High Temperature Gas-cooled Reactor Pebble-bed Module Research Articles

Installation of the graphite internals in HTR-PM

Internals are one of key equipment in the nuclear islands of High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM). Their structural integration is very important to the nuclear safety. Different from the internals in pressurized water reactors, a large number of graphite and carbon components are used in the internals of HTR-PM. These components are connected by dowels and keys with gaps between them to compensate the deformation caused by thermal and irradiation stresses. High installation accuracy of each component is required for the important safety and function related devices installed inside the internals, such as control rod and fuel handling system. Due to the cumulative assembly error of the components and the deformation of the supporting structures, the installation of the graphite and carbon internals is very challenging. Detailed requirements for the examination and installation of the graphite component have been developed and continually optimized for both trial assembly and assembly in the reactor. Research was also carried out for measurement techniques, lifting tools, examination and rapid assessments. The installation of graphite and carbon internals has been successfully completed for both modules. Some major installation results, including the position of the channels and levelness of the graphite components as well as the considerations of their acceptance are introduced in the present paper. The lessons-learned and experience can be applied to the installation of internals in HTRs and other reactors using graphite components, such as molten salt reactor.

Discussion on the accident behavior and accident management of the HTGR

High Temperature Gas-cooled Reactor (HTGR), which plays an important role in the worldwide development of Generation-IV nuclear energy technology, is well-known for the high safety, high efficiency and process heat application. A commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) has been designed and is now under construction in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in the end of 2020.In this paper, the reactor behaviors of the HTR-PM during the typical design basis accidents (DBAs) and beyond design basis accidents (BDBAs) have been introduced. In the DBAs, the maximum fuel temperature will never exceed its design limitation of 1620 °C, below which under the design burn-up the “Tristructural-isotropic” (TRISO) coated particles can effectively retain their fission-product inventory. Even in the typical BDBAs with extremely low probability, there is enough time, e.g. several days, to adopt appropriate measures to mitigate the consequence, so that the large release of the radioactive materials would not happen.Accident management is important for the nuclear power plant. Besides, After the Fukushima Dai-ichi nuclear accident, in the worldwide, people are more concerned about the severe accident management of the nuclear power plant, and some standards and guidelines are issued. Based on the accident analyses of the HTR-PM, its accident management is studied and discussed. The designers believe that the accident management procedures can be simplified and no offsite emergency measures are needed for the HTR-PM due to the inherent safety design. Compared to the other types of the nuclear power plant, e.g. the Pressurized-Water Reactor (PWR) power plant, the Severe Accident Management Guideline (SAMG) can be simplified or even unnecessary for the HTR-PM.Above work also can provide reference for the further study on the safety regulations, standards or guidelines for the design, operation and accident management of the HTGR.

Frequency domain analysis of two-phase flow instabilities in a helical tube once through steam generator for HTGR

Two-phase flow instabilities will disturb the control systems and cause mechanical vibrations and failure of components, which should be avoided. Two-phase flow instabilities in a helical tube Once Through Steam Generator (OTSG) for High Temperature Gas-cooled Reactor (HTGR) were investigated. First, a linear frequency domain model of the helical tube OTSG was developed. The convective evaporation process was divided into single phase water region, two-phase flow boiling region and superheated steam region. Incompressible assumptions were adopted in the single phase water and superheated steam regions. Homogeneous flow model was used in the two-phase region. Small perturbation linearization and Laplace transform were performed to analyze the governing equations in the three regions. The corresponding frequency domain transfer functions describing the flow pulsation and pressure drop pulsation of the two-phase flow systems were derived. By solving the Nyquist curve of the transfer function with matlab software, the two-phase flow instabilities and the stable boundary of the evaporation process in helical tubes can be determined. The new model in this paper was verified using previous experimental results. The two-phase flow stability in the helical tube OTSG of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) was analyzed using the developed frequency domain model at the design power levels. The stability boundary of the two-phase flow and effects of various parameters on flow instabilities were also investigated. Then an accumulator was added to this linear frequency domain model, which was used to study the effect of compressible volume or accumulator on DWO and PDO. Using this model, the effects of throttle valves at different locations on the stability of the system with an accumulator were also studied. The results show that the helical tube OTSG for HTR-PM can operate stably under the designed working conditions with enough margin.

Thermal analysis of HTGR helical tube once through steam generators using 1D and 2D methods

The working temperature of High Temperature Gas-cooled Reactor (HTGR) steam generators is much higher than that of Pressurized Water Reactor (PWR) steam generators. Large temperature non-uniformities will make the HTGR steam generator deviate from the designed working temperatures, which should be avoided. The temperature distribution and thermal deviation should be analyzed. One-dimensional (1D) and two-dimensional (2D) methods for analyzing the thermal hydraulic performance of an HTGR steam generator were developed. In the 1D method, the tube and shell side governing equations were solved. They were coupled through the energy source terms. In the 2D method, simulation of the shell side helium cross flow over tube bundles was based on the two-dimensional CFD code Ansys Fluent. The tube side water flow was coupled to the 2D simulation of the cross flow over tube bundles through the heat transfer between the primary and secondary sides. The 1D and 2D methods were validated using the measured data of the Thorium High Temperature Reactor (THTR) steam generators and the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) test steam generator. The 1D method can give the overall thermal performance of the steam generators. The 2D method can predict temperature non-uniformities in the tube bundles. The overall thermal performance of the steam generator was simulated. The thermal non-uniformities or thermal deviations caused by geometrical and thermal hydraulic deviations were analyzed. Radiation heat transfer contributes to 0.9% of the total thermal power. A one percent change of the secondary side water flow rate of a certain layer of tubes will result in a 5–6 °C variation of the outlet steam temperature of this layer of tubes. A helical diameter deviation of 1 mm will result in a maximum outlet steam temperature variance of 6–7 °C. An inlet helium temperature profile with a 10 °C variance will only result in a variation of 2 °C in the outlet steam temperature.

Supercritical steam generator design and thermal analysis based on HTR-PM

The high temperature gas-cooled reactor pebble-bed module (HTR-PM) being built by the Institute of Nuclear and New Energy Technology of Tsinghua University in China is a typical Generation IV nuclear power system. The subcritical steam generators in HTR-PM can be replaced by the supercritical steam generators to work with the current reactors and a supercritical steam turbine to improve the thermal efficiency, which is being presented in the third technical phase of the HTR-PM program in China. The supercritical high temperature gas-cooled reactor pebble-bed module system scheme is proposed based on HTR-PM. A supercritical steam generator with helically coiled tubes is designed. Parameters of the geometry and the design conditions are listed in detail. The heat transfer performance of the supercritical steam generator is analyzed, and the heat flux, the temperature variation and the heat transfer coefficient along the tube are shown. The effects of the transverse and longitudinal pitches between the tubes in the supercritical steam generator are very important. The tube length, tube bundle height, and the deviation of the tube length in various layers are compared with various transverse and longitudinal pitches and the optimal relative transverse and longitudinal pitches are illustrated.

Effects of ion irradiation and oxidation on point defects in IG-110 nuclear grade graphite

Nuclear grade graphite is a kind of key material in the high temperature gas-cooled reactor pebble-bed module (HTR-PM), where nuclear grade graphite acts as the fuel element matrix material, structural material and neutron reflector. In the reactor, the service environment of nuclear grade graphite suffers high temperature and strong neutron radiation. Both neutron radiation and the oxidation by the oxidizing impurities in HTGR coolant can cause the structure to damage and the properties to deteriorate. Therefore, it is of great significance to study the evolution of defects in nuclear grade graphite for improving the reactor safety. The effects of ion irradiation and oxidation on the point defects in IG-110 graphite are studied in this work. The 190 keV He<sup>+</sup> implantation treatments at room temperature with fluences of 1 × 10<sup>15</sup>, 5 × 10<sup>15</sup>, 1 × 10<sup>16</sup> and 1 × 10<sup>17</sup> cm<sup>–2</sup> are performed to induce 0.029, 0.14, 0.29 and 2.9 displacements per atom respectively. Oxidation treatments are performed at 850 ℃ for 10, 15, 20 and 25 min. Different sequences of He<sup>+</sup> ion irradiation and oxidation are performed, which include irradiation only (Irr.), oxidation only (Ox.), irradiation followed by oxidation (Irr.-Ox.), and oxidation followed by irradiation (Ox.-Irr.). Raman spectrum shows that with the increase of ion irradiation dose, the intensity ratio of D peak to G peak (<i>I</i><sub>D</sub>/<i>I</i><sub>G</sub>) first increases and then decreases, implying that the point defects in graphite are induced by ion irradiation and the point defects evolve as dose increases; the degree of graphitization increases after oxidation, implying that the point defects are recovered by the annealing effect at high temperature, and the point defects decrease after oxidation. This makes Ox.-Irr. samples have a lower point defect content than Irr. samples, and leads Irr.-Ox. samples to possess a higher point defect content than Ox. samples. The positron annihilation Doppler broadening tests reveal that there are only point defects after ion irradiation and oxidation have partially recovered point defects. The ion irradiation and oxidation have opposite effects on the evolution of point defect in graphite. The ion irradiation increases the average <i>S</i>-parameter and reduces the average <i>W</i>-parameter, while oxidation reduces the average <i>S</i>-parameter and increases the average <i>W</i>-parameter. The annealing effect at 850 ℃ cannot completely recover the point defects in Irr.-Ox. samples.

Coupling Methods for HTR-PM Multicircuit Problem

The High-Temperature Gas-Cooled Reactor–Pebble Bed Module (HTR-PM) is a large-scale complex system that includes reactor core, steam generator, helium circulator, and other important components. When integrating these components, coupling problems such as multiphysics problem, multicircuit problem, multiscale problem, and multimodule problem arise in the numerical simulation. The HTR-PM multicircuit system comprises the primary circuit and secondary circuit, which are simulated by two independent codes and coupled by the interface in the once-through steam generator. Although time-consuming, Picard iteration is a feasible and convenient coupling method to integrate two components because oversolving in the early stages of the iteration causes strong fluctuation between circuits. To address this problem, optimization of the maximum subiteration number and convergence precision have been implemented to improve the efficiency and numerical stability of the Picard iteration. The Dynamic Residual Balance method, an improved version of the Residual Balance method, is proposed to prevent oversolving inside the subiterations. It takes into consideration fluctuation between circuits, and this method is robust in a wide range of cases. Moreover, the nonlinear preconditioned Jacobian-Free Newton-Krylov method, which has less fluctuation between circuits than Picard iteration, is a coupling scheme that updates all the solution variables from the primary circuit and the secondary circuit simultaneously. Outstanding convergence and efficiency can be obtained by implementing the proper preconditioner in this HTR-PM multicircuit problem. The downside is that it requires significant modification to the legacy codes.

Radiological habitability evaluation of the control room of the high-temperature reactor pebble-bed module: post-accident dose assessment and sensitivity studies

ABSTRACT The radiological habitability of a control room is important for nuclear emergency response, which is also a licensing prerequisite for nuclear power plants. It is determined by both atmospheric relative concentrations and doses received via different pathways. However, most recent studies have focused only on the former, which may not be adequate. The present study therefore investigates the radiological habitability and its sensitivity to different parameters in the high-temperature gas-cooled reactor pebble-bed module power plant. For three typical accidents, the study estimates the body, thyroid and skin doses received via different pathways using the Nuclear Regulatory Commission recommended ARCON96 and dose calculation method in RG 1.195. To make a realistic evaluation, the latest design and site-specific information, including the unique accidental source term, are collected and used as input parameters. The evaluation results reveal that the total dose of different pathways in the control room is far below the limit, which demonstrates the effectiveness of the current design. The inhalation exposure is the dominant pathway, and iodine and caesium are the primary contributors of the inhalation dose. The particle filter removal fraction is the most influential parameter in an accident which the activities of metal radionuclides are high.

Discussion on emergency response capability configuration for inherent safe HTR

The design philosophy in High-Temperature Gas-Cooled Reactor (HTR) ensures its inherent safety features, and these inherent safety features further determine its different accident characteristics. The severe core damage can be eliminated by adequate arrangements in HTR, and the fuel temperature will never go beyond the failure limitation to cause significant radioactive release. Therefore, the radioactive material released into the atmosphere even for the most severe anticipated accident of HTR is much smaller than that of Light Water Reactors (LWRs) with the same rating. In this sense, the emergency response capacities (ERCs) stipulated for LWRs are not necessarily all suitable for HTR. For example, the study on emergency planning zone (EPZ) for HTR shows that it could be limited within the exclusion area boundary, about several hundred meters away from the reactor, which is much smaller compared to that in LWRs.In this paper, the other prime ERCs besides EPZ will be re-evaluated one by one for HTR nuclear power plant (NPP) with such inherent safety features. The discussion in this paper is based on high temperature gas cooled reactor pebble-bed module (HTR-PM) being constructed at Shidao Bay, Shandong Province, China. It can be concluded that the offsite emergency planning for HTR could be largely replaced by the conventional industrial emergency system for non-nuclear industries; and some onsite ERCs, such as emergency facilities, emergency communication and notification, can be simplified in terms of radiation protection and urgency requirement respectively.

Research on the fuel loading patterns of the initial core in Chinese pebble-bed reactor HTR-PM

Modular high-temperature gas-cooled reactor (HTGR) is a kind of safe and advanced nuclear energy system, which can provide electric power and high-temperature process-heat efficiently. At present, a 200 MWe demonstration plant, named high-temperature gas-cooled reactor – pebble-bed module (HTR-PM), is being built in China and is planned to be critical in one or two years. In this paper, the physical consideration on the fuel loading patterns of the initial core of HTR-PM is given, and the initial core of HTR-PM is proposed to be made up of low-enrichment fuel spheres and graphite spheres. Then several representative schemes are simulated using VSOP code. The relationship between the key physical parameters (fuel enrichment, flux distribution, temperature coefficient, power per fuel sphere, charging rate) and the volume fraction of fuel spheres in the initial core is analyzed. As a result, the feasible region of the volume fraction of fuel spheres in the initial core is determined as [5/15, 11/15]. The work of this paper can provide support for the fuel management of HTR-PM in the future.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

Coupling methods for HTR-PM primary circuit

The purpose of this paper is to analyze coupling methods for High-Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) primary circuit, including reactor core, steam generator, and other components. Each component has individual characteristics and is simulated by an independent code package. A feasible way to integrate these independent codes is by using Picard iteration. In practice, the primary circuit is divided into multiple parts based on physical fields or spatial domains. Another coupling method is Jacobian-free Newton-Krylov (JFNK), which is frequently applied for the simulation of coupled multi-physics systems due to its efficiency and capability for large-scale nonlinear problems. To demonstrate the efficiency of coupling method for tightly coupled primary circuit systems, several Picard and JFNK methods were compared consistently in this paper, where all the models, equations and codes for each method are the same. Especially for neutronics, we use the same additional equation for JFNK as the outer iteration step in power iteration. Results show that the equations in the reactor core and outside the reactor core should be solved simultaneously in Picard iteration, otherwise additional thermal-hydraulics iterations are necessary. JFNK methods have shown better convergence behavior than Picard iteration, especially for computational efficiency. Since the nonlinear residual is evaluated directly from nonlinear equation set without iteration, linear preconditioned JFNK has shown efficient convergence, which is recommended by the authors as the first choice for HTR-PM primary circuit if the linear preconditioned JFNK is applicable and the well-performing preconditioner is available. However, for utilizing physical iteration processes, nonlinear preconditioned JFNK is a practical choice when integrating legacy codes.

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

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.

CFD investigation of bypass flow in HTR-PM

Predictions of various kinds and distributions of bypass flows are important to the thermal hydraulic design of the HTR-PM (High Temperature gas-cooled Reactor-Pebble-bed Module). In the reactor core, gaps among graphite reflectors are widely distributed and connected, which become typical bypass flow paths interacting with the main flow through the pebble bed. In previous researches, simplified flow network was employed to present the main flow through the pebble bed, and a great many interconnected bypass flow paths, while some irregular gaps were modeled by the computational fluid dynamic (CFD) tools to clarify the resistance coefficients. In present paper, a full CFD model was used to further investigate the bypass flow in the HTR-PM reactor core, which could reveal more local helium flow and heat transfer phenomena if vertical gaps close to pebble bed existed. With separately verified CFD models, the vertical gap bypass flow problem was simulated by the full three-dimension symmetrical model including the helium cooling channel, the cold helium plenum, the cavity above the pebble bed, the bottom reflector, the hot helium plenum and the vertical gap from top to bottom reflectors. Flow distributions in the whole height of active core, as well as radial flow directions along the interface of pebble bed and gap, were focused to find out that the bypass flow rate varied in the vertical gap. For horizontal helium flows, the cold helium mainly flowed into the cavity above the pebble bed while hot helium flowed out of pebble bed. In the full power operation condition, the bypass flow rate ratio was between 0.33% and 1.94% of the total helium flow rate when the gap was consistent with 1.6 mm in width. More gap sizes could be modeled in the future to study the effect to the bypass flow rate ratio.

Endurance test of full-scale mock-up helium circulator for HTR-PM

The High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world’s first pebble-bed modular HTGR commercial demonstration plant through the Chinese National Science and Technology program. HTR-PM is expected to generate electricity during 2018 in Shidao Bay, China. Helium circulator is a critical component required for HTR-PM to circulate reactor coolant through the reactor core and steam generators. Since there is no engineering experience with this circulator, a full-scale mock-up was manufactured for tests. The whole parameters of the mock-up were as same as the actual plant circulator for HTR-PM. The mock-up design is a vertical layout with its rotor supported by active magnetic bearings. The whole mock-up including motor and cooler are sealed in a pressure vessel. Since the mock-up is a large mechanical-electrical system and the design includes many new systems, extensive tests are needed to ensure its safety and reliability. An Engineering Test Facility for Helium Circulator (ETF-HC) was then built in Institute of Nuclear and New Energy Technology (INET). Firstly, factory nitrogen tests were conducted by the manufacturer and the customer. After the successful factory tests, the mock-up was delivered to INET to be tested on ETF-HC. One and a half years were spent for the helium endurance tests, including time spent on resolving several helium sealing problems. The mock-up run continuously for 50 h at the rated conditions. Fifty service cycle tests were also conducted. More than 500 operating hours were accumulated during the endurance tests. The final performance shows that the mock-up of helium circulator meets the design requirements and is reliable.

Dynamics and thermal analysis of the auxiliary bearing for the helium circulator of HTR-PM

The high-temperature gas-cooled reactor-pebble bed module (HTR-PM) has been proposed by the Institute of Nuclear and New Energy Technology of Tsinghua University. The helium circulator is equipped in the primary loop of HTR-PM to drive high-temperature helium for energy exchange. The active magnetic bearing (AMB), replaced by the ordinary mechanical bearing, is chosen as the perfect supporting scheme for the helium circulator of HTR-PM. Additionally, the auxiliary bearing is applied in case of AMB failures and has become indispensable to provide mechanical supports and displacement constraints for the rotor. When the rotor drops down, the auxiliary bearing needs to sustain fierce impacts and corresponding thermal growth. While due to the special environment in the primary loop of HTR-PM, oil or grease lubrication can’t be applied in the auxiliary bearing. A new design of unlubricated auxiliary bearing is proposed in this paper. This paper mainly deals with the following two aspects to reveal the rotor drop: dynamic analysis including fatigue life prediction, and corresponding thermal growth. Moreover, dropping experiments are carried out in the HTR-PM helium circulator system. The experimental results can verify the auxiliary bearing capacity to endure the dynamic impact of the dropping rotor, and provide suggestions for the auxiliary bearing type selection. The results in this paper will provide theoretical suggestions for further engineering design and application of the auxiliary bearing in the helium circulator of HTR-PM.

Development of emergency planning zone for high temperature gas-cooled reactor

Emergency planning zone (EPZ) is a basic element of nuclear emergency planning, whose architecture is mainly determined by the characteristics of early light water reactors (LWRs) at present. As a candidate of Gen IV reactors, high temperature gas cooled reactor (HTR) is required to consider the issue of EPZ in current regulation framework of nuclear emergency. The issue we discussed here is based on high temperature gas cooled reactor pebble-bed module (HTR-PM) being constructed at Shidao Bay, Shandong Province, China. Since HTR is much different from LWRs in design concept, system configuration and accident characteristics, the emergency planning developed for HTR can be simplified comparing to the current LWR-based emergency planning. HTR, like HTR-PM, implements inherent safety features and ensures the elimination of severe core damage and large release of radioactive material. So that EPZ needs to be adjusted to match the safety features of HTR to reduce the cost of construction and operation, and enhance the feature of good neighbor. The commonly used technical framework introduced in NUREG-0396, which is documented by Nuclear Regulatory Commission (NRC) is adopted as base to discuss the EPZ issue. The accidents need to be considered during EPZ development, including the design basis accidents (DBAs) and emergency reference release categories (ERRCs), are firstly confirmed according to the accident characteristics of HTR-PM. We conclude that radii of EPZ for plume exposure pathway and ingestion exposure pathway can both be the same as exclusion area boundary, which supports the further simplification of nuclear emergency of HTR.

The influence of nuclear graphite oxidation on air ingress accident of HTR-PM

The high temperature gas-cooled reactor pebble bed module (HTR-PM) has wide potential applications in the development of advanced nuclear energy. The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a kind of beyond design basis accident of HTR-PM, which will result in air ingress into the reactor core. Even though the rare probability of this accident, it still has got a lot of concerns in the safety analyses of HTR-PM, due to the possible severe consequences of the air ingress accident, which would cause oxidation of the graphite material in the reactor. The oxidation of matrix graphite of spherical fuel elements may lead to the exposure of the TRISO fuel particles and the increase of the radioactivity release. The oxidation of the bottom reflector may cause the decrease of mechanical strength of the nuclear grade graphite. Different kinds of nuclear graphite of the bottom reflector will influence the consequences of the DEGB of the horizontal coaxial gas duct accident. This paper mainly analyzed the influence of the nuclear grade graphite used as the bottom reflector on the consequences of this accident by using the system analysis TINTE code.

Development and Application of an Engineering Simulator for HTR-PM Using THERMIX/BLAST and vPower

Different from most current commercial nuclear power plants, the High-Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) power plant consists of two reactor modules connected to a common steam turbine system that will bring a special coupling effect between the two reactor modules. An engineering simulator of the HTR-PM plant was developed by embedding the THERMIX/BLAST code into the vPower simulation platform. Two sets of nuclear steam supply systems of HTR-PM, including two reactors, two steam generators, two helium blowers, and the helium flow ducts, were simulated by two THERMIX/BLAST code modules, respectively. The entire secondary loop system was simulated using intrinsic models of the vPower simulation platform. The vPower platform connects and synchronizes the two THERMIX/BLAST modules, as well as the simulation module for the secondary loop system. The engineering simulator was applied to simulate the behavior of HTR-PM under steady-state operation, startup and shutdown processes, and accident conditions. The coupling effect during the condition conversion process and the thermal characteristics under accident conditions of HTR-PM were analyzed by the engineering simulator.

Numerical simulation on heat transfer process in the reactor cavity of modular high temperature gas-cooled reactor

In order to enhance the safety of the reactor, the passive residual heat removal system (PRHRS) is adopted by both the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) and the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). The heat transfer process in the reactor cavity is very important for the whole PRHRS to remove the heat from the reactor, especially once an accident occurs. A three-dimensional numerical model used to simulate the heat transfer process in the reactor cavity of HTR-10 was developed in this study, in which the flow of water in the water cooled pipe was included. The effects of different symmetry boundary conditions on the heat transfer process were studied. Three experimental conditions of HTR-10 were simulated, and the local and global heat transfer processes were analyzed. The results show that the effects of symmetry boundary condition can be neglected when two or more water cooled pipes are included in the model. The temperature of air varies greatly near the reactor pressure vessel (RPV) and the water cooled panel. The maximum temperature occurs at the top of the RPV. With the elevation increasing, the rising rate of water temperature gradually increases. Moreover, radiation become more dominant as the RPV temperature rises.