scholarly journals MC++: a parallel, portable, Monte Carlo neutron transport code in C++

2002 ◽  
Author(s):  
S.R. Lee ◽  
J.C. Cummings ◽  
S.D. Nolen
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Transport Code

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 Development of an MCNP5-ORIGEN2 coupling scheme for burnup calculation of VVER-1000 fuel assemblies

2016 ◽  
Vol 6 (3) ◽  
pp. 16-30
Author(s):  
Huy Hiep Nguyen ◽  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Shielding Effect ◽  
Coupling Scheme ◽  
Fuel Rods ◽  
Transport Code ◽  
Fuel Assemblies ◽  
Calculation Code ◽  
Burnup Calculation ◽  
Good Agreement

The paper aims to develop an MCNP5-ORIGEN2 coupling scheme for burnup calculation. Specifically, the Monte Carlo neutron transport code (MCNP5) and the nuclides depletion and decay calculation code (ORIGEN2) are combined by data processing and linking files written in the PERL programming language. The validity and applicability of the developed coupling scheme are tested through predicting the neutronic and isotopic behavior of the “VVER-1000 LEU Assembly Computational Benchmark”. The MCNP5-ORIGEN2 coupling results showed a good agreement with the k-inf benchmark values within 600 pcm during the entire burnup history. In addition, the differences of isotopes concentration at the end of the burnup (40 MWd/kgHM) when compared with benchmark values were reasonable and generally within 6.5%. The developed coupling scheme also considered the shielding effect due to gadolinium isotopes and simulated well the depletion of isotopes as a function of the radial position in gadolinium bearing fuel rods.

scholarly journals Analysis of DD, TT and DT Neutron Streaming Experiments with the ADVANTG Code

2020 ◽  
Vol 225 ◽  
pp. 02003
Author(s):  
Bor Kos ◽  
Theodora Vasilopoulou ◽  
Scott W. Mosher ◽  
Ivan A. Kodeli ◽  
Robert E. Grove ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Importance Sampling ◽  
Variance Reduction ◽  
Neutron Transport ◽  
Plasma Source ◽  
Sampling Methodology ◽  
Transport Code ◽  
Weight Window ◽  
Monte Carlo Transport ◽  
Hybrid Methodology

The paper presents an analysis of DD, TT and DT neutron streaming benchmark experiments with the recently released hybrid transport code ADVANTG (AutomateD VAriaNce reducTion Generator). ADVANTG combines the deterministic neutron transport solver Denovo with the Monte Carlo transport code MCNP via the principle of variance reduction. It automatically produces weight-window and source biasing variance reduction parameters based on the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using this novel hybrid methodology Monte Carlo simulations of realistic complex fusion streaming geometries have become possible. In this paper the experimental results from the 2016 DD campaign using measurements with TLDs and activation foils up to 40 m from the plasma source are analyzed. New detailed models of the detector assemblies were incorporated into the JET 360° MCNP model for this analysis. In preparation of the TT and DTE2 campaigns at JET a pre-analysis for these campaigns is also presented.

scholarly journals Development of One-way-coupling Methodology between Severe Accident Integral Code MELCOR and Monte Carlo Neutron Transport Code SERPENT

2016 ◽  
Vol 157 ◽  
pp. 207-213
Author(s):  
Piotr Darnowski ◽  
Kacper Potapczyk ◽  
Michał Gatkowski ◽  
Grzegorz Niewiński
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Severe Accident ◽  
Transport Code

Development of the Assembly-Level Monte Carlo Neutron Transport Code M3C for Reactor Physics Calculations

2019 ◽  
Vol 194 (1) ◽  
pp. 32-43
Author(s):  
Anek Kumar ◽  
Umasankari Kannan ◽  
S. Ganesan
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Physics ◽  
Transport Code

scholarly journals Multi-core performance studies of a Monte Carlo neutron transport code

2013 ◽  
Vol 28 (1) ◽  
pp. 87-96 ◽  
Author(s):  
Andrew R Siegel ◽  
Kord Smith ◽  
Paul K Romano ◽  
Benoit Forget ◽  
Kyle G Felker
Keyword(s):  
Monte Carlo ◽  
Performance Studies ◽  
Neutron Transport ◽  
Transport Code

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.

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 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
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