scholarly journals DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 06041
Author(s):  
Brian J. Ade ◽  
Nicholas P. Luciano ◽  
Andrew J. Conant ◽  
Cole A. Gentry ◽  
Shane G. Stimpson ◽  
...  
Keyword(s):  
Neutron Transport ◽  
National Laboratory ◽  
Test Problems ◽  
High Fidelity ◽  
Validation Data ◽  
Oak Ridge ◽  
Section Data ◽  
Axial Dimension ◽  
And Performance ◽  
Full Core

The MPACT code, jointly developed by Oak Ridge National Laboratory and University of Michigan, is designed to perform high-fidelity light water reactor (LWR) analysis using wholecore pin-resolved neutron transport calculations on modern parallel-computing hardware. MPACT uses the subgroup method for resonance self-shielding, while the primary neutron transport solver uses a 2D/1D method that is based on the method of characteristics (MoC) for the x-y planes coupled with a 1D diffusion or transport solver in the axial dimension. Additional geometry capabilities are currently being developed in MPACT to support hexagonal-pitched lattices, as well as interstitial geometry (i.e., control rods at the corner of four adjacent pin cells). In this research, the MPACT method is tested on gas-cooled reactors by applying MPACT to full-core MAGNOX reactor test problems. MAGNOX test problems were chosen due to the availability of high-quality reactor design and validation data (available through an ongoing collaboration with the National Nuclear Laboratory in the United Kingdom) and the existence of a relatively complex axial power shape that is expected to challenge the MPACT method. MPACT’s convergence for partial- and full-core problems will be tested and verified. MPACT will be compared with high-fidelity continuous-energy Monte Carlo simulations to verify core reactivity, power distributions, and performance of the available cross section data libraries and energy group structures.

scholarly journals Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

2013 ◽  
Vol 2013 ◽  
pp. 1-9 ◽  
Author(s):  
Mario Matijević ◽  
Dubravko Pevec ◽  
Krešimir Trontl
Keyword(s):  
Cross Sections ◽  
Neutron Transport ◽  
National Laboratory ◽  
General Purpose ◽  
Monte Carlo Code ◽  
Oak Ridge ◽  
Benchmark Data ◽  
Selection Of ◽  
Good Agreement ◽  
Reactor Dosimetry

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

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.

Early Results from an Aberration-Corrected JEOL 2200FS STEM/TEM at Oak Ridge National Laboratory

2006 ◽  
Vol 12 (6) ◽  
pp. 483-491 ◽  
Author(s):  
Douglas A. Blom ◽  
Lawrence F. Allard ◽  
Satoshi Mishina ◽  
Michael A. O'Keefe
Keyword(s):  
National Laboratory ◽  
Oak Ridge ◽  
Construction Methods ◽  
Early Results ◽  
Oak Ridge National Laboratory ◽  
Electron Microscopes ◽  
And Performance ◽  
Aberration Corrected ◽  
Suitable Location

The resolution-limiting aberrations of round electromagnetic lenses can now be successfully overcome via the use of multipole element “aberration correctors.” The installation and performance of a hexapole-based corrector (CEOS GmbH) integrated on the probe-forming side of a JEOL 2200FS FEG STEM/TEM is described. For the resolution of the microscope not to be severely compromised by its environment, a new, specially designed building at Oak Ridge National Laboratory has been built. The Advanced Microscopy Laboratory was designed with the goal of providing a suitable location for aberration-corrected electron microscopes. Construction methods and performance of the building are discussed in the context of the performance of the microscope. Initial performance of the microscope on relevant specimens and modifications made to eliminate resolution-limiting conditions are also discussed.

scholarly journals SERPENT/SUBCHANFLOW COUPLED CALCULATIONS FOR A VVER CORE AT HOT FULL POWER

2021 ◽  
Vol 247 ◽  
pp. 04006
Author(s):  
Diego Ferraro ◽  
Manuel García ◽  
Uwe Imke ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
...  
Keyword(s):  
High Performance ◽  
Neutron Transport ◽  
Coupling Scheme ◽  
High Fidelity ◽  
Reactor Physics ◽  
Coupled Calculation ◽  
Final Goal ◽  
Wide Range ◽  
Computational Resources ◽  
Full Core

An increasing interest on the development of highly accurate methodologies in reactor physics is nowadays observed, mainly stimulated by the availability of vast computational resources. As a result, an on-going development of a wide range of coupled calculation tools is observed within diverse projects worldwide. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes. These tools are aimed to be suitable for high-fidelity calculations both for PWR and VVER reactors, with the final goal of performing pin-by-pin coupled calculations at full core scope including burnup. Several intermediate steps are to be analyzed in-depth before jumping into this final goal in order to provide insights and to identify resources requirements. As part of this process, this work presents the results for a pin-by-pin coupling calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel code SUBCHANFLOW (SCF, developed by KIT, Germany) for a full-core VVER model. For such purpose, a recently refurbished master-slave coupling scheme is used within a High Performance Computing architecture. A full-core benchmark for a VVER-1000 that provides experimental data is considered, where the first burnup step (i.e. fresh core at hot-full rated power state) is calculated. For such purpose a detailed (i.e. pin-by-pin) coupled Serpent-SCF model is developed, including a simplified equilibrium xenon distribution (i.e. by fuel assembly). Comparisons with main global reported results are presented and briefly discussed, together with a raw estimation of resources requirements and a brief demonstration of the inherent capabilities of the proposed approach. The results presented here provide valuable insights and pave the way to tackle the final goals of the on-going high-fidelity project.

scholarly journals CONTINUOUS ENERGY COMET SOLUTION TO A SMALL MODULAR ADVANCED HIGH-TEMPERATURE REACTOR BENCHMARK PROBLEM (SmAHTR)

2021 ◽  
Vol 247 ◽  
pp. 05002
Author(s):  
Farzad Rahnema ◽  
Dingkang Zhang
Keyword(s):  
Monte Carlo ◽  
High Temperature ◽  
National Laboratory ◽  
Benchmark Problems ◽  
High Fidelity ◽  
Oak Ridge ◽  
Continuous Energy ◽  
High Temperature Reactor ◽  
Relative Differences ◽  
High Computational Efficiency

The continuous energy coarse mesh transport (COMET) method is a hybrid stochasticdeterministic solver that provides transport solutions to heterogeneous reactor cores. In this paper, COMET is tested against continuous energy Monte Carlo in solving the recently developed stylized Small Modular Advanced High-Temperature Reactor (SmAHTR) Benchmark Problems based on the Oak Ridge National Laboratory pre-conceptual design (core configurations). These problems are well-suited to test the performance of advanced neutronics tools because of their unique neutronics characteristics such as the multiple heterogeneities. The COMET solutions for the three benchmark problems were found to agree very well with the continuous energy Monte Carlo reference solutions. The discrepancy in the core eigenvalue (k-eff) varied from 40 pcm to 51 pcm. The average and maximum relative differences in the pin fission densities were in the range of 0.20% to 0.21% and 0.77% to 0.94%, respectively. It was also found that COMET was more than 2,000 times fast than MCNP. It can be concluded that COMET can model the SmAHTR core configuration with high fidelity and significantly high computational efficiency.

Seismically Detecting Nuclear Reactor Operations Using a Power Spectral Density (PSD) Misfit Detector

2021 ◽  
Author(s):  
David L. Guenaga ◽  
Chengping Chai ◽  
Monica Maceira ◽  
Omar E. Marcillo ◽  
Aaron A. Velasco
Keyword(s):  
Spectral Density ◽  
Power Spectral Density ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Waveform Data ◽  
Oak Ridge ◽  
Power Spectral ◽  
And Performance ◽  
Additional Station ◽  
Initial Results

ABSTRACT To explore the ability to indirectly detect and attribute various operations conducted at a nuclear reactor using waveform data, we investigated the seismic signals recorded near the High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory in Oak Ridge, Tennessee. Specifically, we processed seismic data collected from a single seismoacoustic station, WACO, near the HFIR facility, and employed a power spectral density misfit detector to identify signals of interest and associate the detections with operational events. Initial results suggest that this method provides a promising means of regularly detecting at least 19 unique operations. With additional station deployment and more comprehensive data logs, we anticipate that future analysis will offer an additional means to seismically monitor nuclear reactors (such as HFIR) health and performance more accurately.

The Bio-SANS instrument at the High Flux Isotope Reactor of Oak Ridge National Laboratory

2014 ◽  
Vol 47 (4) ◽  
pp. 1238-1246 ◽  
Author(s):  
William T. Heller ◽  
Volker S. Urban ◽  
Gary W. Lynn ◽  
Kevin L. Weiss ◽  
Hugh M. O'Neill ◽  
...  
Keyword(s):  
Neutron Scattering ◽  
National Laboratory ◽  
Isotopic Labeling ◽  
High Flux ◽  
Oak Ridge ◽  
Oak Ridge National Laboratory ◽  
User Facility ◽  
Technical Specifications ◽  
And Performance ◽  
Sans Instrument

Small-angle neutron scattering (SANS) is a powerful tool for characterizing complex disordered materials, including biological materials. The Bio-SANS instrument of the High Flux Isotope Reactor of Oak Ridge National Laboratory (ORNL) is a high-flux low-background SANS instrument that is, uniquely among SANS instruments, dedicated to serving the needs of the structural biology and biomaterials communities as an open-access user facility. Here, the technical specifications and performance of the Bio-SANS are presented. Sample environments developed to address the needs of the user program of the instrument are also presented. Further, the isotopic labeling and sample preparation capabilities available in the Bio-Deuteration Laboratory for users of the Bio-SANS and other neutron scattering instruments at ORNL are described. Finally, a brief survey of research performed using the Bio-SANS is presented, which demonstrates the breadth of the research that the instrument's user community engages in.

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.

Plug-in scripts for EFTEM automation

1996 ◽  
Vol 54 ◽  
pp. 546-547
Author(s):  
N. D. Evans ◽  
M. K. Kundmann
Keyword(s):  
Electron Microscopy ◽  
National Laboratory ◽  
Oak Ridge ◽  
Multiple User ◽  
Transmission Electron ◽  
Short Period ◽  
User Facility ◽  
Instrument Control ◽  
And Control ◽  
Research Equipment

Post-column energy-filtered transmission electron microscopy (EFTEM) is inherently challenging as it requires the researcher to setup, align, and control both the microscope and the energy-filter. The software behind an EFTEM system is therefore critical to efficient, day-to-day application of this technique. This is particularly the case in a multiple-user environment such as at the Shared Research Equipment (SHaRE) User Facility at Oak Ridge National Laboratory. Here, visiting researchers, who may oe unfamiliar with the details of EFTEM, need to accomplish as much as possible in a relatively short period of time.We describe here our work in extending the base software of a commercially available EFTEM system in order to automate and streamline particular EFTEM tasks. The EFTEM system used is a Philips CM30 fitted with a Gatan Imaging Filter (GIF). The base software supplied with this system consists primarily of two Macintosh programs and a collection of add-ons (plug-ins) which provide instrument control, imaging, and data analysis facilities needed to perform EFTEM.

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