High Temperature Gas-cooled Test Reactor Research Articles

Fabrication and Properties of Boron Carbide/Graphite Ceramic Pellets

Boron carbide (B4C)/graphite (C) ceramic pellets containing 25 wt% B4C as an ideal structural component in Chinese 10 MW high temperature gas-cooled test reactor (HTR-10) were successfully prepared by a two-stage operation method at low temperature. Initially, the samples were pressurelessly sintered at 1373 K and subsequently infiltrated with phenolic resin, pyrolysed at 1473 K, then coated with carbon by chemical vapour deposition at 1273 K. Morphology and structure of the samples were analysed by SEM and X-ray diffraction. The mechanical properties of the sintered bodies were also investigated. Superior properties were obtained due to the formation of pyrolytic carbon during the two-stage operation process. Density and compression strength of the samples were 1.94 g.cm–3 and 620 N, respectively. In addition, the rate of abrasion and moisture resistance were near zero for the samples.

Temperature Response of the HTR-10 during the Power Ascension Test

The 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor in China. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned and performed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. In this study, the power ascension process is preliminarily simulated using the THERMIX code. The code satisfactorily reproduces the reactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature. Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples. THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreement between the calculated temperatures and the measured ones. Based on the comparison results, the THERMIX simulation capability for the HTR-10 dynamic characteristics during the power ascension process can be demonstrated. With respect to the reactor safety features, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower than the fuel temperature limit of 1620°C.

The design and thermohydraulics study of the HTR-10 High Temperature Helium Experimental Loop

A High Temperature Helium Experimental Loop (HTHEL) for the purpose of studying the transportation and deposition behavior of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is studied and designed. Through the optimal design based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly and simulate the physical environment of the heat transfer tube of SG in the HTR-10 in the sense of thermohydraulics, which could be used to study the original source term in the primary circuit. The simulation of the graphite dust particle trajectories in the DSD are shown and it is elucidated that DSD could also be used to study the behavior of graphite dust in the high temperature helium in the SG of HTR, which will provide deeper understanding for the analysis of source terms of HTR-10.

Technical design and principle test of active magnetic bearings for the turbine compressor of HTR-10GT

The 10MW high temperature gas-cooled test reactor coupled with gas-turbine circle (HTR-10GT) has been carried out by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China since year 2002. In the power convention unit (PCU) of the HTR-10GT, the contact-free and no-lubricating active magnetic bearings (AMB) are adopted to support the vertical high-speed turbine machine shaft, because of their numerous advantages over the conventional oil mechanical bearings under the special reactor operating conditions. Based on the previous studies and small tests of the AMBs, the final engineering design of the AMBs for the HTR-10GT turbine compressor rotor has been finished recently. This paper firstly introduces the design principle and technical futures of the AMBs. Then the main structure, as well as the key dimensions and parameters of the different AMB components are illustrated in detail. Besides, the rotor dynamic analysis is conducted to provide the mathematic model for the unbalance attenuation with high performance control system design in order to smoothly pass across the first two bending critical speeds before reaching the rated speed of 15,000rpm. Furthermore, a small AMB flexible test system in compliance with the dynamic similarity principle of the helium turbine compressor rotor of the HTR-10GT, has been set up to validate the advanced control algorithms and accumulate design and operation experiences for the next full scale experiment in the near future.

Benchmark Calculation for the Steady-State Temperature Distribution of the HTR-10 under Full-Power Operation

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230°C. According to the comparison results of code-to-code as well as code-toexperiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem.

Current status and technical description of Chinese 2 × 250 MW th HTR-PM demonstration plant

After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MW th high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10. Due to intensive engineering and R&D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled. The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MW th pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW. A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line. The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures. According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started. After the successful operation of the demonstration plant, commercial HTR-PM plants are expected to be built at the same site. These plants will comprise many NSSS modules and, correspondingly, a larger turbine.

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.

Benchmark Calculation for the Steady-State Temperature Distribution of the HTR-10 under Full-Power Operation

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230°C. According to the comparison results of code-to-code as well as code-toexperiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem.

06/01214 The feasibility study on perfect burning reactor system: Nakae, N. Progress in Nuclear Energy, 2005, 47, (1–4), 139–146

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes. Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results. Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests. The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW. The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.

Fuel irradiation of the first batches produced for the Chinese HTR-10

An irradiation test of four spherical fuel elements (SFE) had been performed in the Russian reactor IVV-2M. The elements were sampled randomly from the first and second product batches which were manufactured for the 10 MW high-temperature gas-cooled test reactor (HTR-10). The maximum burnup of the irradiated fuel elements reached 107,000 MWd/tU and the maximum fast neutron fluence was 1.31 × 10 25 m −2. The release-to-birth rate ratio ( R/ B) did not increase significantly during irradiation. However, an in-pile heating-up test of element SFE 7 in Capsule 5 led to a failure of approximately 6% of the coated particles. After the test it was estimated that the fuel temperature had very likely been much higher than the intended 1600 °C.

Temperature effect on IG-11 graphite wear performance

IG-11 graphite, used in the 10 MW high temperature gas-cooled test reactor (HTR-10), was tested under different temperatures on an SRV standard wear performance tester. The experiment temperatures were room temperature, 100, 200, 300 and 400 °C. According to the reactor structure, the experiments were designed to test graphite–graphite and graphite–stainless steel wear. The wear debris was collected, and the worn surfaces and debris were observed under scanning electronic microscope (SEM). It was found that there were different wear mechanisms at different temperatures. The main wear mechanism at room temperature was abrasive wear; at 200 °C, it was fatigue wear; at 400 °C, adhesive wear was observed. This difference was mainly due to the change of stress distribution at the contact area. The distribution of wear debris was also analyzed by EDX particle analysis software.

High temperature reactor development in China

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.

Economic pre-feasibility study of seawater desalination for a high-temperature gas-cooled reactor by reverse osmosis

The comprehensive study of high-temperature gas-cooled reactors (HTR) has been carried out for many years. In particular, the Chinese 10 MW(t) high-temperature gas-cooled reactor test module (HTR-10) has attained initial criticality since December 20, 2000. The HTR-10 design represents the features of modular HTGR design. The reactor core and steam generator are housed in two steel pressure vessels, which are arranged side by side. These steel pressure vessels are in touch with cold helium of about 250°C coming out from the circulator which sits over the steam generator tubes in the same vessel. A feasibility study for a four-module 200 MW(t) HTR, generating about 300 MW of electricity, has recently been performed by INET, Tsinghua University. This paper presents the economic feasibility of a seawater desalination system for HTR by reverse osmosis (RO)(HTR-300), that produces 480,000m 3/d of freshwater, which was calculated using the Desalination Economic Evaluation Program (DEEP2.0) developed by the IAEA. The results indicated that the averaged discounted production of water cost is $0.571m 3 and the averaged discounted electricity generation cost is $0.038 /kWh, which is lower than that of a nuclear and fossil power plant for desalination plants. It has the advantages of its low electricity cost and water production cost for its inherent safety features of HTR, compared to other same-scale nuclear power and fossil power plants. The sensitive factors for the cost of water desalination and electricity generation were analyzed, which include netthermal efficiency, discount rate/interest rate, specific investment cost and seawater temperature/seawater total dissolved solid (TDS). The discount rate is one of the factors that has the greatest effect on the cost of water and electricity, which are $0.42/m 3 and $0,028/kWh, $0.48/m 3 and $0.032/kWh, $0.571m 3 and $0.038/kWh, W65/m 3 and $0.043/kWh, $0.731m 3 and $0.048/kWh, respectively, while the discount rate/interest rate varies in the range of 3%,5%, 8%,10% and 12%. The water cost is strongly influenced by the TDS of the seawater, but not influenced by seawater temperature. The reference net thermal efficiency is another factor that has a great effect on water and electricity cost: $0.59/m3 and $0.043/kwh at 35%,$0.56W and $0.034/kWh at 47.7%

Prospects for economic fusion electricity : Cook, I. et al. Fusion Engineering and Design, 2002, 63–64, 25–33

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.

Digital signal processing on the fuel ball counting system for HTR-10

It is critical for the normal operation of the 10 MW high-temperature gas-cooled test reactor (HTR-10) to accurately monitor the number and direction of the fuel balls passing through the fuel handling system. This is the role for the fuel ball counting system. Due to the complexity of the signals from the fuel counters, the digital signal-processing method is required in order to count the number of the fuel balls accurately, continuously, and automatically. This paper focuses on the method and algorithms of the signal processing and implementation of the system, i.e. how to search the signal peaks, how to deal with the floating and varying signals, how to eliminate the noises and pseudo peaks, how to analyze the time relationship, and how to coordinate the functions of sampling signals, counting fuel balls, communicating with PLC, displaying all results in a windows-based system, etc.

Design and full scale test of the fuel handling system

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m −1 °C −1 under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.

Thermal hydraulic instrumentation system of the HTR-10

The 10 MW high temperature gas-cooled reactor—test module (HTR-10) has many different characteristics compared with the pressure water reactor (PWR), such as special fuel elements, core construction, helium coolant and so on. Thus, the design of thermal hydraulic instrumentation system of the HTR-10 is also different from that of a PWR. This paper describes the design criteria and system integrating method of thermal hydraulic instrumentation systems in details. The main thermal instrumentation and devices specially developed for the HTR-10 are also introduced.

Simulating test for thermal mixing in the hot gas chamber of the HTR-10

The helium coolant at the outlet of the pebble bed core of the 10 MW High Temperature Gas-cooled Reactor-Test Module exhibits a severe radial temperature deviation. In order to avoid damages at the downstream components due to alternating thermal loads such as the steam generator, a hot gas chamber is especially designed to solve the problem. Thermal mixing performance of the coolant in the hot gas chamber is experimentally investigated on a 1:1.5 scale model by air. The experimental result shows that within the Reynolds number range of 1.4×10 5–5.8×10 5, the hot gas chamber with a radial mixer reaches excellent thermal mixing of the coolant of about 94%. The flow resistance coefficient for the hot gas chamber is also presented.