A review of recent advances in HTGR CFD and thermal fluid analysis

Huning, Alexander J.; Chandrasekaran, Sriram; Garimella, Srinivas

doi:10.1016/j.nucengdes.2020.111013

Cited by 28 publications

(7 citation statements)

References 96 publications

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“…Lower plenum hot streaking can occur when certain coolant channels and bypass gaps experience higher than average heat flux values due to power peaking [19]. Coolant temperatures near the core center and periphery will be less than that of coolant in the annular active core region.…”

Section: Lower Plenummentioning

confidence: 99%

High-Temperature Gas-Cooled Reactor Research Survey and Overview: Preliminary Data Platform Construction for the Nuclear Energy University Program

Qin

Minseop

Vietz

et al. 2022

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Since the U.S. Department of Energy Office of Nuclear Energy initiated the Nuclear Energy University Program (NEUP) in 2009, there are 29 NEUP projects focusing on high-temperature gas-cooled reactor (HTGR) research up to July 2022. The resultant research product, either experimental or computational, were published as final NEUP reports, journal articles and conference proceedings. However, these federally funded products have been scattered and sometimes cannot be easily accessed.To improve access to this valuable HTGR validation data and optimize the return on the significant investment made by the Department of Energy, the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program started a survey of completed and ongoing HTGR NEUP projects to develop a public-access database specific for HTGRs applications that can be used to retrieve computational fluid dynamics and system code validation data. This effort will help guide future NEUP-funded research, define new state of the ART Phenomena Identification and Ranking Table (PIRT), and promote the usage of this data in the codes validation matrices. This report provides an overview of the NEUP-funded HTGR-related research projects from Fiscal Year (FY) 2009-2021 and identifies validation knowledge gaps still existing in HTGR thermal-fluid research.A preliminary data platform has been developed for the 29 NEUP projects investigating HTGR thermal hydraulics, including their final reports as well as the available scientific publications. As an ultimate goal for this work, the ART-GCR program will create a central database at Idaho National Laboratory to identify, organize, and store these datasets generated by experimental investigations or computational models, experimental facility descriptions, and publicly-available academic products from the HTGR-related NEUP projects and provide future guidance for the storage and transmission of important project documentations for later NEUP projects as well.

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Section: Lower Plenummentioning

confidence: 99%

High-Temperature Gas-Cooled Reactor Research Survey and Overview: Preliminary Data Platform Construction for the Nuclear Energy University Program

Qin

Minseop

Vietz

et al. 2022

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“…Computational fluid dynamics (CFD) has been used to examine TH parameters in complex heat exchanger systems with simplifications/assumptions to reduce the computational cost without affecting the results accuracy. [33][34][35][36] In a similar manner and due to the complex geometry of the TES (1000 blocks with each block containing 11 000 modular containers that are triangularly positioned), the approach adopted in the current study focuses on conducting a TH analysis of a single modular container. This can be justified by the fact that the flow conditions and melting/solidification processes within the modular containers are almost identical due to similar inlet and outlet boundary conditions.…”

Section: Tes-npp Coupled System and Model Descriptionmentioning

confidence: 99%

“…This approach has been used by many researchers to perform TH analysis not only for TES and heat exchanger 33 but also for sub-channel TH analysis within nuclear reactor cores. [34][35][36] Figure 2 presents the simplified schematic of the TES block, cross-sectional view of three PCM containers, and simplified one modular container with helium region. Given the large perimeter/diameter ratio, the assumption that there is no secondary flow between adjacent sub-channels is considered.…”

Section: Tes-npp Coupled System and Model Descriptionmentioning

confidence: 99%

Numerical investigation of a vertical triplex‐tube latent heat storage/exchanger to achieve flexible operation of nuclear power plants

Ali

Alkaabi

Addad

2021

Intl J of Energy Research

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The current numerical investigation is carried out to assess the possibility of integrating a latent heat thermal energy storage (TES) in a nuclear power plant (NPP) system. This would allow improving the capability of such plants to operate in a load-following mode by eliminating the gap between energy supply and energy demand by the grid. In contrary to previously published TES designs, the phase change material (PCM) triplex-tube containers are vertically oriented in the present study to take advantage of melting/solidification enhancements induced by the buoyancy forces. In addition, two different fluids simultaneously co-circulating within the storage, hence, allowing for this component to operate as a storage system (under extreme charging/discharging conditions) or as simple heat exchanger (under normal steady-state conditions). The computational domain, in this study, is reduced from the full TES geometry to a single triplextube container owing to the homogeneous triangular positioning of the annular tubes within the storage vessel. The results show that the time required to fully melt the solid PCM is around 27 hours during charging mode, thus, allowing for a relatively long-time margin even under an extreme postulated accident scenario consisting in the inability of the secondary loop to remove heat from the system. In the opposite discharging mode, the time needed to fully solidify the 100% liquid PCM is 17 hours. During the simultaneous charging and discharging (SCD) mode, with constant load, the fraction of liquid PCM is nearly fixed with time. Hence, TES in this mode is said to take the role of a standard heat exchanger. Finally, simulations conducted for the SCD mode, with variable load, demonstrate that the coupled system can follow the daily variations in the grid demand without having any significant impact on the constant operation of the reactor. In this operating mode, the PCM is changing phases from solid to liquid and vice versa to adapt to the energy demand variations as anticipated.

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“…To accurately predict the peak fuel temperature during such transient events, it requires the simulation tools to capture the heat transfer in both length scales. For prismatic HTGRs, two popular methods for performing thermal analysis on prismatic HTGR cores include the unit-cell and CFD methods [2]. The unit-cell approach can be seen as a simplified and homogenized method that uses a unit cell to represent the averaged behavior of a larger domain, such as a fuel assembly.…”

Section: Introductionmentioning

confidence: 99%

Ring Model Development and Validation for Prismatic HTGR Core Thermal-hydraulics and Safety Analysis

Hua¹,

Ooi²,

Hu³

et al. 2021

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The Laboratory's main facility is outside Chicago, at 9700 South Cass Avenue, Argonne, Illinois 60439. For information about Argonne and its pioneering science and technology programs, see www.anl.gov. DOCUMENT AVAILABILITYOnline Access: U.S. Department of Energy (DOE) reports produced after 1991 and a growing number of pre-1991 documents are available free at OSTI.GOV (http://www.osti.gov/), a service of the US Dept. of Energy's Office of Scientific and Technical Information.

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A review of recent advances in HTGR CFD and thermal fluid analysis

Cited by 28 publications

References 96 publications

High-Temperature Gas-Cooled Reactor Research Survey and Overview: Preliminary Data Platform Construction for the Nuclear Energy University Program

High-Temperature Gas-Cooled Reactor Research Survey and Overview: Preliminary Data Platform Construction for the Nuclear Energy University Program

Numerical investigation of a vertical triplex‐tube latent heat storage/exchanger to achieve flexible operation of nuclear power plants

Ring Model Development and Validation for Prismatic HTGR Core Thermal-hydraulics and Safety Analysis

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