Institute Of Nuclear And New Energy Technology Research Articles

Experimental measurement of stiffness coefficient of high-temperature graphite pebble fuel elements in helium at high temperatures

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.

Numerical simulation of the transient graphite oxidation process based on a dynamic mesh method

The high-temperature gas-cooled reactor uses graphite-matrix spherical fuel elements to coat the tri-structural isotropic fuel particles (TRISO). In accident conditions, graphite oxidation may cause fuel particle exposure, which further leads to severe radioactive contamination. Therefore, it is very important to study the oxidation behavior of graphite during air-ingress accidents. This work aimed to develop a reaction kinetic model of graphite oxidation based on the Institute of Nuclear and New Energy Technology (INET) experiment correlations and to study the transient process of graphite oxidation using computational fluid dynamics and a dynamic mesh method. The results show that convective diffusion and mass diffusion dominate the distribution of reaction products at different flow rates. The reaction rate in the early stage of oxidation was high and then decreased rapidly until the oxygen consumption rate and supplementation rate reached equilibrium. After that, the oxidation rate decreased slowly due to the reduction of graphite surface area. The fluctuating flow around the graphite leads to anisotropic oxygen concentrations at different inflow angles on the graphite surface. The fluctuations caused a non-uniform ellipsoidal graphite structure, in which graphite has a smaller radius of curvature and pits on the windward side and a larger radius of curvature on the leeward side. The oxidation mass of graphite is in good agreement with the INET experimental data, which shows that this study can provide a method for the estimation of the mass and shape of graphite oxidation.

The direct measurement of HTR-10 in-core neutron flux

The direct measurement of neutron flux in reactor core is of great significance for accurately understanding the physical characteristics of reactors, verifying the correctness of core simulation calculations, and developing and improving physical analysis models. The high-temperature gas-cooled reactor HTGR is a pebble-bed type modular reactor which is developed by the Institute of Nuclear and New Energy Technology (INET) in Tsinghua University. Large number of spherical fuel elements are piled up in the reactor core. During operation, the fuel elements flow downward slowly under the actions of gravity. As a consequence, the in-core neutron flux cannot be measured directly in the reactor core for it is impossible to arrange measurement points in the pebble bed. In a recent study, we took advantage of the neutron activation features of the temperature-measuring graphite pebbles (monitors) which were put into the core during the temperature measurement experiment period, and measured the in-core neutron flux of HTR-10. The cobalt element in the monitors was activated during the reactor’s operation period and was utilized to reflect the level of neutron flux. The experimental data including the peak value and radial distribution of in-core neutron flux was obtained. The results show that the neutron flux in the radial direction presents a distribution law with higher value in center and lower value in periphery, with an overall uniform distribution. In addition, the experimental and theoretical calculation data are compared, and consistent results have been shown. The study of in-core neutron flux measurement provides an important experimental evidence for verifying the physical analysis model of HTGRs, and develops an effective method for direct measurement of neutron flux of pebble bed type reactors.

ARCHER - a new Three-Dimensional method of characteristics neutron transport code for Pebble-bed HTR with coarse mesh finite difference acceleration

A new deterministic code ARCHER using Three-Dimensional (3-D) Method of Characteristics (MOC) has been developed by Institute of Nuclear and new Energy Technology (INET), Tsinghua University, aiming to obtain high-fidelity transport solution of pebble-bed High Temperature gas-cooled Reactor (HTR) with explicit pebble-bed geometry. However, 3-D MOC usually suffers from the slow convergence rate without efficient acceleration methods. In this work, the Coarse Mesh Finite Difference (CMFD) based on regular mesh is implemented to accelerate 3-D MOC calculation, which is available for the problems with cylinder and cuboid geometry. The complex geometry of reactor core with randomly stacked spherical fuel elements and irregular helium flow area, as well as the topological relationships between fine Flat Source Region (FSR) mesh for MOC and coarse CMFD mesh, are the key but challenging tasks for high-fidelity transport solution. Constructive Solid Geometry (CSG) and Boolean operation are constructed to describe problem geometry, which has the ability to deal with complex topological relationships between meshes. In addition, tree grid structure is utilized to reduce the time complexity of ray tracing and obtain the grid mapping relationship between coarse mesh and fine mesh. Spatial domain decomposition parallel based on MPI is utilized to alleviate the memory pressure and pursue high computational performance. Ray parallel is further achieved through OpenMP with shared memory. Several problems have been used to verify the geometric modeling capability and computational performance of this new 3-D MOC code for pebble-bed HTR.

An improved homogenization method for the treatment of strong absorbers in pebble-bed HTGR

In pebble-bed high temperature gas-cooled reactors (HTGRs), strong absorbers including control rods and absorber balls are placed in the reflector. The strong absorber regions need to be homogenized for the convenience of downstream whole-core neutronics diffusion calculation. Previously, a homogenization method based on the generalized equivalence theory (GET) was studied. This paper proposes an improved homogenization method by employing a more rigorous discontinuity factor (DF) model, for more accurate treatment of the strong absorbers in the HTGR reflector. The proposed method has been implemented in the PANGU code developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. Numerical results suggest that the newly proposed homogenization method achieves good agreement with the heterogeneous Monte Carlo transport solution and experimental result.

A comparison study on mechanical models for TRISO fuel particle in HTGR

TRISO fuel performance is crucial for the safety of High Temperature Gas-cooled Reactor (HTGR). In the engineering design process of HTR-PM by Institute of Nuclear and New Energy Technology (INET), PANAMA code was adopted to evaluate the mechanical performance and failure probability of TRISO coated particle fuels. However, PANAMA results are quite conservative compared with the recent HTR-PM fuel irradiation experiment. INET develops a new TRISO fuel performance code FRAT that employs more delicate models involving internal gas pressure buildup, mechanical stress analysis as well as temperature and burnup effects. In this paper, we use FRAT code and do a comparison study of mechanical analysis model of TRISO fuel particle. Results yield a high sensitivity of calculated TRISO fuel failure probability to mechanical models adopted in TRISO fuel performance. This work helps to further understand the mechanical models of TRISO fuel particle.

Steady-state thermal fluids analysis for the HTR-PM equilibrium core

The high temperature gas-cooled reactor-pebble bed module (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is a demonstration project of nuclear power plant with modular high temperature reactors (HTRs) in China. So far the detailed engineering design of the HTR-PM has been completed, and now the plant is at its commissioning stage. In the design process of the reactor, the thermal hydraulics design is of extreme importance for the evaluation of the reactor performance under various operation conditions. In this study, the steady-state thermal fluids analysis is performed for the HTR-PM equilibrium core. The analysis methods are introduced and the calculating models are presented. The key thermal-fluid parameters under different power levels are obtained and evaluated, including the temperature field of the reactor, maximum temperatures of the fuel and other components, pressure loss of the core and the primary circuit, and so on. As the most important safety-related parameter, the maximum fuel temperature is much lower than its safety limit for normal operation. The analysis results show that the thermal hydraulics design provides sufficient cooling capability to the core and thus ensures the reactor safety under normal operation.

The NUIT code for nuclide inventory calculations

This work introduces the NUIT (Nuclide Inventory Tool) code for nuclide inventory calculations, which has been developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. NUIT is able to compute isotope density, decay heat, particle spectrum and radioactivity in fuel burnup, material activation and decay calculations. The burnup equations are solved by transmutation trajectory analysis method (TTA) and matrix exponential methods in the NUIT code. The code libraries account for very detailed burnup chains with thousands of isotopes and all major neutron-induced reactions and decays, which adapt to general purpose applications in fission reactors. Besides, NUIT code provides friendly interface to utilize problem-specific neutron spectrum and tabular effective one-group cross sections produced by neutron transport codes. The NUIT code is validated by analytical solutions, experimental data and code-to-code comparisons. The validation results demonstrate the NUIT code’s capabilities in wide-range nuclide inventory calculations.

PANGU code for pebble-bed HTGR reactor physics and fuel cycle simulations

PANGU is a modern computer code for pebble-bed High Temperature Gas-cooled Reactor (HTGR) physics analyses and fuel cycle simulations, developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. This paper gives a comprehensive review on PANGU code’s features including its calculation scheme, methodologies, models and capabilities. Numerical tests are presented for typical pebble bed reactor cases, to verify PANGU code’s applicability. It demonstrates that PANGU code is promising for physical designs of traditional and new-conceptual pebble-bed HTGRs.

Study on the DLOFC and PLOFC accidents of the 200 MWe pebble-bed modular high temperature gas-cooled reactor with TINTE and SPECTRA codes

Validation and verification (V&V) is important and receives significant attention during the design and analysis of the nuclear power plant. As the first commercial modular high temperature gas-cooled reactor (HTR) demonstration plant, during the regulatory review of the preliminary safety analysis report (PSAR) and the final safety analysis report (FSAR) of the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM), the authorities also pay high attention to the code validation and verification.Within the cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China, a code to code verification was carried out between the TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG. Three typical accidents scenarios, two depressurized loss of forced cooling (DLOFC) accidents scenarios and one pressurized loss of forced cooling (PLOFC) accident scenario, have been analyzed both with the TINTE code and the SPECTRA code.The simulation results show good agreement between TINTE code and SPECTRA code. The results also indicate that, due to the inherent safety feature of the HTR-PM design, the maximal fuel temperature during above accidents would never exceed the limitation of 1620 °C, below which there is no additional damage for the TRISO particles, and the SiC layer of each particle can guarantee the fission-product-retention capacity. This work will be helpful to the further study of the HTGR, and more code to code verification is planned.

Measurement on effective thermal diffusivity and conductivity of pebble bed under vacuum condition in High Temperature Gas-cooled Reactor

A full-radius-scale heat test facility has been developed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China to measure the effective thermal diffusivity and conductivity of pebble bed in High Temperature Gas-cooled Reactor (HTGR). In present test, a separate-effect transient experiment under vacuum condition is conducted successfully up to 1200 °C. Through inverse method and experimental temperature data, this paper presents the process of determining these thermal parameters and compares the results with previous researches, which reveals that the results are comparable in low-temperature region and valid in high-temperature region. In addition, it also studies the experimental errors and the results' uncertainties in this test. The correct knowledge on effective thermal diffusivity and conductivity of pebble bed inside the reactor core is significant to ensure the valid design and inherent safety, which will give a better balance between safety and economic competitiveness.

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.

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.

Development of a neutronics analysis code for pebble-bed HTRs

A new code named HINT (High Temperature Reactor Neutronics Tool) is under development in Institute of Nuclear and New Energy Technology (INET). This paper introduces the HINT code’s computational framework and methodologies in lattice physics and core physics computations. Emphases are given to the double heterogeneity treatment and the leakage feedback in broad-group cross-section generation, as these two issues are specific challenges for high temperature reactors (HTR) compared with light water reactors (LWR). Numerical results are presented to verify the HINT code’s capability in HTR fuel pebble and full-core calculations. It demonstrates that the HINT code is promising for pebble-bed HTR neutronics analysis.

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.

INET's catalysts study for HI decomposition in the thermochemical water-splitting iodine–sulfur process for hydrogen production

The iodine–sulfur thermochemical cycle (IS cycle) can be applied to realize water decomposition and hydrogen production by using the solar energy or the nuclear heat from High Temperature Gas-cooled Reactor as heat source. Among the three reactions of the IS process (Bunsen reaction, H2SO4 decomposition and HI decomposition), the HI decomposition has the slowest reaction rate and the lowest thermodynamic equilibrium conversion (about 23% at 500 °C). In order to make HI decompose at the workable reaction rate, catalysts had to be applied in the reaction system. This paper outlines our study on the monometallic and bimetallic catalysts for HI decomposition at INET (Institute of Nuclear and New Energy Technology, Tsinghua University, China). The influences of different supports, transition metals, bimetallic compositions, reaction time on the catalytic performance for HI decomposition were investigated. The optimized catalyst of 2.5 wt.%Pt–2.5 wt.%Ir/C was proposed and adopted to catalyze the HI decomposition in the close-cycle operation of IS-100 facility with the H2 production rate of 100 NL/h. Both the conversion of HI (21%) and the H2 production rate (60 NL/h) reached the technical targets of the nuclear hydrogen project (National S&T Major Projects 2010zx06901: Research on key technologies of hydrogen production from high temperature gas cooled reactor).

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.

Improvement of discontinuity factor for strong absorber region

At Institute of Nuclear and New Energy Technology (INET) the discontinuity factor corrected diffusion method with the homogenization technology was developed and applied in the control rod worth calculation of the pebble bed high temperature gas cooled reactor. But the result with the normal procedure is not accurate enough for a strong absorber. The numerical analysis shows that the strong absorber still has great influence on the flux distribution in the nearby graphite region, so that the flux distribution obtained by the normal diffusion method does not agree with the transport result. Thus, two improvements were proposed in this paper. First, instead of the neutron flux in the middle of the fine mesh, the surface flux of the absorber region was calculated through the net current in the boundary of the region; and then, while the discontinuity factor of the homogenized absorber region should be calculated, the discontinuity factor of the neighboring graphite region on the other side of the interface should also be calculated to eliminate the influence of the strong absorber. The numerical results demonstrate that, based on the improved method, the accuracy of heterogeneous transport calculation can be achieved by a diffusion calculation.

HTR steady state and transient thermal analyses

A thermal model to simulate the steady state and transient behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10) is presented in this work. The helium cooled HTR-10 was designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. In this study, a simulation is performed using the RELAP5 code. 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. The simulation is demonstrated to be in good agreement, showing that the developed model is capable of reproducing the thermal behavior of the HTR-10 in steady state and transient operation conditions.

Thermal Analysis and Simulation of Auxiliary Bearings and Its Application in the High Temperature Reactor-10

The auxiliary bearing is applied to provide mechanical uphold for the rotational dropping rotor when contact event happens due to the active magnetic bearing (AMB) failure emergencies. During the rotor drop process, the auxiliary bearing will endure huge impact force and friction heat generation. The thermal behavior will affect the mechanical interaction and dynamic behavior of the auxiliary bearing and even induce rapid failure especially when excessive temperature growth occurs. The Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China has proposed the 10 MW high-temperature gas-cooled reactor (HTR-10). It is designed to guarantee the inherent safety and economic competitiveness. The dry-lubricated ceramic auxiliary bearing is utilized to protect the AMB and aims to ensure the safety of the AMB system in the HTR-10 in the case of the special operational requirements in the reactor. This paper simulates the process of the rotor drop on the auxiliary bearings in the AMB system of the helium blower of the HTR-10, including the analysis of thermal growth based on the Hertzian contact model and a one-dimensional (1D) thermal heat transfer network model. The study results demonstrate the validation of the bearing models and elucidate different responses between mechanical and ceramic auxiliary bearings during contact events. The research in this paper offers important theoretical bases for the auxiliary bearing design to guarantee the safety of the whole system.