scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

scholarly journals Bypass flow in small absorber sphere channels of the high-temperature gas-cooled reactor pebble-bed module

2017 ◽  
Vol 10 (3) ◽  
pp. 128-139 ◽  
Author(s):  
Ziping Liu ◽  
Zeguang Li ◽  
Jun Sun
Keyword(s):  
High Temperature ◽  
Network Model ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Bypass Flow ◽  
Pebble Bed ◽  
Flow Network ◽  
Control Rod ◽  
High Temperature Gas

In the high-temperature gas-cooled reactor pebble-bed module, the helium bypass flow among graphite blocks cannot be ignored due to its effect on the temperature distribution as well as the maximum temperature in the reactor core. Bypass flow was previously analyzed in the discharging tube, in vertical gaps between graphite reflectors, and in control rod channels. The focus of this study is on the bypass flow that connects the small absorber sphere channels. Different from bypass flow connecting the control rod channels, there was no evident inlet or outlet flow paths into or out of the small absorber sphere channels at the top or bottom of the reactor core. Therefore, the bypass flow connecting the pebble bed with the small absorber sphere channels was mainly caused by the horizontal gaps, in which those gaps would also be irregular due to installation, thermal expansion, or irradiation of the graphite reflectors. After clarifying the resistant coefficients of those gaps by computational fluid dynamic tools, the bypass flow distribution was calculated by the flow network model including the flow in the reactor core, small absorber sphere channels, as well as horizontal gaps. Cases with various size combinations of gaps were adopted into the flow network model to test the sensitivity of bypass flow distribution to those parameters. Finally, the bypass flow in the small absorber sphere channels was concluded to be not significant in the reactor core.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

2010 ◽  
Author(s):  
H. K. Cho ◽  
B. J. Yun ◽  
I. K. Park ◽  
J. J. Jeong
Keyword(s):  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Physical Models ◽  
Injection System ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Rate Condition ◽  
The Core ◽  
Benchmark Calculation ◽  
Thermal Hydraulic Analysis

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

Research on Heat Transfer Characteristics of the Thermometric Sphere in HTR-10

2018 ◽  
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng ◽  
Xiang Fang ◽  
Xiaoyong Yang
Keyword(s):  
Heat Transfer ◽  
Thermal Equilibrium ◽  
Reactor Core ◽  
Fluid Temperature ◽  
Heat Transfer Characteristics ◽  
Pebble Bed ◽  
The Core ◽  
Temperature Measuring ◽  
Transfer Characteristics ◽  
Metal Wires

During the operation of the High Temperature Gas-cooled Reactor (HTGR), the hot-spot temperature in the reactor core must be lower than the maximum permissible temperature of the fuel elements and the materials of construction, so that the reactor kept safe. However, no fixed temperature-measuring devices can be set in a pebble-bed core. A special spherical temperature-measuring device is adopted to make sure it brings as small impact to the reactor operation as possible. There are several metal wires with different melting points inside. The graphite thermometric balls will be put onto the top of HTR-10 reactor core, and they record and reflect the highest temperature in different positions in the core when flowing in the pebble bed. Before the reactor core temperature-measuring experiment of HTR-10, we must study the heat transfer characteristics of the graphite thermometric sphere to find out the relationship of the melting conditions and the temperature in the reactor core. A 3-D model of the graphite thermometric ball is established, and CFD method is adopted to research and figure out the thermal equilibrium time and temperature difference between the metal wires in the ball and the hot fluid outside the balls. Multiple situations are simulated, and the heat transfer process of the thermometric sphere is comprehensively studied. The heat convection is certified the most important aspect. Thermal equilibrium can be achieved within 19 minutes, far shorter than the period while the spheres flowing through the core. The simulation results can also applied to derive the thermal fluid temperature backward.

scholarly journals THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

2016 ◽  
Vol 4 ◽  
pp. 22
Author(s):  
Filip Fejt
Keyword(s):  
Atmospheric Pressure ◽  
Room Temperature ◽  
Nuclear System ◽  
Hydraulic Analysis ◽  
Research Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Coolant Circulation ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

scholarly journals Preliminary thermal-hydraulic analysis of the EU-DEMO Helium-Cooled Pebble Bed fusion reactor by using the RELAP5-3D system code

2021 ◽  
Vol 162 ◽  
pp. 112111
Author(s):  
Salvatore D’Amico ◽  
Pietro Alessandro Di Maio ◽  
Xue Zhou Jin ◽  
Francisco Alberto Hernández Gonzalez ◽  
Ivo Moscato ◽  
...  
Keyword(s):  
Fusion Reactor ◽  
Hydraulic Analysis ◽  
Pebble Bed ◽  
Thermal Hydraulic Analysis ◽  
The Eu

Numerical Simulation of Core Temperature and Melting Process of IVR Core After a Severe Water Loss Accident

2017 ◽  
Author(s):  
Yiqun Liu ◽  
Xiaoying Zhang ◽  
Jingya Li ◽  
Biao Wang ◽  
Dekui Zhan ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Water Level ◽  
Water Loss ◽  
Reactor Core ◽  
Melting Process ◽  
Nucleate Boiling ◽  
Maximum Temperature ◽  
Conduction Model ◽  
Fuel Rods ◽  
The Core

After the occurrence of severe water loss accident in a PWR, the water level in the reactor core would decrease gradually, leading to heat up and melted down of the core, threatening safety of the nuclear power plant and the surrounding environment. In this paper, the 1/4 core of AP1000 PWR was adopted for study, a numerical method has been established to calculate the transient change of temperature and melting process of the core and envelope structure (boarding, basket and RPV) after the severe water loss accident. A two-dimensional conduction model with cylindrical coordinate has been used to simulate heat transfer along the radius and height direction of fuel rods and control rods in fuel assemblies. Heat transfer condition on rod surface considers nucleate boiling for rod surface below the water level, while radiative heat transfer among neighboring rods and natural convection with water vapor was considered for rod surface above the water level. Heat transfer along thickness of envelope structures were modeled with the one-dimensional conduction model. The results show that the maximum temperature of the whole reactor core does not exceed 3000K and AP1000 will not meet the melting of fuel rods with the help of RPV external water chamber cooling. The temperature values of the fuel rods and the control rod show the characteristic distribution of the two regions. At 4904s, the maximum temperature of the rod rises to 2900K, and then stabilize. The temperature of the shell is up to 2000K, the maximum temperature of the basket is to 1260K, the variation of RPV wall temperature is not obvious.

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