin powers than the linear approximation, depending on the degree of axial variation of angular fluxes. The quadratic approximation also allows the use of about 3 times coarser axial meshes than the linear approximation for comparable pin power accuracy, which consequently reduces the memory requirement by about 2 times. The memory reduction is not proportional because of the increased number of coefficients in each element from 2 to 3. However, the quadratic approximation did not reduce the computational time as expected because of the deteriorated performance of the pCMFD acceleration scheme due to large axial mesh sizes.

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