Temperature Distribution in the Duct Wall and at the Exit of a 19-Rod Simulated LMFBR Fuel Assembly (FFM Bundle 2A)

1974 ◽  
Vol 24 (2) ◽  
pp. 176-200 ◽  
Author(s):  
M. H. Fontana ◽  
R. E. MacPherson ◽  
P. A. Gnadt ◽  
L. F. Parsly ◽  
J. L. Wantland
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Duct Wall

Investigation of Wire Effect in a Wire-Wrapped 37-Pin Fuel Assembly

2015 ◽  
Author(s):  
Jae-Ho Jeong ◽  
Jin Yoo ◽  
Kwi-Lim Lee ◽  
Kwi-Seok Ha
Keyword(s):  
Fuel Assembly ◽  
Relative Position ◽  
Large Scale ◽  
Vortex Structures ◽  
Secondary Flows ◽  
Vortical Flow ◽  
Small Scale ◽  
Multi Scale ◽  
Duct Wall ◽  
The Wire

The wire effect in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor (SFR), Monju, has been investigated through a numerical analysis using a general-purpose commercial computational fluid dynamics (CFD) code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-Averaged Navier-Stokes (RANS) flow simulation with a shear stress transport (SST) turbulence model. The CFD results show good agreement with Rehme’s friction factor correlation model, which can consider the number of wire-wrapped pins in the fuel assembly. Three-dimensional multi-scale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Large-scale and small-scale vortex structures are generated in the corner and edge, and interior sub-channel, respectively. The behavior of the large-scale vortex structures in the corner and edge sub-channel are closely related to the relative position between the hexagonal duct wall and the wire spacer. Regardless of the relative position between the adjacent rod and wire spacer, a small-scale vortex is axially developed in the interior sub-channels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and wire spacer. It is expected that the multi-scale vortex structures in the fuel assembly play a significant role in the convective heat transfer characteristics.

Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment

2019 ◽  
Vol 111 ◽  
pp. 174-182 ◽  
Author(s):  
Tangtao Feng ◽  
Wenxi Tian ◽  
Ping Song ◽  
Jun Wang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Spatial Temperature Distribution ◽  
Spatial Temperature

scholarly journals Sleeve leakage gas impact on fuel assembly temperature distribution

2016 ◽  
Vol 30 (6) ◽  
pp. 419-424 ◽  
Author(s):  
Charles Moulinec ◽  
Juan C. Uribe ◽  
Jim Gotts ◽  
Bing Xu ◽  
David R. Emerson
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly

CFD investigation for a 7-pin wire-wrapped fuel assembly with different shapes of fuel duct wall

2020 ◽  
Vol 141 ◽  
pp. 107272
Author(s):  
Yuanyuan Zhao ◽  
Mei Huang ◽  
Jiyuan Huang ◽  
Xiaoping Ouyang ◽  
Rongbin Hou
Keyword(s):  
Fuel Assembly ◽  
Duct Wall ◽  
Different Shapes

Investigation of Mixing Flow Process Using PIV and PLIF Techniques

2013 ◽  
Vol 816-817 ◽  
pp. 1054-1058
Author(s):  
Ezddin Hutli ◽  
Dániel Tar ◽  
Valer Gottlasz ◽  
Gyorgy Ezsol
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Cross Section ◽  
Temperature Difference ◽  
Power Distribution ◽  
Coolant Temperature ◽  
Mixing Process ◽  
Flow Process ◽  
Flow Rates ◽  
Rod Bundle

A coolant mixing investigation in a head of a half-size model of VVER-440 fuel assembly (simulator) has been performed at KFKI. The PIV and PLIF measurements have been done under a selected list of power distribution options, flow rates and powers. The experiments were focused on obtaining a data for investigating the trends in temperature difference between the value registered by a thermocouple and that obtained using PLIF technique. The coolant temperature distribution has been measured in many positions along the coolant trajectory and where coolant flow leaves the rod bundle and in the cross section location of thermocouple, thus the dynamics of effect of mixing process is also declared. PIV and LPIF results show their ability to verify the primary results of CFD calculations.

scholarly journals Temperature Profile Measurement in Simulated Fuel Assembly Structure with Wire-Mesh Technology

2018 ◽  
Vol 2018 ◽  
pp. 1-9
Author(s):  
Hiroki Takiguchi ◽  
Masahiro Furuya ◽  
Takahiro Arai
Keyword(s):  
Temperature Distribution ◽  
Flow Field ◽  
Fuel Assembly ◽  
Temperature Profile ◽  
Wire Mesh ◽  
Calibration Line ◽  
Temperature Calibration ◽  
Water Reactor ◽  
Simulated Fuel ◽  
Assembly Structure

When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

scholarly journals Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

2017 ◽  
Vol 67 (1) ◽  
pp. 69-76
Author(s):  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Vladimír Kutiš ◽  
Gabriel Gálik
Keyword(s):  
Steady State ◽  
Temperature Distribution ◽  
Pressure Drop ◽  
Fuel Assembly ◽  
Nuclear Reactor ◽  
Software Tool ◽  
Operating Conditions ◽  
Coolant Flow ◽  
Cfd Modelling ◽  
Ansys Cfx

AbstractThe paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Norihiro Kikuchi ◽  
Yasutomo Imai ◽  
Ryuji Yoshikawa ◽  
Norihiro Doda ◽  
Masaaki Tanaka ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Flow Rate ◽  
Fast Reactor ◽  
Numerical Results ◽  
Operating Conditions ◽  
Low Flow ◽  
Thermal Hydraulics ◽  
Local Flow ◽  
Subchannel Analysis

In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.

Thermal-Hydraulic Analysis of Fuel Assembly With Inner Duct Structure of an Advanced Loop-Type Sodium-Cooled Fast Reactor Using ASFRE Code

2017 ◽  
Author(s):  
Norihiro Kikuchi ◽  
Yasutomo Imai ◽  
Ryuji Yoshikawa ◽  
Norihiro Doda ◽  
Masaaki Tanaka ◽  
...  
Keyword(s):  
Experimental Data ◽  
Temperature Distribution ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Mixing Model ◽  
Thermal Hydraulics ◽  
Local Flow ◽  
Wide Range ◽  
Resistance Model ◽  
Turbulence Mixing

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan Atomic Energy Agency, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor during the core disruptive accident. Thermal-hydraulics evaluations in FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, thermal-hydraulics in FAIDUS are investigated by using a subchannel analysis code ASFRE, which is applicable to a wire-wrapped fuel pin bundle with a distributed resistance model and a simplified turbulence mixing model. At first, the distributed resistance model was validated by comparison of pressure drop coefficients with experimental data obtained in water experiments with simulated FAs under the condition of wide-range Reynolds number. And then, the turbulence mixing model was validated by comparison of temperature distribution in the pin bundle with experimental data obtained in sodium experiments with simulated FAs. After the applicability of ASFRE to FAs was confirmed through these validations, thermal-hydraulic analyses of a FA with 271 fuel pins without the inner duct and a FAIDUS with 255 fuel pins were conducted. The obtained results indicate that no significant asymmetric temperature distribution occurs in a FAIDUS as a FA without an inner duct. In addition, the temperature distribution of FAIDUS with 255 fuel pins under the low flow rate condition tended to be the same as that of a FA with 271 fuel pins due to the local flow acceleration and the flow redistribution caused by the buoyancy force.

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