Design of the Essential Material Test Equipment for the Pebble Bed Effective Thermal Conductivity Measurement Experiment

2013 ◽  
Author(s):  
Cheng Ren ◽  
Xing-Tuan Yang ◽  
Cong-Xin Li ◽  
Zhi-Yong Liu ◽  
Sheng-Yao Jiang
Keyword(s):  
Thermal Conductivity ◽  
High Temperature ◽  
Effective Thermal Conductivity ◽  
Reactor Core ◽  
Test Equipment ◽  
Normal Operation ◽  
Pebble Bed ◽  
Material Test ◽  
Temperature Detector ◽  
High Temperature Gas

High Temperature Gas-cooled Reactor (HTGR) is a typical representation of Generation IV nuclear power system for its advantages like inherent safety, high efficiency, widely application as high-temperature heat source. The first two 250-MWt high temperature reactor pebble bed modules (HTR-PM) have be installing at the Shidaowan plant in Shandong Province, China, which have the cylindrical core structure with thousands of spherical fuel elements randomly packed inside. The values of the effective thermal conductivity of the pebble bed core under different temperatures are essential parameters for the design of HTGR, which are needed to analyze the maximum fuel temperature, temperature distribution and residual heat releasing ability in reactor core. For this purpose, Tsinghua University in China has proposed a full-scale heat transfer experiment to conduct comprehensive thermal transfer tests in packed pebble bed and to determine the effective thermal conductivity through the pebble bed under vacuum condition and helium environment with temperature up to 1600°C. An essential material test equipment is built in advance to provide reliable materials and technical support for the design of the final experimental device aimed at measuring the effective thermal conductivity of pebble bed type reactor core of the high temperature gas-cooled reactor. The design of the essential material test equipment is introduced in detail, including the heat element, the insulation structure, the temperature detector, cooling water system, vacuum system, hydraulic lifting system, data acquisition system and so on. Several key technologies in design are described in detail. Test temperature in the equipment was elevated up to 1600°C, which covers the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor. The construction and commissioning of the test equipment shows that the test equipment has met the design requirements and verified the feasibility of the related materials and structures.

Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor

2014 ◽  
Author(s):  
Yanhua Zheng ◽  
Fubing Chen ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Thermal Inertia ◽  
Hydraulic Calculation ◽  
Normal Operation ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Cooling Flow ◽  
Safety Features ◽  
High Temperature Gas

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs. According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.

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.

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 Bypass flow in small absorber sphere channels of the high-temperature gas-cooled reactor pebble-bed module

2017 ◽  
Vol 10 (3) ◽  
pp. 128-139 ◽  
Author(s):  
Ziping Liu ◽  
Zeguang Li ◽  
Jun Sun
Keyword(s):  
High Temperature ◽  
Network Model ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Bypass Flow ◽  
Pebble Bed ◽  
Flow Network ◽  
Control Rod ◽  
High Temperature Gas

In the high-temperature gas-cooled reactor pebble-bed module, the helium bypass flow among graphite blocks cannot be ignored due to its effect on the temperature distribution as well as the maximum temperature in the reactor core. Bypass flow was previously analyzed in the discharging tube, in vertical gaps between graphite reflectors, and in control rod channels. The focus of this study is on the bypass flow that connects the small absorber sphere channels. Different from bypass flow connecting the control rod channels, there was no evident inlet or outlet flow paths into or out of the small absorber sphere channels at the top or bottom of the reactor core. Therefore, the bypass flow connecting the pebble bed with the small absorber sphere channels was mainly caused by the horizontal gaps, in which those gaps would also be irregular due to installation, thermal expansion, or irradiation of the graphite reflectors. After clarifying the resistant coefficients of those gaps by computational fluid dynamic tools, the bypass flow distribution was calculated by the flow network model including the flow in the reactor core, small absorber sphere channels, as well as horizontal gaps. Cases with various size combinations of gaps were adopted into the flow network model to test the sensitivity of bypass flow distribution to those parameters. Finally, the bypass flow in the small absorber sphere channels was concluded to be not significant in the reactor core.

scholarly journals Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

2021 ◽  
Vol 927 (1) ◽  
pp. 012037
Author(s):  
Daddy Setyawan
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
The Core ◽  
Full Power ◽  
Double Heterogeneity ◽  
High Temperature Gas

Abstract In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.

scholarly journals Computational Fluid Dynamics Applied to Study Coolant Loss Regimes in Very High Temperature Reactors

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Uebert Gonçalves Moreira ◽  
Dany Sanchez Dominguez ◽  
Leorlen Yunier Rojas Mazaira ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
Carlos Rafael García Henández
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
High Temperature ◽  
Reactor Core ◽  
Normal Operation ◽  
Pebble Bed ◽  
Continuous Increase ◽  
High Temperature Reactors ◽  
Fuel Elements ◽  
Very High

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its   inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). We obtain the temperature profile distribution in the core for regimes where the coolant flow rate is smaller than recommended in a normal operation. In general, the temperature distributions calculated are consistent with phenomenological behavior. Even without considering the reactivity changes to reduce the reactor power or other safety mechanisms, the maximum temperatures do not exceed the recommended limits for TRISO fuel elements.

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