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

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.

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

Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy ◽

10.1115/icone14-89561 ◽

2006 ◽

Cited By ~ 3

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.

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Flux extrapolation models used in the DOT IV discrete ordinates neutron transport code

10.2172/5361239 ◽

1980 ◽

Cited By ~ 2

Author(s):

E. T. Tomlinson ◽

W. A. Rhoades ◽

W. W. Engle, Jr.

Keyword(s):

Neutron Transport ◽

Discrete Ordinates ◽

Transport Code

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

EPJ Web of Conferences ◽

10.1051/epjconf/202124706023 ◽

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.

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Deterministic Phonon Transport Predictions of Thermal Conductivity in Uranium Dioxide With Xenon Impurities

Journal of Heat Transfer ◽

10.1115/1.4038554 ◽

2018 ◽

Vol 140 (5) ◽

Cited By ~ 3

Author(s):

Jackson R. Harter ◽

Laura de Sousa Oliveira ◽

Agnieszka Truszkowska ◽

Todd S. Palmer ◽

P. Alex Greaney

Keyword(s):

Thermal Conductivity ◽

Uranium Dioxide ◽

Neutron Transport ◽

Mean Free Path ◽

Phonon Transport ◽

Multiscale Approach ◽

First Principles Calculations ◽

Boltzmann Transport ◽

Experimental Heat ◽

Transport Code

We present a method for solving the Boltzmann transport equation (BTE) for phonons by modifying the neutron transport code Rattlesnake which provides a numerically efficient method for solving the BTE in its self-adjoint angular flux (SAAF) form. Using this approach, we have computed the reduction in thermal conductivity of uranium dioxide (UO2) due to the presence of a nanoscale xenon bubble across a range of temperatures. For these simulations, the values of group velocity and phonon mean free path in the UO2 were determined from a combination of experimental heat conduction data and first principles calculations. The same properties for the Xe under the high pressure conditions in the nanoscale bubble were computed using classical molecular dynamics (MD). We compare our approach to the other modern phonon transport calculations, and discuss the benefits of this multiscale approach for thermal conductivity in nuclear fuels under irradiation.

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ARRC: A random ray neutron transport code for nuclear reactor simulation

Annals of Nuclear Energy ◽

10.1016/j.anucene.2017.10.015 ◽

2018 ◽

Vol 112 ◽

pp. 693-714 ◽

Cited By ~ 4

Author(s):

John R. Tramm ◽

Kord S. Smith ◽

Benoit Forget ◽

Andrew R. Siegel

Keyword(s):

Nuclear Reactor ◽

Neutron Transport ◽

Transport Code

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

Nuclear Science and Technology ◽

10.53747/jnst.v6i3.160 ◽

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.

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