scholarly journals The Effect of the Flux Separability Approximation on Multigroup Neutron Transport

2021 ◽  
Vol 2 (1) ◽  
pp. 86-96
Author(s):  
Adam G. Nelson ◽  
William Boyd ◽  
Paul K. Romano
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Neutron Transport ◽  
Group Structure ◽  
Fine Tuning ◽  
Computational Time ◽  
Discrete Ordinates ◽  
Energy Structure ◽  
The Impact ◽  
Monte Carlo Transport

The angular dependence of flux-weighted multigroup cross sections is commonly neglected when generating multigroup libraries. The error of this flux separability approximation is typically not isolated from other error sources due to a lack of availability of library generation and corresponding solvers that cannot relax this approximation. These errors can now be isolated and quantified with the availability of a multigroup Monte Carlo transport and multigroup library-generation capability in the OpenMC Monte Carlo transport code. This work will discuss relevant details of the OpenMC implementation, provide an example case useful for detailing the type of errors one can expect from making the flux separability approximation, and end with more realistic problems which show the impact of the approximation and highlight how it can strongly arise from an energy-dependent resonance absorption effect. Since the angle-dependence is intrinsically linked to the energy group structure, these examples also show that relaxing the flux separability approximation with angle-dependent cross sections could be used to reduce either the fine-tuning required to set a multigroup energy structure for a specific reactor type or the number of energy groups required to obtain a desired level of accuracy for a given problem. This trade-off could increase the costs of generating multigroup cross sections, and has the potential to require more memory for storing the multigroup library during the transport calculations, but it can significantly reduce the computational time required since the runtime of a discrete ordinates or method of characteristics neutron transport solver scales roughly linearly with the number of groups.

Analysis of Major Group Structures Used for Nuclear Reactor Simulations

2018 ◽  
Author(s):  
Marco Di Filippo ◽  
Jiri Krepel ◽  
Konstantin Mikityuk ◽  
Horst-Michael Prasser
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Group Structure ◽  
Major Group ◽  
Computational Time ◽  
Sensitivity Threshold ◽  
Energy Structure ◽  
Reactor Spectrum ◽  
Group Structures ◽  
The Impact

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation. The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated. The aim was to identify a group structure with very low dependency on the actual reactor spectrum.

scholarly journals GPU ACCELERATION OF DOPPLER BROADENING FOR NEUTRON TRANSPORT CALCULATIONS1

2021 ◽  
Vol 247 ◽  
pp. 04017
Author(s):  
Paul E. Burke ◽  
Kyle E. Remley ◽  
David P. Griesheimer
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Radiation Transport ◽  
Neutron Transport ◽  
Processing Unit ◽  
Complex Data ◽  
Doppler Broadening ◽  
Central Processing ◽  
Monte Carlo Transport

In radiation transport calculations, the effects of material temperature on neutron/nucleus interactions must be taken into account through Doppler broadening adjustments to the microscopic cross section data. Historically, Monte Carlo transport simulations have accounted for this temperature dependence by interpolating among precalculated Doppler broadened cross sections at a variety of temperatures. More recently, there has been much interest in on-the-fly Doppler broadening methods, where reference data is broadened on-demand during particle transport to any temperature. Unfortunately, Doppler broadening operations are expensive on traditional central processing unit (CPU) architectures, making on-the-fly Doppler broadening unaffordable without approximations or complex data preprocessing. This work considers the use of graphics processing unit (GPU)s, which excel at parallel data processing, for on-the-fly Doppler broadening in continuous-energy Monte Carlo simulations. Two methods are considered for the broadening operations – a GPU implementation of the standard SIGMA1 algorithm and a novel vectorized algorithm that leverages the convolution properties of the broadening operation in an attempt to expose additional parallelism. Numerical results demonstrate that similar cross section lookup throughput is obtained for on-the-fly broadening on a GPU as cross section lookup throughput with precomputed data on a CPU, implying that offloading Doppler broadening operations to a GPU may enable on-the-fly temperature treatment of cross sections without a noticeable reduction in cross section processing performance in Monte Carlo transport codes.

scholarly journals IMPROVEMENT OF SCALE-XSProc MULTIGROUP CROSS SECTION PROCESSING BASED ON THE CENTRM POINTWISE SLOWING DOWN CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 02011
Author(s):  
Seog Kim Kang ◽  
Andrew M. Holcomb ◽  
Friederike Bostelmann ◽  
Dorothea Wiarda ◽  
William Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Section ◽  
Cross Sections ◽  
Reaction Rate ◽  
Neutron Transport ◽  
Energy Structure ◽  
Reactor Physics ◽  
Slowing Down ◽  
Monte Carlo Calculations ◽  
Scattering Matrices

The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations. This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.

Monte Carlo Neutron Transport on GPUs

2014 ◽  
Author(s):  
Ryan M. Bergmann ◽  
Jasmina L. Vujić
Keyword(s):  
Monte Carlo ◽  
Lower Energy ◽  
Neutron Transport ◽  
Memory Bandwidth ◽  
Performance Gain ◽  
Computational Power ◽  
Sorting Algorithms ◽  
Main Factors ◽  
Gpu Architectures ◽  
Monte Carlo Transport

GPUs have gradually increased in computational power from the small, job-specific boards of the early 90s to the programmable powerhouses of today. Compared to CPUs, they have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU optimized parallel algorithms are not directly portable to GPU architectures (or at least without losing substantial performance gain), transport codes need to be rewritten in order to execute efficiently on GPUs. Unless this is done, we cannot take full advantage of these new supercomputers for reactor simulations. In this work, we attempt to efficiently map the Monte Carlo transport algorithm on the GPU while preserving its benefits, namely, very few physical and geometrical simplifications. Regularizing memory access and introducing parallel-efficient search and sorting algorithms are the main factors in completing the task.

scholarly journals Collisions of Nucleons with Atoms: Calculated Cross Sections and Monte Carlo Simulation

2021 ◽  
Vol 9 ◽  
Author(s):  
Francesc Salvat ◽  
José Manuel Quesada
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Impact Ionization ◽  
Neutron Transport ◽  
Emission Spectra ◽  
Monte Carlo Code ◽  
Dose Distributions ◽  
Elastic Collisions ◽  
Proton Impact

After a summary description of the theory of elastic collisions of nucleons with atoms, we present the calculation of a generic database of differential and integrated cross sections for the simulation of multiple elastic collisions of protons and neutrons with kinetic energies larger than 100 keV. The relativistic plane-wave Born approximation, with binding and Coulomb-deflection corrections, has been used to calculate a database of proton-impact ionization of K-shell and L-, M-, and N-subshells of neutral atoms These databases cover the whole energy range of interest for all the elements in the periodic system, from hydrogen to einsteinium (Z = 1–99); they are provided as part of the penh distribution package. The Monte Carlo code system penh for the simulation of coupled electron-photon-proton transport is extended to account for the effect of the transport of neutrons (released in proton-induced nuclear reactions) in calculations of dose distributions from proton beams. A simplified description of neutron transport, in which neutron-induced nuclear reactions are described as a fractionally absorbing process, is shown to give simulated depth-dose distributions in good agreement with those generated by the Geant4 code. The proton-impact ionization database, combined with the description of atomic relaxation data and electron transport in penelope, allows the simulation of proton-induced x-ray emission spectra from targets with complex geometries.

Evaluation of the Activated Radioactivity of Turbine Equipments in BWR

2010 ◽  
Author(s):  
Masato Watanabe ◽  
Motonori Nakagami
Keyword(s):  
Cross Sections ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Group Structure ◽  
Track Detector ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Main Stream ◽  
Water Reactor ◽  
Main Steam

The activated radioactivity of turbine equipments irradiated by neutron originating from 17N in the main stream is evaluated for an introduction of clearance system to boiling-water reactor (BWR) plant. The 17N, main neutron source is generated by 17O(n, p)17N reaction in the core region. The evaluation results clarified that the activated radioactivity of the turbine equipment is extremely small comparing to the clearance level. The feature of the evaluation is as follows. (1) Actual radioactive concentration of the 17N in the main steam in Hamaoka nuclear power station unit 5 (Hamaoka-5) which is an advanced boiling-water reactor (ABWR) was measured with solid-state track detector (SSTD). The 17N concentration is used for the neutron transport calculation as initial neutron sources. (2) The turbine equipments were modeled as two-dimensional geometry for DORT code. (3) Activation cross-sections for major nuclides subject to the clearance evaluation were based on JENDL3.3 on 175 energy group structure (VITAMIN-J). (4) Minor nuclides subject to the clearance evaluation were calculated with ORIGEN-S code.

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 Demonstration of On-the-Fly Generation of Unresolved Resonance Region Cross Sections in a Monte Carlo Transport Code

2014 ◽  
Author(s):  
Jonathan Walsh ◽  
Brian Kiedrowski ◽  
Benoit Forget ◽  
Kord Smith ◽  
Forrest Brown
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Resonance Region ◽  
Transport Code ◽  
Monte Carlo Transport
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