Characteristics of Droplet Formation During Spout Process of Water Ingress Accident

2014 ◽  
Author(s):  
Yanhua Zheng ◽  
Jun Sun ◽  
Zhipeng Chen ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Flow Line ◽  
Reactor Core ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Flash Evaporation ◽  
Water Ingress ◽  
Multi Phase Flow ◽  
Different Diameters ◽  
High Temperature Gas

Water ingress is one of the peculiar accidents of high temperature gas-cooled reactors (HTR). Since the pressure of the secondary circuit is higher than that of the primary circuit, when the break of heating tubes of the SG occurs, liquid water and steam will enter the primary circuit from the secondary side, and then enter the reactor core together with the helium coolant. This will result in introduction of a certain positive reactivity, rising of the primary pressure and corrosion of the in-core fuel elements and graphite components. Therefore, water ingress is a complex physical processes involving water spout, flash evaporation, multi-phase flow, heat transfer and other phenomena, and in-depth research of this accident will benefit further understanding of the characteristics of HTRs. Based on the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), assuming that the break of steam generator (SG) heating tube can be considered as a round pressure nozzle, the characteristics of the flow line at breaks with different diameters, as well as the property of the droplet spectrum are studied. According to the analysis of droplet properties of flow and evaporation in the primary circuit, the preliminary result shows that under the design basis accidents of water ingress, the water will enters the reactor core in the form of steam. This analysis result can provide a basis for the further research of water ingress of HTR in the future.

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.

scholarly journals Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach

2019 ◽  
Vol 2019 ◽  
pp. 1-12
Author(s):  
Chuan Li ◽  
Wenqian Li ◽  
Lifeng Sun ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Point Of View ◽  
Normal Operation ◽  
Primary Circuit ◽  
Chemical Thermodynamics ◽  
Chemical Forms ◽  
Operational Conditions ◽  
Water Ingress ◽  
High Temperature Gas

The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

Graphite aerosol release to the containment in a water ingress accident of high temperature gas-cooled reactor (HTGR)

2019 ◽  
Vol 342 ◽  
pp. 170-175 ◽  
Author(s):  
Mingzhe Wei ◽  
Yiyang Zhang ◽  
Zhu Fang ◽  
Xinxin Wu ◽  
Libin Sun
Keyword(s):  
High Temperature ◽  
Water Ingress ◽  
High Temperature Gas

scholarly journals Design of the Online Gross γ Monitoring Instrument at the Exit of the Helium Purification System in HTR-PM

2018 ◽  
Vol 2018 ◽  
pp. 1-12 ◽  
Author(s):  
Mengqi Lou ◽  
Liguo Zhang ◽  
Feng Xie ◽  
Jianzhu Cao ◽  
Jiejuan Tong ◽  
...  
Keyword(s):  
High Temperature ◽  
Detection Efficiency ◽  
Primary Circuit ◽  
Source Terms ◽  
Primary Loop ◽  
Monitoring Instrument ◽  
Purification System ◽  
Design Values ◽  
Under Construction ◽  
High Temperature Gas

After the successful construction and operation experience of the 10 MW high-temperature gas-cooled reactor (HTR-10), a high-temperature gas-cooled pebble-bed modular (HTR-PM) demonstration plant is under construction in Shidao Bay, Rongcheng City, Shandong province, China. An online gross γ monitoring instrument has been designed and placed at the exit of the helium purification system (HPS) of HTR-PM and is used to detect the activity concentration in the primary circuit after purification. The source terms in the primary loop of HTR-PM and the helium purification process were described. The detailed configuration of the gross γ monitoring instrument was presented in detail. The Monte Carlo method was used to simulate the detection efficiency of the monitoring system. Since the actual source terms in the primary loop of HTR-PM may be different than the current design values, a sensitivity analysis of the detection efficiency was implemented based on different relative proportions of the nuclides. The accuracy and resolution of the NaI(Tl) detector were discussed as well.

Steady-State Neutrbnic Investigations to the Accident of Water Ingress in Systems with Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel

1987 ◽  
Vol 97 (1) ◽  
pp. 72-88 ◽  
Author(s):  
F. Schürrer ◽  
W. Ninaus ◽  
K. Oswald ◽  
R. Rabitsch ◽  
Hj. Müller ◽  
...  
Keyword(s):  
Steady State ◽  
High Temperature ◽  
Reactor Fuel ◽  
Pebble Bed ◽  
Water Ingress ◽  
High Temperature Gas

Study on Heating Process of Dehumidifying Experiment in HTGR

2020 ◽  
Author(s):  
Kaiyue Shen ◽  
Wei Zheng ◽  
Shengchao Ma ◽  
Huaqiang Yin ◽  
Xuedong He ◽  
...  
Keyword(s):  
High Temperature ◽  
Kinetic Energy ◽  
Thermal Energy ◽  
Reactor Core ◽  
Primary Circuit ◽  
Heating Process ◽  
Transfer Model ◽  
Thermal Source ◽  
Flow And Heat Transfer ◽  
Selection Of

Abstract A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.

Thermal Modeling of the HTR-10 Using the RELAP5-3D Code

2014 ◽  
Author(s):  
Maria Elizabeth Scari ◽  
Antonella Lombardi Costa ◽  
Claubia Pereira ◽  
Clarysson Alberto Mello da Silva ◽  
Maria Auxiliadora Fortini Veloso
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Initial Study ◽  
Industrial Applications ◽  
Pebble Bed ◽  
State Behavior ◽  
Energy Agency ◽  
Helium Gas ◽  
New Energy ◽  
High Temperature Gas

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

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