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CFD Analysis of Nuclear Engineering

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A special issue of Applied Sciences (ISSN 2076-3417). This special issue belongs to the section "Applied Physics General".

Deadline for manuscript submissions: closed (31 December 2024) | Viewed by 5653

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Special Issue Editors

Dr. Amir F. Ali

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Guest Editor

Nuclear Engineering Department, Idaho State University, Pocatello, ID 83209, USA
Interests: accident tolerant fuel; liquid metals reactors; thermal hydraulics; CFD; fire safety

Dr. Prashant K Jain

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Guest Editor

Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831, USA
Interests: thermal hydraulics; CFD; parallel programming nuclear safety

Prof. Dr. Chad L. Pope

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Guest Editor

Nuclear Engineering Department, Idaho State University, Pocatello, ID 83209, USA
Interests: risk analysis; reactor physics; nuclear fuel; safeguards

Special Issue Information

Dear Colleagues,

The method known as ‘computational fluid dynamics’ (CFD) constitutes a crucial tool for developing diverse engineering applications, predominantly nuclear iterations. CFD improved the analysis and prediction of single-phase and multi-phase flows under steady-state or transient scenarios in nuclear reactor engineering. It has been utilized for current and advanced reactor fleets, including liquid metal, molten salt, and gas-cooled reactors. CFD also drives the integration of validation experiments and advances their coupling with other modeling tools to address many aspects of nuclear reactor sciences including, but not limited to, neotropics, and material science.

In this Special Issue, we invite submissions exploring cutting-edge research and recent advances in CFD, including development, current status, and challenges for nuclear application. We particularly welcome research that develops new methods in CFD for high-fidelity results and the utilization of CFD in designing separate and integral effects experiments for validation. Potential submission topics include CFD single- and multi-phase simulation, open-source CFD code development and application, and porous medium model application. Discussion of multiphysics modeling using coupled CFD tools with MCNP, RELAP5, and more for conducting coupled neutronics, thermal hydraulics, and thermal–structural analysis is also welcome.

Dr. Amir F. Ali
Dr. Prashant K Jain
Prof. Dr. Chad L. Pope
Guest Editors

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Keywords

CFD
nuclear reactor
thermal hydraulics
high-fidelity simulation
multiphysics simulation
CFD validation

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Published Papers (4 papers)

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Research

26 pages, 8106 KiB

Open AccessArticle

A Framework for Multi-Physics Modeling, Design Optimization and Uncertainty Quantification of Fast-Spectrum Liquid-Fueled Molten-Salt Reactors

by David Holler, Sandesh Bhaskar, Grigirios Delipei, Maria Avramova and Kostadin Ivanov

Appl. Sci. 2024, 14(17), 7615; https://doi.org/10.3390/app14177615 - 28 Aug 2024

Viewed by 1273

Abstract

The analysis of liquid-fueled molten-salt reactors (LFMSRs) during steady state, operational transients and accident scenarios requires addressing unique reactor multi-physics challenges with coupling between thermal hydraulics, neutronics, inventory control and species distribution phenomena. This work utilizes the General Nuclear Field Operation and Manipulation (GeN-Foam) code to perform coupled thermal-hydraulics and neutronics calculations of an LFMSR design. A framework is proposed as part of this study to perform modeling, design optimization, and uncertainty quantification. The framework aims to establish a protocol for the studies and analyses of LFMSR which can later be expanded to other advanced reactor concepts too. The Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) statistical analysis tool was successfully coupled with GeN-Foam to perform uncertainty quantification studies. The uncertainties were propagated through the input design parameters, and the output uncertainties were characterized using statistical analysis and Spearman rank correlation coefficients. Three analyses are performed (namely, scalar, functional, and three-dimensional analyses) to understand the impact of input uncertainty propagation on temperature and velocity predictions. Preliminary three-dimensional reactor analysis showed that the thermal expansion coefficient, heat transfer coefficient, and specific heat of the fuel salt are the crucial input parameters that influence the temperature and velocity predictions inside the LFMSR system. Full article

(This article belongs to the Special Issue CFD Analysis of Nuclear Engineering)

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19 pages, 62520 KiB

Open AccessArticle

Investigation of Point-Contact Strategies for CFD Simulations of Pebble-Bed Reactor Cores

by Nolan Goth, Thien Nguyen and William David Pointer

Appl. Sci. 2024, 14(16), 7343; https://doi.org/10.3390/app14167343 - 20 Aug 2024

Viewed by 1438

Abstract

This study numerically investigated the effects of various contact strategies on the thermal hydraulic behavior within a structured bed of 100 explicitly modeled pebbles. Four contact strategies and two thermal hydraulic conditions were considered. The strategies to avoid contact singularities include decreasing the pebble diameter, increasing the pebble diameter, bridging the pebble surfaces near the contact region, and capping the pebble surfaces near the contact region. One strategy, Strategy 3a, which involves bridging with a cylinder equal to 10% of the pebble diameter, was selected as the baseline strategy because it addressed the contact singularity while minimizing the geometric changes that affect the bed porosity. The two thermal hydraulic conditions were full-power operation (Case 1) and pressurized loss of forced cooling or PLOFC (Case 2). Simulations of the conjugate heat transfer within the structured bed were performed using the Reynolds-averaged Navier–Stokes approach with the realizable k-

ϵ

turbulence model and two-layer all

y^{+}

wall treatment. The thermal-fluid quantities of interest were compared between the contact strategies for each case. In Case 1, the hydraulic behavior was sensitive to the contact strategy, with large differences in the pressure drop (30%) and volume-average velocity (4%). The thermal behavior was not sensitive, with less than a 0.5% difference across the strategies. To better understand the separate effects of each heat transfer mode, Case 2 was divided into the following subcases: conduction (Case 2a); conduction/radiation (Case 2b); and conduction/radiation/convection (Case 2c). Case 2a represents an early phase of the PLOFC transient. Case 2b represents an intermediate phase of the PLOFC transient, with the pebble temperatures sufficiently high for the radiative heat transfer to be non-negligible. Case 2c represents a late phase of the PLOFC transient after the establishment of the natural circulation of the heat transfer fluid. For Case 2, large differences in the contact strategy were observed only in Case 2a with only conduction. The difference in the maximum pebble temperature was 23% in Case 2a, 2% in Case 2b, and 0.3% in Case 2c. Full article

(This article belongs to the Special Issue CFD Analysis of Nuclear Engineering)

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21 pages, 10037 KiB

Open AccessArticle

Validation and Application of a Code for Three-Dimensional Analysis of Hydrogen–Steam Behavior in a Nuclear Reactor Containment during Severe Accidents

by Jongtae Kim and Kukhee Lim

Appl. Sci. 2024, 14(15), 6695; https://doi.org/10.3390/app14156695 - 31 Jul 2024

Cited by 1 | Viewed by 1196

Abstract

In a pressurized water reactor (PWR) during a loss of coolant accident (LOCA) or a station blackout (SBO) accident, water and steam are released into the containment building. The water vapor mixes with the atmosphere, partially condensing into droplets or condensing on the containment walls. Although a significant amount of water vapor condenses, it coexists with hydrogen generated by the reactor core oxidation. As water vapor condenses, the volume fraction of hydrogen increases, raising the risk of explosion or flame acceleration. As such, water vapor’s behavior directly affects hydrogen distribution. To conservatively evaluate hydrogen safety in a PWR during a severe accident, lumped-parameter codes have been heavily used. As a best-estimate approach for hydrogen safety analysis in a PWR containment, a turbulence-resolved CFD code called contain3D has been developed. This paper presents the validation results of the code and simulation results of hydrogen behavior affected by water vapor condensation and hydrogen removal by passive autocatalytic recombiners (PARs) in the APR1400 containment. The results provide insight into the three-dimensional behaviors of the hydrogen in the containment. Full article

(This article belongs to the Special Issue CFD Analysis of Nuclear Engineering)

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24 pages, 24251 KiB

Open AccessArticle

A New Development of Cross-Correlation-Based Flow Estimation Validated and Optimized by CFD Simulation

by Xiong Gao, Lane B. Carasik, Jamie B. Coble and J. Wesley Hines

Appl. Sci. 2024, 14(15), 6687; https://doi.org/10.3390/app14156687 - 31 Jul 2024

Cited by 1 | Viewed by 958

Abstract

The accurate measurement of mass flow rates is important in nuclear power plants. Flow meters have been invented and widely applied in several industries; however, the operating environment in advanced nuclear power plants is especially harsh due to high temperatures, high radiation, and potentially corrosive conditions. Traditional flow meters are largely limited to deployment at the outlet of pumps, on pipes, or in limited geometries. Cross-correlation function (CCF) flow estimation, on the other hand, can estimate the flow velocity indirectly without any specific instruments for flow measurement and in any geometry of the flow region. CCF flow estimation relies on redundant instruments, typically temperature sensors, in series in the direction of flow. One challenge for CCF flow estimation is that the accuracy of the flow measurement is mainly determined by inherent, common local process variation across the sensors, which may be small compared to the uncorrelated measurement noise. To differentiate the process variations from the uncorrelated noise, this research implements periodic fluid injection at a different temperature than the bulk fluid before the temperature sensors to amplify process variation. The feasibility and accuracy of this method are investigated through flow loop experiments and Computational Fluid Dynamics (CFD) simulations. This paper focuses on a CFD simulation model to verify the previous experimental results and optimize CCF flow estimation with different configurations. The optimization study is carried out to perform a grid search on the optimal location of the sensor pair under different flow rates. The CFD results show that the optimal sensor spacing depends on the flow rate being measured and provides guidance for sensor location implementation under various anticipated flow rates. Full article

(This article belongs to the Special Issue CFD Analysis of Nuclear Engineering)

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