scholarly journals Influence of void fraction in the power distribution for a GE-12 fuel assembly

2017 ◽  
Vol 792 ◽  
pp. 012051
Author(s):  
S J Castillo ◽  
G A Vargas ◽  
E del Valle Gallegos
Keyword(s):  
Fuel Assembly ◽  
Void Fraction ◽  
Power Distribution

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

scholarly journals Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

2021 ◽  
Vol 31 (1) ◽  
pp. 60-71
Author(s):  
Leonardo Acosta Martínez ◽  
Carlos Rafael García Hernández ◽  
Jesus Rosales García ◽  
Annie Ortiz Puentes
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Capital Investment ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Geometric Model ◽  
Modular Construction ◽  
Construction Time ◽  
Hydraulic Simulation ◽  
In Series

One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

Measurement of Three-Dimensional Time-Averaged Void Fraction Distribution in Rod Bundle in Air-Water System by X-Ray CT Technique

2013 ◽  
Author(s):  
Kenichi Katono ◽  
Jun Nukaga ◽  
Takuji Nagayoshi ◽  
Kenichi Yasuda
Keyword(s):  
Fuel Assembly ◽  
Void Fraction ◽  
Two Phase Flow ◽  
Three Dimensional ◽  
Ct Images ◽  
Phase Flow ◽  
Two Phase ◽  
X Ray ◽  
Void Fraction Distribution ◽  
Fraction Distribution

We have been developing a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray CT (computed tomography) system to understand two-phase flow behavior inside a fuel assembly for BWR (boiling water reactor) thermal hydraulic conditions of 7.2 MPa and 288 °C. Unlike CT images of a normal standstill object, we can obtain 3D CT images that are reconstructed from time-averaged X-ray projection data of the intermittent two-phase flow. We measured the 3D void fraction distribution in a vertical square (5 × 5) rod array that simulated a BWR fuel assembly in the air-water test. From the 3D time-averaged CT images, we confirmed that the void fraction at the center part of the channel box was higher than that near the channel box wall, and the local void fraction at the central region of a subchannel was higher than that at the gap region of the subchannel. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer in a void fraction range from 0.05 to 0.40 showed satisfactory agreement within a difference of 0.03.

scholarly journals Void Fraction Correlation for Rod Bundle. Void Fraction Correlation Based on Void Data of BWR Fuel Assembly Measured by X-Ray CT Scanner.

1995 ◽  
Vol 37 (2) ◽  
pp. 133-143 ◽  
Author(s):  
Takao ISHIZUKA ◽  
Akira INOUE ◽  
Tatuo KUROSU ◽  
Toshimasa AOKI ◽  
Masanobu FUTAKUCHI ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Void Fraction ◽  
Ct Scanner ◽  
X Ray ◽  
Rod Bundle

Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics

2010 ◽  
Author(s):  
Zhenyang Li ◽  
Tao Zhou ◽  
Canhui Sun ◽  
Xiaozhuang Liu
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Power Distribution ◽  
Neutron Cross Section ◽  
Physical Characteristics ◽  
Supercritical Pressure ◽  
Light Water ◽  
Assembly Design ◽  
Temperature Distributions ◽  
Parallel Channels

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

scholarly journals LATERAL COOLANT FLOW BETWEEN FUEL ASSEMBLIES IN MIXED CORES

2018 ◽  
Vol 14 ◽  
pp. 1
Author(s):  
Vojtech Caha ◽  
Jiří Čížek
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Coolant Flow ◽  
Coolant Temperature ◽  
Hydraulic Characteristics ◽  
Axial Coordinate ◽  
Comparison Of Results ◽  
Ansys Cfx ◽  
Fuel Assemblies ◽  
Fuel Reload

This paper presents the results of an analysis of lateral coolant flow between adjacent fuel assemblies with non-identical spacing grids in a mixed core consisting of TVSA-T mod.1 and TVSA-T mod.2 fuel assemblies. The calculation was carried out using modified subchannel code SUBCAL which allows to calculate 3D thermo-hydraulic characteristics of the coolant flow in the full three fuel assemblies model. This full three fuel assemblies model was created in two variants. The first variant consisted of three hydraulically identical fuel assemblies TVSA-T mod.1, whereas the second variant consisted of two fuel assemblies TVSA-T mod.1 and one fuel assembly TVSA-T mod.2 which mainly differ in types, number and axial coordinate of spacing grids and also in diameter of guide tubes. The influence of mixed core to lateral coolant flow and hence coolant temperature was obtained by comparing these two variants. The power distribution was taken from presumed mixed core fuel reload calculated by macro-code ANDREA. Finally there were also provided a comparison of results achieved by subchannel analysis approach with calculation of similar problem using CFD code ANSYS CFX by TVEL, the fuel supplier.

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