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

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.

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A parametric study of graphite dust deposition on high-temperature gas-cooled reactor (HTGR) steam generator tube bundles

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.09.009 ◽

2019 ◽

Vol 123 ◽

pp. 135-144 ◽

Cited By ~ 7

Author(s):

Mingzhe Wei ◽

Yiyang Zhang ◽

Xinxin Wu ◽

Libin Sun

Keyword(s):

High Temperature ◽

Steam Generator ◽

Parametric Study ◽

Dust Deposition ◽

Steam Generator Tube ◽

Tube Bundles ◽

High Temperature Gas

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

Volume 6B: Thermal-Hydraulics and Safety Analyses ◽

10.1115/icone26-82043 ◽

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.

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Analysis of Fission Products’ Release in Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel Elements Using a Modified FRESCO II Numerical Model

Science and Technology of Nuclear Installations ◽

10.1155/2021/6627789 ◽

2021 ◽

Vol 2021 ◽

pp. 1-8

Author(s):

Chao Fang ◽

Chuan Li ◽

Jianzhu Cao ◽

Ke Liu ◽

Sheng Fang

Keyword(s):

High Temperature ◽

Source Term ◽

Fission Products ◽

Release Rates ◽

Time Step ◽

The Difference ◽

Advanced Nuclear System ◽

Fuel Elements ◽

Term Analysis ◽

High Temperature Gas

The radiation safety design and emergency analysis of an advanced nuclear system highly depends on the source term analysis results. In modular high-temperature gas-cooled reactors (HTGRs), the release rates of fission products (FPs) from fuel elements are the key issue of source term analysis. The FRESCO-II code has been established as a useful tool to simulate the accumulation and transport behaviors of FPs for many years. However, it has been found that the mathematical method of this code is not comprehensive, resulting in large errors for short-lived nuclides and large time step during calculations. In this study, we used the original model of TRISO particles and spherical fuel elements and provided a new method to amend the FRESCO-II code. The results show that, for long-lived radionuclides (Cs-137), the two methods are perfectly consistent with each other, while in the case of short-lived radionuclides (Cs-138), the difference can be more than 1%. Furthermore, the matrix method is used to solve the final release rates of FPs from fuel elements. The improved analysis code can also be applied to the source term analysis of other HTGRs.

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