MW High Temperature Gas-cooled Reactor Research Articles

Thermal-Hydraulic Analysis of Pebble Bed High Temperature Gas-Cooled Reactor After Shutdown Based on FLUENT Software

In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media model and user-defined scalar (UDS) function of FLUENT software. Then, the model was applied to the numerical simulation of the shutdown test of the 10 MW high temperature gas-cooled test reactor (HTR-10) at 9 MW power level, and the temperature distribution and flow field distribution in the reactor were obtained and compared with the results of the experimental data. The reliability of the model in this paper was verified. Based on the model, the effects of the water-cooled panel temperature and the initial core temperature on the thermal-hydraulic characteristics of HTR-10 after shutdown were further explored. The results show that there is a decoupling phenomenon between the residual heat transfer within the core and the heat dissipation of the pressure vessel. The initial core temperature has relatively little effect on the heat dissipation and maximum temperature of the pressure vessel, but it has a significant impact on the thermal characteristics of the core area.

Inventories of Short-Lived Fission Gas Nuclides in Nuclear Reactors

Taking inventories in reactor cores is critical for understanding their radioactive source terms and establishing the relationship between the activity concentration in the primary loop and the status of the reactor core’s fuel. However, there is a niche in which a simple but accurate relationship between reactor conditions and nuclide inventories can reliably predict the fission gas nuclide activities of the reactor core in the primary loop. In this study, a simple and efficient model called “Inventories of a Point Reactor for Fission Gas Nuclides” (IPRFGN) was proposed to calculate and interpret such inventories, in which a 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was used as the test case. The present study findings were consistent with those of a general point–depletion burnup code such as the KORIGEN code. Here, the relative error was <1%. Based on the application of the IPRFGN model in HTR-10, the results indicate that the proposed IPRFGN model has provided the relationship between the inventories of fission gas nuclides in the core and the reactor conditions in all types of nuclear fission reactors. In the future, the IPRFGN model will be used for calculating fission gas nuclide inventories in various reactors.

Preliminary study on HTR-10 operating at higher outlet temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003.Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850–1000 °C is carried out.In this paper, the preliminary thermal hydraulic design of the HTR-10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature.Due to the lower power and power density, as well as the large heat capacity of the graphite reflector and the carbon brick, the maximal fuel temperature during the DLOFC would not exceed the limitation of 1620 °C (250th meeting of the reactor safety commission, 1990). Nevertheless, the higher fuel temperature during the normal operation should be of concern.

Preliminary Study on HTR-10 Operating in Higher Outlet Temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003. Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850~1000 °C is carried out. In this paper, the preliminary thermal hydraulic design of the HTR- 10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the typical DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature. The preliminary results show that, operated at the higher outlet helium temperature, the original acceptance criteria for HTR-10 will be challenged. In the future, the design optimization, as well as the possible modification of these acceptance criteria, which were set more than two decades ago, should be studied based on the current knowledge of the fuel element properties and structure material properties.

Source term analysis and sorting scheme in post-irradiation spherical pyrometric elements

The 10 MW high-temperature gas-cooled reactor (HTR-10) is the most advanced of the fourth generation reactors, and has become a good test platform for very high-temperature operations. Spherical fuel elements are utilized in HTR-10, and the maximum temperature of the fuel element is one of the primary technical factors restricting the outlet temperature of the core. In our experiment, a pyrometric element with different wire melting points, is added to the HTR-10 reactor core to measure the maximum temperature distribution. How to sort out the pyrometric element, and analyze the source terms in post-irradiation spherical pyrometric elements when cycled out of the core, are the key issues of this study. This study can provide technical support for the internal temperature measurement, and analysis of the core, as well as for very high-temperature operation of HTR-10.

Accuracy and efficient solution of helical coiled once-through steam generator model using JFNK method

The Helical coiled tube Once-Through Steam Generator (H-OTSG) is key equipment in the High-Temperature Gas-cooled Reactor (HTGR) system, where the fission energy is transferred from the helium side to the water side, together with the complicated heat transfer and phase change phenomenon. In detail, the water is heated to the water-steam two-phase flow by the hot helium and further to the superheated steam. As a result, the H-OTSG is a multi-phases, multi-physics, and multi-domain highly nonlinear system due to the complex constitutive equations in the two-phase flow model. The accuracy and efficient solution of this highly nonlinear system is a key point for the successful simulation of the H-OTSG. In this work, the advanced Jacobian-Free Newton Krylov (JFNK) method is used instead of the traditional semi-implicit iteration method to pursue high computational performance. An efficient physical-based preconditioner is proposed, which is derived from the original semi-implicit based method. Moreover, the numerical technologies of the phase generation/disappearance and the global convergence algorithms of JFNK method are implemented and discussed. The two-phase drift-flux model is used to describe the two-phase flow in the waterside and the second-order accuracy temporal and spatial numerical scheme is employed to discrete the conservation equations. A new H-OTSG simulation code based on JFNK method is developed in this work and applied to analyze the steady and transient state behavior of the H-OTSG in the 10 MW high-temperature gas-cooled reactor (HTR-10). The simulation results show that the accuracy of the newly developed H-OTSG simulation code agrees well with the design data under different operating conditions. Furthermore, the computational performance of the JFNK method with the physical-based preconditioner is the best compared with the traditional Picard iteration method, Newton–Krylov method and JFNK method without preconditioner.

Summary of Tritium Source Term Study in 10 MW High Temperature Gas-Cooled Test Reactor

Tritium (3H) has been increasingly researched when assessing the environmental impact of nuclear reactors and other nuclear facilities because it is widely present in nuclear systems and can easily enter the environment. The first pebble-bed gas-cooled test reactor in China, the 10 MW high temperature gas-cooled test reactor (HTR-10), uses helium, graphite, and graphite spheres containing embedded tristructural-isotropic–coated particles as primary coolant, reflectors, and fuel elements, respectively. Several experiments that involved the 3H source term in HTR-10 were performed, and they measured the 3H specific activity and its distribution in the irradiated graphite spheres from the core, 3H activity concentration in the primary helium, 3H activity concentration during the regeneration of the molecular sieve adsorber in the helium purification system, and 3H amount in the gaseous effluent discharge from the stack. The experimental data were summarized and compared with the theoretical predictions. The balance diagram of the 3H source term in HTR-10 is introduced in this paper. Sensitivity analysis was performed to illustrate the effect of the 3He abundance in the primary helium and Li content in the graphite reflectors on the 3H activity concentration in the primary coolant of HTR-10. The interactions between graphite and different hydrogen isotopes (1H, 3H, 1H2, 1H3H, and 3H2) were investigated using first-principles calculations and the diffusion theory. The results indicated that molecular 3H tended to diffuse in graphite.

Coaxial stacking disjoint model study of 10 MW high temperature gas-cooled reactor (HTR-10)

HTR-10 is quite different than conventional reactors due to its geometry, spherical fuel elements and random distribution of fuel elements and graphite balls. It is desired to built realistic and accurate reactor model. In this paper, the detailed mathematic geometrical relationships of HTR-10 are studied on HTR-10′s coaxial stacking disjoint model. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by SuperMC. Compared with available data, this model can achieve accurate results in terms of initial criticality of the HTR-10 reactor.

A Comprehensive Study of the 14C Source Term in the 10 MW High-Temperature Gas-Cooled Reactor

ABSTRACTWhile assessing the environmental impact of nuclear power plants, researchers have focused their attention on radiocarbon (14C) owing to its high mobility in the environment and important radiological impact on human beings. The 10 MW high-temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China that adopted helium as primary coolant and graphite spheres containing tristructural-isotropic (TRISO) coated particles as fuel elements. A series of experiments on the 14C source terms in HTR-10 was conducted: (1) measurement of the specific activity and distribution of typical nuclides in the irradiated graphite spheres from the core, (2) measurement of the activity concentration of 14C in the primary coolant, and (3) measurement of the amount of 14C discharged in the effluent from the stack. All experimental data on 14C available for HTR-10 were summarized and analyzed using theoretical calculations. A sensitivity study on the total porosity, open porosity, and percentage of closed pores that became open after irradiating the matrix graphite was performed to illustrate their effects on the activity concentration of 14C in the primary coolant and activity amount of 14C in various deduction routes.

Experimental Investigation of 14C in the Primary Coolant of the 10 MW High Temperature Gas-Cooled Reactor

ABSTRACTThe very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 (14C) was established in the helium purification system of the HTR-10 primary loop, which could sample 14C from the coolant at three locations. The results showed that activity concentration of 14C in the HTR-10 primary coolant was 1.2(1) × 102 Bq/m3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14C in the core of HTR-10 and its concentration in the primary coolant. The activation reaction of 13C has been identified to be the dominant 14C source in the core, whereas in the primary coolant, it is the activation of 14N. These results can supplement important information for the source term analysis of 14C in HTR-10 and promote the study of 14C in HTGRs.

The application of the DEM-based burnup construction method in the pebble burnup history analysis in HTR-10

In the 10 MW high temperature gas-cooled test reactor (HTR-10) operation history, the reactor experienced frequent shutdown periods and irradiation periods with varied power. The complicated operation history is supposed to influence the pebble burnup history. Therefore, this paper firstly introduces a DEM-based pebble burnup tracking method, which combines the DEM pebble flow simulation and the pebble burnup calculation. The method can track the burnup of each pebble in the core individually. Then the method is applied to construct the pebble burnup under the complicated HTR-10 real operation history, and the construction results are analyzed to investigate the influence of operation history on the pebble burnup history. The analyzing results show that the average final pebble burnup at the end of the history is 17–18 GWd/tU. Most of the pebbles experience 2–3 cycles before the reactor finally stops. Moreover, the correlation between the pebble burnup and the Cs-137 activity is found to be influenced by the operation history. The tail of the original single linear distribution turns to be double linear distribution, which brings system error for determining the pebble burnup by Cs-137 activity. In addition, a method that can correct the system error is promised with the help of some recommended auxiliary nuclides.

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

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.

A one-dimensional code of the passive residual heat removal system for the modular high temperature gas-cooled reactor

The characteristics of the passive residual heat removal system (PRHRS) are very important for the safety of the modular high temperature gas-cooled reactor (MHTGR). In this study, a one-dimensional code used to calculate the transient characteristics of the PRHRS of the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) was developed using the C programming language. Water loop and air cooled condenser are the crucial parts of the PRHRS. Their characteristics under the startup condition, normal operating condition and shutdown condition of HTR-10 were respectively calculated with the one-dimensional code. To validate the code, the computational results were compared with the experimental results obtained from the HTR-10. Also, the flow and heat transfer conditions in the water loop and the air cooled tower were also analyzed. The results show that the computational results are in good agreement with the experimental results. Under the normal operating condition, the characteristics of the PRHRS are mainly affected by the environmental temperature. Under the startup condition and the shutdown condition, the characteristics of the PRHRS are affected by both the nuclear power and the environmental temperature, and the effects of the nuclear power play a dominant role.

A comprehensive study on source terms in irradiated graphite spheres of HTR-10

With previously developed experimental methods which include the preparation and measurement process for the graphite sample, two new irradiated graphite spheres with surface γ dose rates of 51.00 μSv/h and 0.14 μSv/h from the reactor core of the 10 MW high temperature gas-cooled reactor (HTR-10) have been investigated experimentally. The total β counting rate, the β spectra and the γ spectra for each graphite sample of irradiated graphite spheres were recorded with a total α/β counting measuring apparatus, a liquid scintillation counter and a high-purity germanium detector connected to a multichannel analyzer, respectively. Combined with previous experimental data of two irradiated graphite spheres with surface γ dose rates of 25.10 μSv/h and 1.17 μSv/h, the types of key nuclides in the irradiated graphite sphere of HTR-10 were determined, which were H-3, C-14, Co-60, Cs-137, Eu-152 and Eu-154. The distributions for each nuclide in four irradiated graphite spheres were compared. The generation mechanisms of H-3, C-14, Co-60, Cs-137, Eu-152 and Eu-154 in the irradiated graphite sphere of HTR-10 were discussed and analyzed. Based on all the experimental data regarding impurities and uranium contamination in the matrix graphite of HTR-10 available, a sensitivity analysis was performed to explain the effect of impurities and uranium contamination on the specific activity of key nuclides in the graphite sphere. The influence of the neutron flux and the dwell time in the core on the specific activity of key nuclides was also considered. The differences of experimental specific activities among these irradiated graphite spheres were compared and explained. Current comprehensive studies on irradiated graphite spheres of HTR-10 can provide valuable information for the source term analysis, waste minimization and radiation protection of high temperature gas-cooled reactors (HTGRs).

Three-dimensional numerical investigation of heat transfer process in the reactor cavity of HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is the first Modular High Temperature Gas-cooled Reactor (MHTGR) in China. The heat transfer process in the reactor cavity of HTR-10, which was very important for the decay heat to be removed, was investigated with a three-dimensional numerical model in this study. The temperature distributions of air, water cooled panel and water in five experimental conditions were analyzed. The heat transfer rates on the reactor pressure vessel (RPV) wall, front and back wall of the water cooled panel, and outer wall of the water cooled pipe were calculated. Besides, the percentages of radiation heat transfer rate on these walls were given, and the variations of them with the RPV temperature rising were also analyzed. The results show that radiation between the RPV and the steel sheet, heat conduction between the steel sheet and the water cooled pipe, and convection between the water cooled pipe and the water form the main way that the heat is removed, especially when the RPV temperature is high. The maximum temperature of air occurs on the top of the RPV. In addition, as the RPV temperature rises, the heat transfer rates and the percentages of radiation all increase.

Dynamic Analysis for the Rotor Drop Process and Its Application to a Vertically Levitated Rotor/Active Magnetic Bearing System

Active magnetic bearings (AMBs) have been utilized widely to support high-speed rotors. However, in the case of AMB failure, emergencies, or overload conditions, the auxiliary bearing is chosen as the backup protector to provide mechanical supports and displacement constraints for the rotor. With lack of support, the auxiliary bearing will catch the dropping rotor. Accordingly, high contact forces and corresponding thermal generation due to mechanical rub are applied on the dynamic contact area. Rapid deterioration may be brought about by excessive dynamic and thermal shocks. Therefore, the auxiliary bearing must be sufficiently robust to guarantee the safety of the AMB system. Many approaches have been put forward in the literature to estimate the rotor dynamic motion, nonetheless most of them focus on the horizontal rotor drop and few consider the inclination around the horizontal plane for the vertical rotor. The main purpose of this paper is to predict the rotor dynamic behavior accurately for the vertical rotor drop case. A detailed model for the vertical rotor drop process with consideration of the rotating inclination around x- and y-axes is proposed in this paper. Additionally, rolling and sliding friction are distinguished in the simulation scenario. This model has been applied to estimate the rotor drop process in a helium circulator system equipped with AMBs for the 10 MW high-temperature gas-cooled reactor (HTR-10). The HTR-10 has been designed and researched by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University. The auxiliary bearing is utilized to support the rotor in the helium circulator. The validity of this model is verified by the results obtained in this paper as well. This paper also provides suggestions for the further improvement of auxiliary bearing design and engineering application.

Experimental verification on design model of the passive residual heat removal system of MHTGR

The passive residual heat removal system (PRHRS) of Modular High Temperature Gas-cooled Reactor (MHTGR) plays an important role in removing the residual heat of the reactor core and thus keeps nuclear power plant in safe. The PRHRS has been adopted for both the 10 MW High Temperature Gas-cooled Reactor (HTR-10) and the High Temperature Reactor Pebble-bed Modules (HTR-PM). A specific steady-state model has been built to calculate whether the system has sufficient heat transfer capacity under design parameters. The accuracy of the steady-state model is the key factor. Experiments of PRHRS was designed and conducted on HTR-10. The model's computation results were compared with the experimental data to analyze the deviations and to further improve the model. Other improvements are also suggested.

Unbalance Compensation of a Full Scale Test Rig Designed for HTR-10GT: A Frequency-Domain Approach Based on Iterative Learning Control

Unbalance vibrations are crucial problems in heavy rotational machinery, especially for the systems with high operation speed, like turbine machinery. For the program of 10 MW High Temperature gas-cooled Reactor with direct Gas-Turbine cycle (HTR-10GT), the rated operation speed of the turbine system is 15000 RPM which is beyond the second bending frequency. In that case, even a small residual mass will lead to large unbalance vibrations. Thus, it is of great significance to study balancing methods for the system. As the turbine rotor is designed to be suspended by active magnetic bearings (AMBs), unbalance compensation could be achieved by adequate control strategies. In the paper, unbalance compensation for the Multi-Input and Multi-Output (MIMO) active magnetic bearing (AMB) system using frequency-domain iterative learning control (ILC) is analyzed. Based on the analysis, an ILC controller for unbalance compensation of the full scale test rig, which is designed for the rotor and AMBs in HTR-10GT, is designed. Simulation results are reported which show the efficiency of the ILC controller for attenuating the unbalance vibration of the full scale test rig. This research can offer valuable design criterion for unbalance compensation of the turbine machinery in HTR-10GT.

Research on dynamics and experiments about auxiliary bearings for the helium circulator of the 10 MW high temperature gas-cooled reactor

The 10MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University. The auxiliary bearing is utilized in this system to meet particular requirements for the reactor. The main role of the auxiliary bearing is to constrain rotor displacements and also to support the rotor when the rotor drops down, which is caused by the active magnetic bearing (AMB) failure. The auxiliary bearing needs to endure huge impact, rapid angular acceleration and thermal shock. On the one hand, complex geometrical constructions and forces applied on the system bring difficulties and restrictions to establish an appropriate model to reveal the actual dynamic process. On the other hand, large volumes of data obtained from experiments show velocities and displacements of the rotor during the rotor drop process and then can indicate the actual dynamic interactions to a great extent. The research in this paper is based on the test rig of the AMB helium circulator of HTR-10. This paper aims to analyze the dynamic performance and contact forces of the rotor by processing experimental data. A measurement to estimate forces developed due to impacts of the rotor and the auxiliary bearings is presented. It is of great significance and provides certain foundation to elaborate the rotor drop process for the AMB helium circulator of HTR-10.

Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10)

Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR). To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10) in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% forγdose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.