A Numerical Study on Graphite Dust Deposition on Steam Generator Tubes in the High-Temperature Gas-Cooled Reactor (HTGR)

2018 ◽  
Author(s):  
Mingzhe Wei ◽  
Yiyang Zhang ◽  
Zhu Fang ◽  
Xinxin Wu ◽  
Libin Sun
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Deposition Rate ◽  
Numerical Study ◽  
Fem Simulation ◽  
Dust Deposition ◽  
Lagrangian Simulation ◽  
Large Particles ◽  
Steam Generator Tubes ◽  
High Temperature Gas

Graphite dust is an important issue for the operation and maintenance of high-temperature gas-cooled reactor (HTGR), because the transport of fission product (FP) is coupled closely with graphite dusts. For instance, vapor phase FP could condense as flowing through the steam generator (SG) and deposit on the surface of graphite dusts that are either air-borne or already deposited on SG tubes. In water ingress or loss-of-coolant accidents, these dusts may re-suspend and contribute to the source term. Despite the importance of graphite dusts in HTGRs, the transport and deposition of dust particle are far from being fully understood, neither particle-fluid nor particle-wall interactions. In this work we present a numerical study on the particle transport through upper 5 layers of SG tubes. Particularly, the particle impaction process is simulated by Finite Element Method (FEM) with adhesion and dissipation specially accounted. The FEM simulation predicts the critical adhesion velocity and restitution coefficient when rebound occurs. Then we substitute the particle impaction model into Eulerian-Lagrangian simulation of flow field and extract the deposition rate statistically. The result shows that for small particles (< 5 μm), the deposition rate is controlled by the collision rate, which is mainly determined by the interaction between turbulence and thermophoresis. The particle-vortex interaction is essentially important for the distribution of particles near wall and thus influences the deposition rate. For large particles the deposition rate is more affected by the sticking efficiency, which is simultaneously controlled by both the critical adhesion velocity and normal impaction velocity. Therefore, the deposition rate first increases then decreases with particle size and reaches maximum at about 5 μm.

A Numerical Analysis on Graphite Dust Deposition and Resuspension in HTR-10 Steam Generator

2014 ◽  
Author(s):  
Wei Peng ◽  
Tian-qi Zhang ◽  
Ya-nan Zhen ◽  
Su-yuan Yu
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Load Capacity ◽  
Fission Products ◽  
Dust Deposition ◽  
Preliminary Calculation ◽  
Transfer Tube ◽  
High Load Capacity ◽  
Fuel Elements ◽  
High Temperature Gas

The behavior of graphite dust is important to the safety analysis of High-Temperature Gas-cooled Reactor (HTGR). The fission products released by fuel elements would enter the primary loop and combine with dust, resulting in that the dust has a high load capacity of cesium, strontium, iodine and tritium. It would bring difficulty and inconvenience to the maintenance and repair of steam generator. Therefore, the behavior of graphite dust in the steam generator is essential to the safety of High Temperature Gas-cooled Reactors. The present study focused on the deposition and resuspension of graphite dust in steam generator of HTR by numerical method. The results show that the graphite dust in steam generator deposits on the surface of heat transfer tube through turbulent deposition, thermophoretic deposition, and other depositional mechanisms, of which thermophoretic deposition is the main mechanism for the particles with the diameter of 2.2μm in the present study. The preliminary calculation result shows that about 6760mg/m2 of graphite dust tends to load on the tube surface.

Numerical Simulation of Accident Scenario in High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors

2012 ◽  
Author(s):  
Geoffrey J. Peter
Keyword(s):  
Numerical Simulation ◽  
High Temperature ◽  
Boundary Conditions ◽  
Nuclear Reactors ◽  
Numerical Study ◽  
Packed Bed ◽  
Pebble Bed ◽  
Accident Scenario ◽  
Set Up ◽  
High Temperature Gas

The accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors considered for the next generation of Advanced High Temperature Reactors (AHTR), for nuclear power production, and for high-temperature hydrogen production using nuclear reactors to reduce the carbon footprint is examined in this paper. Blockages can cause local variations in flow and heat transfer that may lead to hot spots within the bed that could compromise reactor safety. Therefore, it is important to know the void fraction distribution and the interstitial velocity field in the packed bed. The blockage for this numerical study simulated a region with significantly lower void than that in the rest of the bed. Finite difference technique solved the simplified continuity, momentum, and energy equations. Any meaningful outcome of the solution depended largely upon the validity of the boundary conditions. Among them, the inlet and outlet velocity profiles required special attention. Thus, a close approximation to these profiles obtained from an experimental set-up established the boundary conditions. This paper presents the development of the elliptic-partial differential equation for a bed of pebbles, and the solution procedure. The paper also discusses velocity and temperature profiles obtained from both numerical and experimental setup, with and without effect of blockage. In addition, the paper compares the results obtained from the experimental set-up with numerical simulation using a commercially available code that uses finite element techniques.

The Water Ingress Mass Analysis on Steam Generator Heat-Exchange Tube Rupture Accident of High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Yan Wang ◽  
Yanhua Zheng ◽  
Fu Li ◽  
Lei Shi ◽  
Zhiwei Zhou
Keyword(s):  
High Temperature ◽  
Heat Exchange ◽  
Steam Generator ◽  
Mechanical Load ◽  
Radioactive Isotopes ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Water Ingress ◽  
Heat Exchange Tube ◽  
High Temperature Gas

The module high temperature gas-cooled reactor (HTGR) is an advanced reactor with high safety level. The steam generator heat-exchange tube rupture (SGTR) accident (or water ingress accident) is an important and particular accident which will result in water ingress to the primary circuit of reactor. Water ingress may, in turn, result in chemical reaction of graphite fuel and structure with water, causing release of radioactive isotopes and generation of explosive gaseous in large quantity. The analysis of SGTR is significant for verifying the inherent safety characteristics of HTGR. One of the key factors is to estimate the amount of water ingress mass which is used to evaluate the severity of the accident consequence. The 200MWe high temperature gas-cooled reactor, which is designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected as an example to analyze. The accident scenarios of double-ended rupture of both single and two heat-exchange tubes at the inlet and outlet of steam generator are simulated respectively by RETRAN-02. The results show that the amount of water ingress mass is related to the break location, the number of ruptured tubes (or the break size). The greater the number of ruptured tubes or the break size, the larger the amount of water ingress mass. It is important to design the draining pipe line with reasonable diameter, which should be optimized based on economy and safety considerations for preventing large water ingress to the reactor primary circuit, restricting the change rate of mechanical load on SG, and reducing the radioactive isotopes release to the secondary circuit.

Numerical Study of the Deposition of Radioactive Nucleoids (Aerosol) in PWR Steam Generator

2021 ◽  
Author(s):  
Muhammad Aadil ◽  
Rab Nawaz ◽  
Ajmal Shah ◽  
Kamran Rasheed Qureshi
Keyword(s):  
Steam Generator ◽  
Numerical Study ◽  
Deposition Efficiency ◽  
Decontamination Factor ◽  
Radioactive Nuclide ◽  
Discrete Phase Model ◽  
Velocity Magnitude ◽  
Discrete Phase ◽  
Steam Generator Tubes ◽  
Secondary Side

Abstract This research presents numerical study of deposition efficiency and decontamination factor of radioactive nuclide in steam generator tubes of a typical 325 MWe PWR. To find out the deposition of aerosol, the discrete phase model (DPM) has been used. The flow has been characterized as compressible, adiabatic, turbulent and wall bounded. When steam generator tube gets ruptured, the radioactive nuclides can escape from primary side and create a radioactive field in the secondary side. This can be harmful for the personnel working at the plant. Therefore, in order to ensure the safety of the plant and personnel, it is important to study the particles deposition on the wall of steam generator tubes. In the present study, a CFD methodology has been first developed and validated with the published results. After methodology validation, it has been applied to the U-tube of a typical PWR steam generator. It has been observed that due to the action of centrifugal force near the bent, the velocity magnitude is high towards the inner wall and the flow separates at the bent entrance. Furthermore, the flow inside the tube is rotational with vortices throughout the domain due to the presence of the bent. Finally, the deposition efficiency and decontamination factor have been calculated and it has been observed that both increase with the increase in particle size due to inertial effects.

Simulation of a Once-Through, Helical-Coiled Steam Generator of High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Jinliang Ye ◽  
Yangping Zhou ◽  
Xiaoming Chen ◽  
Yuanle Ma ◽  
Fu Li ◽  
...  
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Static Model ◽  
Mass Energy ◽  
Design Data ◽  
The Difference ◽  
Steady State Simulation ◽  
Energy And Momentum ◽  
Blast Program ◽  
High Temperature Gas

A quasi-static model of a helical coiled Once-Through Steam Generator of High Temperature gas-cooled Reactor-Pebble Bed Module (HTR-PM) is developed based on the fundamental conservation of fluid mass, energy and momentum. The steam generator is handled with single tube concept and is divided into three regions as subcooled region, boiling region and superheated region. The equations are solved by Rung-Kutta method. The steady-state simulation results agree well with the design data. Furthermore, the results are compared with the results gotten from THERMIX/BLAST program, and the difference between them is small which shows the model and the methodology are reasonable.

Sign in / Sign up

Export Citation Format

Share Document