Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

2011 ◽  
Vol 241 (8) ◽  
pp. 3076-3085 ◽  
Author(s):  
Min-Hwan Kim ◽  
Hong-Sik Lim
Keyword(s):  
Hot Spot ◽  
Bypass Flow ◽  
The Core

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.

VHTR Core Preliminary Analysis Using NEPHTIS3 / CAST3M Coupled Modelling

2008 ◽  
Author(s):  
Fre´de´ric Damian
Keyword(s):  
Fuel Element ◽  
System Analysis ◽  
Hot Spot ◽  
Operating Conditions ◽  
Maximum Temperature ◽  
Design Parameters ◽  
Block Type ◽  
Fuel Temperature ◽  
Coupled Modelling ◽  
The Core

Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3, 4]. In the framework of the European RAPHAEL project, this paper presents the results of the preliminary investigations carried out on the VHTR design. These studies aimed at understanding the physical aspects of the annular core and to identify the limits of a standard block type VHTR with regard to a degradation of its passive safety features. Analysis was performed considering various geometrical scales: fuel cell and fuel column located at the core hot spot, 2D and 3D core configurations including the coupling between neutronic and thermal-hydraulic. From the thermal analysis performed at the core hot spot, the capability to reduce the maximum fuel temperature by modifying the design parameters such as the fuel compact and the fuel block geometry was assessed. The best performances are obtained for an annular fuel compact geometry with coolant flowing inside and outside the fuel compact (ΔT > 50°C). The reliability of such design option should however be addressed with respect to its performance during the LOFC transient (the residual decay heat will be evacuated by radiation during the transient instead of conduction through graphite). As far as the fuel element geometry is concerned, a gain of approximately 50°C can be achieved by making limited changes on the fuel compact distribution in the prismatic block: reduction of the number of fuel compact in the outer ring of the fuel element where the average ratio between coolant channels and fuel compact is smaller. On the other hand, the adopted modifications should also be evaluated with respect to the maximum temperature gradient achieved in the fuel (amoeba effect). In the end, calculations performed on the full core configuration taking into account the thermal feedback showed that the radial positioning of the fuel elements allows to reduce significantly the power peaking factor and the maximum fuel temperature. The gain on the fuel temperature, which varies during the core irradiation, is in the range 100 – 150°C. Several modifications such as the increase of the bypass fraction and the replacement of a part of the graphite reflector by material with better thermal properties were also addressed in this paper.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injector: Part 14—Cooling Characteristics of the Particulate Core Material Debris Accumulated in Horizontal Narrow Gap

2006 ◽  
Author(s):  
Y. Hirao ◽  
G. Su ◽  
K. Sugiyama ◽  
T. Narabayashi ◽  
M. Mori ◽  
...  
Keyword(s):  
Heat Supply ◽  
Nuclear Power ◽  
High Efficiency ◽  
Hot Spot ◽  
Research Work ◽  
Lower Surface ◽  
Injection System ◽  
Core Material ◽  
Metal Debris ◽  
The Core

When LOCA occurs in proposed nuclear reactor systems, the coolability of the core would be kept by the SI core injection system and therefore the probability of the core meltdown is negligible small. In this research work, we make it clear that the coolability of the RPV bottom is secured even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. In this report, we examined experimentally the coolability of the RPV bottom that a Zircaloy-based loose debris layer is deposited on. We set up a heat supply section made by SUS304 on the loose debris layer and measured the heat flux released into the loose debris bed and the temperature at the lower surface of the heat supply section. In addition, we measured the temperature distribution at the bottom of the loose debris bed. It became clear in this study that the coolability depends on the amount of coolant supplied, and the hot spot would not occur when coolant is supplied. Even if a hotspot should occur in the oxidization of loose metal debris accompanied with rapid heat generation. It is found that when a small amount of coolant can be supplied, it disappears because of a high capillary force of oxidized loose debris. So it is confirmed that the soundness of RPV is basically maintained.

Dynamic Trust Model of Internet-Ware Based on Task Classification

2012 ◽  
Vol 155-156 ◽  
pp. 221-225 ◽  
Author(s):  
Gui Sheng Yin ◽  
Shu Yin Wang
Keyword(s):  
Information Security ◽  
Trusted Computing ◽  
Hot Spot ◽  
Trust Model ◽  
Information Storage ◽  
Security Design ◽  
The Core ◽  
Task Classification ◽  
Present Information ◽  
Dynamic Trust

with the Internet popularity and the widely application of information storage technology, people pay more and more attention on the information security. Trust model is the hot spot of present information security field. It is connected with the safety interaction and operation of the entire system. Internet has the characteristic of openness, anonymity, and autonomy. All these extrude the security problems. How to build the safety and reliable trust model is the most reliable approach to solve the network security problems. Trusted computing is the core center of trust model. Through the trusted computing, we will get the perfect security design.

Characterization and Decontamination of Contaminated Structures Using a Profiling Technology

2003 ◽  
Author(s):  
S. Aggarwal ◽  
G. Charters ◽  
D. Thacker
Keyword(s):  
Hot Spot ◽  
Cross Contamination ◽  
Chemical Contamination ◽  
Considerable Time ◽  
Activity Profile ◽  
Concrete Core ◽  
Porous Concrete ◽  
The Core ◽  
Waste Segregation ◽  
Contamination Problem

Certain radioisotopes (tritium, radium, cobalt, plutonium, and cesium) can penetrate porous concrete and contaminate the concrete well below the easily measured surface. Certain radioisotopes can penetrate concrete and contaminate the concrete well below the surface. The challenge is to determine the extent of the contamination problem and the magnitude of the problem in a real-time. Currently, concrete core bores are shipped to certified laboratories where the concrete residue is run through a battery of tests to determine the contaminants. The existing core boring operation volatilizes some of the contaminants (like tritium) and oftentimes cross-contaminates the area around the core bore site. The volatilization of the contaminants can lead to airborne problems in the immediate vicinity of the core bore. Cross-contamination can increase the contamination area and thereby increase the amount of waste generated. The goal is to avoid those field activities that could cause this type of release. The concrete profiling technology, TRUPROSM in conjunction with portable radiometric instrumentation produces a profile of radiological or chemical contamination through the material being studied. The data quality, quantity, and representativeness may be used to produce an activity profile from the hot spot surface into the material being sampled. This activity profile may then be expanded to ultimately characterize the facility and expedite waste segregation and facility closure at a reduced cost and risk. Performing a volumetric concrete or metal characterization safer and faster (without lab intervention) is the objective of this characterization technology. This way of determining contamination can save considerable time and money.

Effects of Bypass Flow on the LOFA Transient Computations in a VHTR

2013 ◽  
Author(s):  
Yu-Hsin Tung ◽  
Richard W. Johnson ◽  
Yuh-Ming Ferng ◽  
Ching-Chang Chieng
Keyword(s):  
Nuclear Power ◽  
Natural Circulation ◽  
Cfd Model ◽  
Bypass Flow ◽  
The Core ◽  
Active Core ◽  
Very High Temperature Reactor ◽  
High Temperature Reactor ◽  
Very High ◽  
Helium Coolant

The prismatic gas-cooled very high temperature reactor (VHTR) is one possible option for the generation IV nuclear power plant. The prismatic VHTR basically involves stacks of hexagonal graphite blocks that are drilled to accept cylindrical fuel compacts and provide coolant channels for the helium coolant. Between the hexagonal blocks, there are gaps, which allow the coolant flow to bypass the coolant channels. The gaps are not intentionally designed to occur in the core, but are present because of tolerances in machining the blocks, imperfect installation and expansion and shrinkage from heating and irradiation. Based on previous studies of a loss of flow accident (LOFA), the cooling provided by flow in the bypass gaps has a significant effect on the nature and strength of the attendant natural circulation. One of the mechanisms that occurs after a LOFA for the transport of heat out of the core is by the natural convection of the coolant. It is of interest to know if there are problems for the core associated with the natural circulation and what is the role played by the bypass flow in such an event. The distribution of heat generation and the separation of the partial columns included in the CFD model of the heated core have a strong effect on the natural circulation. In the present paper, a 1/12 symmetric section of the active core is considered for the CFD model. Two regions of the 1/12 section are employed to perform the LOFA transient calculations. Several scenarios are investigated including with and without the bypass gap in the model. The present study also reports the effects of bypass flow on the natural circulation with time for these cases.

DEM-CFD Study on Fluid Flow Through Radial Layered Composite Packed Beds

2018 ◽  
Author(s):  
Zehua Guo ◽  
Zhongning Sun ◽  
Nan Zhang ◽  
Ming Ding ◽  
Haozhi Bian ◽  
...  
Keyword(s):  
Fluid Flow ◽  
Packed Bed ◽  
Layered Composite ◽  
Wall Region ◽  
Bypass Flow ◽  
Porosity Distribution ◽  
Container Wall ◽  
The Core ◽  
Core Part ◽  
Near Wall

The radial porosity generally have a higher value at the container wall than that in the core part. Consequently, the fluid flows are mal-distributed in packed bed with significant bypass flow at the wall, which lowers the convective heat transfer performance inside the packed bed. To overcome this drawback of packed bed, we developed an effective way to construct the radial layered composite packed bed, which can easily realize placing small particle at the near wall region and large spheres in the core region. Therefore, smaller pores forms close to the container wall and larger channels presented in the core part. This could result in a much homogenous radial porosity distribution, which is benefit to restrain the bypass flow near the wall. In present paper, the packing procedure is simulated by the discrete element method (DEM). Radial layered composite packed bed and traditional packed bed with uniform spheres are compared on radial porosity distribution. By altering the size of spheres ratios in the near wall region different radial layered packings are also generated and compared. Then, the geometries of these packed bed are imported into the computational fluid dynamics (CFD) simulation. The fluid flow inside the packed bed is investigated. It finds that the radial layered packed bed has a lower pressure drop with the ordered packing structure. And a much homogenous fluid flow distribution is obtained than the traditional, which is benefit for the heat removal inside the packed bed. This would be useful for the optimum design of packed bed in industry applications.

CFD Analysis on the Influence of Coolant Distribution Blocks on the Core Hot Spot Temperature in a Prismatic Very High Temperature Reactor

2011 ◽  
Author(s):  
Min-Hwan Kim ◽  
Nam-il Tak ◽  
Jae Man Noh ◽  
Goon-Cherl Park
Keyword(s):  
High Temperature ◽  
Hot Spot ◽  
Maximum Velocity ◽  
Unit Cell ◽  
Normal Operation ◽  
The Core ◽  
Very High Temperature Reactor ◽  
Velocity Deviation ◽  
High Temperature Reactor ◽  
Very High

Two design options of core distribution block (CDB) for a cooled-vessel design in the Very High Temperature Reactor (VHTR) were developed and the influence on the core hot spot was investigated by the commercial computational fluid dynamics (CFD) code, CFX-11. Isothermal CFD analyses were performed to estimate the coolant flow variation at the inlet of the coolant channel. The results predicted about 5% of the maximum velocity deviation when applying the core pressure drop of NHDD PMR200. A unit-cell CFD model was used to access the effect of the velocity deviation on the core hot spot. The unit-cell analyses were carried out for the velocity deviation of 0%, 5%, and 10%. Not only a constant power but also a local maximum power profile was considered. According to the results, the maximum fuel temperature was increased by about 30°C for the velocity deviation of 10% but still below the normal operation limit of 1250°C.

scholarly journals Analysis of the Flow Distribution in a Particle Bed Reactor for Nuclear Thermal Propulsion

Energies ◽  
2019 ◽  
Vol 12 (19) ◽  
pp. 3590 ◽  
Author(s):  
Ji ◽  
Liu ◽  
Sun ◽  
Shi
Keyword(s):  
Hot Spot ◽  
Flow Distribution ◽  
Pressure Profile ◽  
Gas Channel ◽  
The Core ◽  
Hydraulic Design ◽  
Distribution Process ◽  
Pros And Cons ◽  
Particle Bed ◽  
Chemical Propulsion

Nuclear thermal propulsion (NTP) is regarded as the preferred option for the upcoming crewed interstellar exploration due to its excellent performance compared to the current most advanced chemical propulsion systems. Over the past several decades, many novel concepts have been proposed, among which the particle bed reactor (PBR) is the most efficient, compact, and lightweight method. Its unique features, such as the extremely high power density and the radial flow path of coolant in the fuel region, introduce many challengeable issues to the thermal hydraulic design of PBR, with the flow distribution being representative. In this work, the flow distribution process within the core is analyzed based on the understanding of the axial pressure profile in a dummy PBR. A “flow shift” phenomenon leading to the hot spot in the core is introduced first, and three methods, i.e., decreasing the pressure drop within the hot gas channel, increasing the flow resistance on the cold frit or hot frit, and changing the flow pattern from “Z” to “U”, are proposed to reduce the “flow shift” and the consequent temperature mal-distribution. The pros and cons of using cold frit or hot frit to distribute the coolant are also discussed. Finally, by using three numerical examples, these analyses are demonstrated. The findings here may provide technical support for PBR design.

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