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.

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

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.

Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2012 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi ◽  
Fubing Chen
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Normal Operation ◽  
Helium Temperature ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Decay Heat ◽  
The Core ◽  
Core Cooling ◽  
High Temperature Gas

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.

Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Depressurized Loss of Coolant Accidents

2008 ◽  
Author(s):  
Yanhua Zhengy ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Safety Margin ◽  
Reactor Core ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Loss Of Coolant ◽  
High Temperature Gas ◽  
Residual Heat

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.

scholarly journals The R&D of HTR-STAC Program Package: Source Term Analysis Codes for Pebble-Bed High-Temperature Gas-Cooled Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-9 ◽  
Author(s):  
Hongyu Chen ◽  
Chuan Li ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Nuclear Reactor ◽  
Source Term ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Water Reactors ◽  
Term Analysis ◽  
High Temperature Gas

Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

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