scholarly journals Thermal-fluid-structure coupling analysis on plate-type fuel assembly under irradiation. Part-II Mechanical deformation and thermal-hydraulic characteristics

2020 ◽  
Author(s):  
Yuanming Li ◽  
Quan-yao Ren ◽  
Pan Yuan ◽  
Guanghui Su ◽  
Hongxing Yu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Mechanical Deformation ◽  
Hydraulic Characteristics ◽  
Coupling Analysis ◽  
Fluid Structure ◽  
Type Fuel ◽  
Plate Type

Coupled Calculation on Fluid Structure Interaction in Plate-Type Fuel Element

2018 ◽  
Author(s):  
Yiqi Yu ◽  
Elia Merzari ◽  
Jerome Solberg
Keyword(s):  
Fuel Element ◽  
Fluid Structure Interaction ◽  
Initial Step ◽  
University Of Missouri ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Coupled Calculation ◽  
Enriched Uranium ◽  
Type Fuel ◽  
Plate Type

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.

scholarly journals Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

2020 ◽  
Vol 2020 ◽  
pp. 1-12 ◽  
Author(s):  
Linrong Ye ◽  
Mingjun Wang ◽  
Xin’an Wang ◽  
Jian Deng ◽  
Yan Xiang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Coupling Method ◽  
High Fidelity ◽  
Monte Carlo Code ◽  
Coupling Analysis ◽  
Fuel Reactor ◽  
Nuclear Reactor Core ◽  
Type Fuel ◽  
Plate Type

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

scholarly journals DROP IMPACT ANALYSIS OF PLATE-TYPE FUEL ASSEMBLY IN RESEARCH REACTOR

2014 ◽  
Vol 46 (4) ◽  
pp. 529-540 ◽  
Author(s):  
HYUN-JUNG KIM ◽  
JEONG-SIK YIM ◽  
BYUNG-HO LEE ◽  
JAE-YONG OH ◽  
YOUNG-WOOK TAHK
Keyword(s):  
Fuel Assembly ◽  
Impact Analysis ◽  
Research Reactor ◽  
Drop Impact ◽  
Type Fuel ◽  
Plate Type

Uncertainty analysis of criticality safety for the plate type fuel assembly storage rack

2013 ◽  
Vol 53 ◽  
pp. 82-89 ◽  
Author(s):  
Tae Young Han ◽  
Chang Je Park ◽  
Byung Chul Lee ◽  
Jae Man Noh
Keyword(s):  
Uncertainty Analysis ◽  
Fuel Assembly ◽  
Type Fuel ◽  
Plate Type ◽  
Criticality Safety

Development of Real-Time Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor

2010 ◽  
Author(s):  
Lei Li ◽  
Zhijian Zhang
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Real Time ◽  
Transient Flow ◽  
Rectangular Channel ◽  
Hydraulic Analysis ◽  
Fuel Reactor ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type

A multi-channel model thermal-hydraulic analysis code in real-time for plate type fuel reactor is developed in this paper. In this code, every fuel assembly in reactor is divided into a subchannel. A series of reasonable mathematical and physical model are set up based on the structure and operational characteristics of plate type fuel core. As for the choice of flow friction and heat transfer models, all possible flow regimes which include the laminar flow, transient flow and turbulent flow, and heat transfer regimes which include single liquid phase heat transfer, sub-cooled boiling, saturation boiling, film boiling and single vapor phase heat transfer, are considered. The correlations and constitutive equations used in the code are fit for the rectangular channel. Look-up table method is used to calculate the properties of water and steam. The code has been loaded on the real-time simulation supporting system SimExec. The reactivity insertion accident and loss of flow accident, which has been defined in the IAEA 10MW MTR benchmark program, were calculated by the code in this paper for validation. Furthermore, the steady state of CARR (China Advanced Research Reactor) is analyzed by this code. The detailed flow distribution in each fuel assembly is obtained. The temperature of coolant, quality, void fraction, DNBR in each subchannel is calculated. The results show that the recently developed code can be used for real time thermal hydraulic analysis of plate type fuel reactor.

Measurement of core flow distribution in a research reactor using plate-type fuel assembly

2020 ◽  
Vol 148 ◽  
pp. 107750
Author(s):  
Jonghark Park ◽  
Hyung Min Son ◽  
Kiwon Song ◽  
Ragai M. Altamimi
Keyword(s):  
Fuel Assembly ◽  
Research Reactor ◽  
Flow Distribution ◽  
Core Flow ◽  
Type Fuel ◽  
Plate Type

Electron Beam Weldability of AA6061-T6 Aluminum Straps for U-Mo Follower Fuel Assembly Manufacturing

2016 ◽  
Author(s):  
Soo-sung Kim ◽  
Yong-jin Jeong ◽  
Jong-man Park ◽  
Yoon-sang Lee ◽  
Chong-tak Lee
Keyword(s):  
Electron Beam ◽  
Fuel Assembly ◽  
Cross Sections ◽  
Electron Beam Welding ◽  
High Vacuum ◽  
Welding Parameters ◽  
Depth Of Penetration ◽  
Shearing Strength ◽  
Type Fuel ◽  
Plate Type

A procedure for Electron Beam Welding (EBW) was developed for the manufacturing of a follower fuel assembly made of an AA 6061-T6 aluminum straps for a U-Mo plate-type fuel proposed to be used in the future in Korea’s Kijang Research Reactor (KJRR) project. The initial welding trials of the weld samples were carried out with a high vacuum chamber using the EBW process. After investigating the welds, EB welding parameters for the full-sized samples were optimized for the required depth of penetration and weld quality. Subsequently, the weld samples made by the filler specimens showed higher shearing strengths than those of the non-filler specimens. This procedure made by EBW process was also confirmed based on the results of the shearing strength test, an examination of the macro-cross sections, and the fracture surfaces of the welded specimens.

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