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.

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.

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.

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.

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.

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.

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.

The Gasification of Graphite Matrix in Pebble Bed Reactors

2009 ◽  
Author(s):  
Xinli Yu ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Corrosion Rate ◽  
Fuel Element ◽  
Pebble Bed ◽  
The Core ◽  
Graphite Matrix ◽  
Matrix Graphite ◽  
Element Matrix ◽  
Fuel Elements ◽  
High Temperature Gas

This paper mainly deals with the simulations of graphite matrix of the spherical fuel elements by steam in normal operating conditions. The fuel element matrix graphite was firstly simplified to an annular part in the simulations. Then the corrosions to the matrix graphite in 10 MW High Temperature Gas-cooled Reactor (HTR-10) and the High Temperature Gas-cooled Reactor—–Pebble-bed Module (HTR-PM) were investigated respectively. The results showed that the gasification of fuel element matrix graphite was uniform and mainly occurred at the bottom of the core in both of the reactors in the mean residence time of the spherical fuel elements. This was mainly caused by the designed high temperature at the bottom. The total mass gasified in HTR-PM was much greater than the HTR-10, while it did not mean much severer corrosion occurred there. As it is known the core volume of HTR-PM is much larger than the HTR-10, which will result in much greater consumed graphite even for the same corrosion rate. The steam only lost about 1 to 3 percent after flowing through the cores in both reactors for different steam conditions. The corrosion of graphite became worse when the steam concentrations increased in helium coolant. The results also indicated that the corrosion rate of fuel element matrix graphite tended to increase slightly with the prolonging of the service time.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

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.

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