Institute Of Nuclear Energy Technology Research Articles

Preparation of UO2 Kernel for HTR-10 Fuel Element

A 10 MW high temperature gas-cooled reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, China. HTR-10 reached its first criticality successfully in December, 2000 and realized its full power operation in the beginning of 2003. Fabrication technology for HTR-10 spherical fuel element has been established and developed in INET through a lot of R&D activities during the past 20 years. A process known as total gelation process of uranium (TGU) has been developed to prepare UO2 kernel for HTR-10 fuel element in INET. The TGU process combined advantages of both external gelation and internal gelation method of conventional sol-gel process. The sol prescription consists of not only urea, polyvinylalcohol (PVA), tetrahydrofurfuyl alcohol (THFA), but also hexamethylene tetra-amine (HMTA) that is always employed in internal gelation process. The TGU process and facilities are described in this paper. Fuel kernel production for HTR-10 fuel element using this TGU process has completed up to April, 2002. This paper also reported the quality inspection data of kernel production.

Experimental and Operational Verification of the HTR-10 Once-Through Steam Generator (SG)

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor was operated smoothly at the designated parameters. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. Design of the SG includes hydraulics, heat transfer and stability designs. Based on the design requirement, it is necessary to ensure sufficient heat removal from the reactor in order to maintain stable operation. In order to confirm the thermal hydraulic reliability of the SG, a series of experiments had been carried out. The purpose of this paper is to introduce the design features and experimental verification of HTR-10 SG, and the research results of small bending radius helical coil-pipe used in HTR-10, for example, the heat transfer coefficient of water, superheat steam and the two phase flow in the helical tube, the heat transfer coefficient of the He flow across the helical tube, and the centrifugal force effect on the heat transfer for the helical tube. In the paper, some operational experimental data of the HTR-10 SG have been presented.

Experiment and Verification Test of the Once-through Steam Generator of the 10 MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out. The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements. In the paper, some operational experimental data of the HTR-10 S.G have been presented.

A Personal Computer-Based Simulation-and-Control-Integrated Platform for 10-MW High-Temperature Gas-Cooled Reactor

A personal computer—based simulation-and-control-integrated platform for the 10-MW high-temperature gas-cooled reactor (HTR-10), HTRSIMU, has been developed by the Institute of Nuclear Energy Technology (INET) of Tsinghua University in China to meet the requirements of safety analysis, operator training, and control system design. The HTRSIMU runs on a personal computer Windows2000 operating system and consists of three parts: simulation computing system (SCS), man/machine interface (MMI) system, and control system design platform (CDP). Simulation models and equations of the SCS are given, including models of the reactor core, the fuel ball, the primary loop, and the steam generator. Furthermore, functions and characteristics of the MMI and CDP are also described in detail. Moreover, steady state, several typical accidents, and a power control process of HTR-10 are simulated by using the HTRSIMU to demonstrate its simulation and control system design capability.

Preparation of UO2 Kernel for HTR-10 Fuel Element

A 10 MW high temperature gas-cooled reactor (HTR-10) was constructed in Institute of Nuclear Energy Technology (INET) of Tsinghua University, China. HTR-10 reached its first criticality successfully in December, 2000 and realized its full power operation in the beginning of 2003. Fabrication technology for HTR-10 spherical fuel element has been established and developed in INET through a lot of R&D activities during the past 20 years. A process known as total gelation process of uranium (TGU) has been developed to prepare UO2 kernel for HTR-10 fuel element in INET. The TGU process combined advantages of both external gelation and internal gelation method of conventional sol-gel process. The sol prescription consists of not only urea, polyvinylalcohol (PVA), tetrahydrofurfuyl alcohol (THFA), but also hexamethylene tetra-amine (HMTA) that is always employed in internal gelation process. The TGU process and facilities are described in this paper. Fuel kernel production for HTR-10 fuel element using this TGU process has completed up to April, 2002. This paper also reported the quality inspection data of kernel production.

Experiment and Verification Test of the Once-through Steam Generator of the 10MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out. The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements. In the paper, some operational experimental data of the HTR-10 S.G have been presented.

ICONE11-36622 PRELIMINARY DESIGN OF THE POWER CONVERSION UNIT WITH DIRECT GAS-TURBINE CYCLE FOR THE HTR-10

After the first phase of the 10MW high temperature gas-cooled test module reactor (HTR-10) with steam-turbine cycle, the preliminary design of the power conversion unit (PCU) with direct gas-turbine cycle for the next phase of HTR-10 project was finished by the Institute of Nuclear Energy Technology (INET) of Tsinghua University and OKBM (Russia). The main characteristics of the preliminary design, including the PCU structure, the thermal circuit and equipment parameters are described detailed in this paper. Implementation of research program for gas-turbine cycle on the HTR-10 reactor will make it possible to gain necessary experience in gas-turbine cycle design for high-temperature gas-cooled reactor and creation of equipment of large commercial nuclear power plant in the future.

Temperature Measuring System of the In-core Components for Chinese 10 MW High Temperature Gas-cooled Reactor

The 10 MW High Temperature Gas-cooled Reactor (HTR-10) constructed at the Institute of Nuclear Energy Technology (INET), Tsinghua University in China reached its first criticality by the end of 2000. The temperature measuring system of the in-core components is described in this paper. This system consists of the thermocouple penetration assembly of the reactor pressure vessel (TPARPV), the thermocouple penetration assembly of the reactor containment (TPARC) and the distributed computer-based data acquisition and processing system (DCS). Some new techniques were developed and applied, such as the thermocouple penetration technology under the high temperature and high-pressure environment and the laser welding technique. The TPARPV is the key measurement device and is described in detail. The general behavior of the TPARPV and TPARC was confirmed under HTR-10 operating conditions. The helium leakage rate of the TPARPV is 1×10-7Pa-m3/s while the helium leakage rate of the TPARC is less than 1×10-2 Pa-m3/s. The insulation resistance of the sheathed thermocouple is more then 109Ω. The temperature measurement error of the system is 2.3°C. The results of testing and field inspection and operation demonstrate that the design of the temperature measuring system is reasonable and reliable and that the performance of the system satisfies the design requirements of the HTR-10. These new techniques used in the temperature measuring system can be applied not only to other high temperature gas-cooled reactors but to various reactor types as well.

Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor

The Chinese 10 MW high temperature gas-cooled reactor (HTR-10) attained its first criticality on December 21, 2000. The fabrication of the first fuel for the HTR-10 started in February 2000 at the Institute of Nuclear Energy Technology (INET), Tsinghua University. Up to September 2000, a total of 11 721 spherical fuel elements were successfully produced. The average free uranium fraction of the first fuel-determined by the burn-leach method-was 5.0×10 −5. So far, the release rate R/B of the fission gas, measured in the irradiation test, shows that not a single particle in three irradiated spherical fuel elements failed as the results of the irradiation test carried out in Russia. This paper describes the design parameter, the fabrication technology and the performance data of the HTR-10 first fuel, and the production and quality control experiences obtained from the manufacture of the first fuel for the HTR-10.

Prediction calculations and experiments for the first criticality of the 10 MW High Temperature Gas-cooled Reactor-Test Module

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.

Coupling of nuclear heating reactor with desalination process

A seawater desalination plant using a nuclear heating reactor (NHR) coupled with the multi-effect distillation (MED) process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. The seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy sources are severely lacking. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the NHR-5 experiences. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR and MED desalination systems. With the power range of the heating reactor within 10 to 200 MWt, the desalination plant could provide 8,000 to 160,000 m 3/d of potable water of appropriate quality. The design concept and parameters, safety features, and comparative investigation of coupling schemes are presented in the paper.

Temperature Measuring System of the In-core Components for Chinese 10 MW High Temperature Gas-cooled Reactor.

The 10 MW High Temperature Gas-cooled Reactor (HTR-10) constructed at the Institute of Nuclear Energy Technology (INET), Tsinghua University in China reached its first criticality by the end of 2000. The temperature measuring system of the in-core components is described in this paper. This system consists of the thermocouple penetration assembly of the reactor pressure vessel (TPARPV), the thermocouple penetration assembly of the reactor containment (TPARC) and the distributed computer-based data acquisition and processing system (DCS). Some new techniques were developed and applied, such as the thermocouple penetration technology under the high temperature and high-pressure environment and the laser welding technique. The TPARPV is the key measurement device and is described in detail. The general behavior of the TPARPV and TPARC was confirmed under HTR-10 operating conditions. The helium leakage rate of the TPARPV is 1×10-7Pa-m3/s while the helium leakage rate of the TPARC is less than 1×10-2 Pa-m3/s. The insulation resistance of the sheathed thermocouple is more then 109Ω. The temperature measurement error of the system is 2.3°C. The results of testing and field inspection and operation demonstrate that the design of the temperature measuring system is reasonable and reliable and that the performance of the system satisfies the design requirements of the HTR-10. These new techniques used in the temperature measuring system can be applied not only to other high temperature gas-cooled reactors but to various reactor types as well.

Experiment Study on Rod Ejection Accident of Hydraulic Control Rod Driving System

The rod ejection experiment has been done in the 1:1 experiment loop of the hydraulic control rod driving system (HCRDS) of the 200 MW nuclear heating reactor (NHR-200). The result indicated that under different experiment conditions, the control rod was not ejected when the pressure vessel of the loop lost pressure. Based on the depressurization accident analysis of the NHR-200, performed at the Institute of Nuclear Energy Technology (INET), Tsinghua University, the depressurization rate of experiment is much higher than that of analysis. Thus, the HCRDS has inherent safety and no hidden trouble of rod ejection when the depressurization accident happens.

Research Method and Two-Phase Flow Stability of the Steam Generator of HTR-10

A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute ofNuclear Energy Technology (INET) of Tsinghua University is being constructed now. The steam generator (SG) of the HTR-10 is one of the most important facilities for reactor safety. In order to investigate the thermal-hydraulic performance of the SG, a full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 in detail. The test assembly of the SGTM-10 simulates practical thermal and structural parameters of the HTR-10. The SGTM-10 consisted of three separated loops, primary-helium loop, secondary-water loop, and third-cooling water loop. There are two parallel tubes arranged in the test assembly. The main experimental equipment is shown in this paper. Analysis shows that for once-through steam generator simulation experiment, the electric-heated simulation method could not match practical operating condition. The results may not reflect true phenomena. The main results of experiments, for example effects of the outlet pressure, effects of the heating power, effects of the inlet sub-cooling are described. Experiments indicated, when the heat load of the HTR-10 is more than 30% the SG will be stable.

Experiment Study on Rod Ejection Accident of Hydraulic Control Rod Driving System.

The rod ejection experiment has been done in the 1:1 experiment loop of the hydraulic control rod driving system (HCRDS) of the 200 MW nuclear heating reactor (NHR-200). The result indicated that under different experiment conditions, the control rod was not ejected when the pressure vessel of the loop lost pressure. Based on the depressurization accident analysis of the NHR-200, performed at the Institute of Nuclear Energy Technology (INET), Tsinghua University, the depressurization rate of experiment is much higher than that of analysis. Thus, the HCRDS has inherent safety and no hidden trouble of rod ejection when the depressurization accident happens.

Step Dynamic Process of the Hydraulically-driven Control Rod, (II)

The HCRDS (hydraulic control rod driving system) is a new type of control rod driving system, which is designed by INET (Institute of Nuclear Energy Technology) and has been put into use in 5 MW nuclear heating reactor in Tsinghua University. The purpose of this paper is to theoretically analyze the step-down process of this new technology and establish fundamental basis for further analysis and research. The experimental loop of the HCRDS and the working principle on the step-down process are introduced in this paper. The theoretical model is established on the basis of analysis, simplification and hypothesis. Also given is the accurate mathematical description of this model. The comparison between the results of this model and that of the experiment proves the rationality and feasibility of the model. The selection of the working point is also introduced.

Experimental research on the gravity-driven boron injection system for a 200 MW nuclear heating reactor

The gravity-driven boron injection system (GDBIS), designed by the Institute of Nuclear Energy Technology (INET) of the Tsinghua University, PR China, is a new type of passive system to be applied in the 200 MW nuclear heating reactor (NHR-200), also designed by INET. The function of this system is to shut down the reactor in an emergency, in case control rods do not operate properly. A borate water tank is located 10 m above the top of the pressure vessel. When the pressure of the reactor and the boron tank balances, the borate water will be driven by gravity to flow into the reactor, and thus shut down the reactor. The thermal hydraulic performances of the system for cold (room temperature nitrogen) and hot (mixture of hot steam and nitrogen) operating conditions, especially the response time of pressure and water injection, have been researched under different initial conditions. Firstly, several factors, e.g. orifice on steam lines, and the volume ratio of the gas–steam spaces of the reactor and the boron tank, have effects on the pressure and water injection response time and other thermal hydraulic performance of the system. Secondly, the steam and liquid communication modes, namely the acting time and sequence of the action of valves connecting steam and liquid lines, have great influences on the performance of the system. Thirdly, the limited pressure balance time (about 1.0 s) can be achieved under the cold condition. This investigation shows that GDBIS can be properly used in the 200 MW nuclear heating reactor.

Experimental Study of Two-Phase Flow Oscillation in Natural Circulation

The experiment was performed on the test loop HRTL-5, which simulates the geometry and system design of the 5-MW nuclear heating reactor developed by the Institute of Nuclear Energy Technology, Tsinghua University. The flow behavior for a wide range of inlet subcoolings, in which the flow experience varies from single- to two-phase, is described in a natural circulation system at different pressures (p = 0.1, 0.24, and 1.5 MPa). Several kinds of flow instability are investigated, including geysering, flashing-related flow instability, and high-frequency flow oscillation at p = 0.1 and 0.24 MPa, as well as low steam quality density wave oscillation at p = 1.5 MPa. The mechanisms of geysering, which has new features, and flashing-related flow instability, which has never been studied well enough in this field, are particularly interpreted. The experimental results show the following: First, for a low-pressure natural circulation system, the two-phase flow is unstable in most inlet subcooling conditions, and the two-phase stable flow can be reached only with very low inlet subcoolings. Second, at high inlet subcoolings, the flow instability is dominated by subcooling boiling in the heated section, and at intermediate inlet subcoolings, it is dominated by void flashing in the adiabatic long riser. Third, in the two-phase stable flow region, the conditions for boiling out of the core, namely, single-phase flow in the heated section and two-phase flow in the riser due to vapor flashing, can be realized. The experimental results are of significance for the design and accident analysis of vessel and swimming pool–type natural circulation nuclear heating reactors.

Experimental investigation on the pressure response characteristic of the gravity driven boron injection system / Experimentelle Untersuchung zur Druckantwortcharakteristik eines gravitationsgetriebenen Boreinspeisesystems

The gravity driven boron injection system (GDBIS), designed by the Institute of Nuclear Energy Technology (INET) of the Tsinghua University, PR China, is a new type of passive system to be applied in the 200 MW nuclear heating reactor also designed by INET, The purpose of this system is to shutdown the reactor in an emergency, in case the control rods do not operate properly, The pressure response time and the thermal hydraulic performance of the system have been investigated under different initial conditions. The research results for cold (room temperature nitrogen) and hot (mixture of hot steam and nitrogen) operating conditions are presented. Firstly several factors, e. g. the flow resistance character of the steam and liquid lines, and the volume ratio of the upper spaces of reactor and boron injection tank, affect the pressure response time and the thermal hydraulic performance of the system. Secondly, the steam and liquid communication modes, namely, the time and sequence of the action of valves connecting steam and liquid lines, have great influence on the performance of the system. Thirdly, the limited pressure balance time (about 1.5∼2.0 s) can be achieved under cold operating conditions. This investigation shows that GDBIS can be properly used in the 200 MW nuclear heating reactor.

Investigation on two-phase flow stability in a natural circulation system / Untersuchung der Stabilität einer Zweiphasenströmung in einem System mit Naturumlauf

A research program on thermal-hydrodynamic stability of the two-phase flow, simulating the behavior in the primary loop of a nuclear heating reactor developed by the Tsinghua University Institute of Nuclear Energy Technology (INET) in China has been executed for several years. In the integrated primary loop of the NHR heating reactor, the natural circulation of the coolant water was adopted with low system pressure, low steam quality at the exit of the core and a relatively long riser above the core. It is important to keep the reactor operating under stable conditions with enough safety margins. The program was aimed at: (1) accumulating experimental data for verification of the models and codes used in the design and safety analysis of this type of reactor; (2) understanding the unstable behavior, its physical mechanisms and parameter effects. The results of the study show that under certain geometric conditions and operating parameters a self-sustaining, low frequency, even amplitude mass flow oscillation may be excited at very low steam qualities. Stability maps under different conditions are provided. An unstable region, which exists between the single-phase stable region and the low steam quality bulk boiling stable region, was experimentally demonstrated.