Zuoyi Zhang

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.

Effects of reaction temperature and inlet oxidizing gas flow rate on IG-110 graphite oxidation used in HTR-PM

ABSTRACT The oxidation behavior of a selected nuclear graphite (IG-110) used in Pebble-bed Module High Temperature gas-cooled Reactor was investigated under the condition of air ingress accident. The oblate rectangular specimen was oxidized by oxidant gas with oxygen mole fraction of 20% and flow rates of 125–500 ml/min at temperature of 400–1200u2009°C. Experiment results indicate that the oxidation behavior can also be classified into three regimes according to temperature. The regime I at 400–550u2009°C has lower apparent activation energies of 75.57–138.59 kJ/mol when the gas flow rate is 125–500 ml/min. In the regime II at 600–900u2009°C, the oxidation rate restricted by the oxygen supply to graphite is almost stable with the increase of temperature. In the regime III above 900u2009°C, the oxidation rate increases obviously with the increase of temperature. With the increase of inlet gas flow from 125 to 500 ml/min, the apparent activation energy in regime I is increased and the stableness of oxidation rate in regime II is reduced.

A Personal Computer-Based Simulator for Nuclear-Heating Reactors

Abstract A personal computer (PC)-based simulator for nuclear-heating reactors (NHRs), PC-NHR, has been developed to provide an educational tool for understanding the design and operational characteristics of an NHR system. A general description of the reactor system as well as the technical basis for the design and operation of the heating reactor is provided. The basic models and equations for the NHR simulation are then given, which include models of the reactor core, the reactor coolant system, the containment, and the control system. The graphical user interface is described in detail to provide a manual for the user to operate the simulator properly. Steady state and several transients have been simulated. The results of PC-NHR are in good agreement with design data and the results of RETRAN-02. The real-time capability is also confirmed.

Progress of the HTR-10 Measured Data Utilization

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), which was designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is the first High Temperature Gas-cooled Reactor (HTGR) in China. Stepping into the commissioning phase in April, 2000, the HTR-10 attained the first criticality in December, 2000 and achieved its full power operation in January, 2003. Up to now, the HTR-10 has been successfully operated for more than ten years with different power levels. During the relatively long period of commissioning and operation, various kinds of tests were carried out on this reactor. Within the scope of this paper, the commissioning stage, the operation history and the test implementation of the HTR-10 are briefly summarized.At this stage, the HTR-10 is the only pebble bed HTGR under operation in the world, so the measured data from this reactor are extremely valuable for verifying the inherent safety features incorporated in small modular HTGRs as well as for testing the computer programs employed in the HTGR design process. With the purpose of ensuring the code credibility, validation work using the HTR-10 operation and test data has been performed for several years in INET. What is more, these data were partly shared with different countries through some collaborative research projects related to code development and assessment. In this paper, progress of the HTR-10 measured data utilization is reviewed. Meanwhile, existing problems observed from the code-to-test as well as code-to-code comparisons are pointed out. In addition, possible reasons of such problems are discussed in detail.Copyright

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on 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 characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data 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 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.Copyright

Characterizing thermal-oxidation behaviors of nuclear graphite by combining O 2 supply and micro surface area of graphite

The effects of different parameters on oxidation rate are non-linear, interactive and diversified in which the change of adequacy of O2 supply is an important indicator. The influence of microstructure on oxidation rate became stronger worsening the fitting linearity to calculate the activation energy based on present method with the decreased adequacy of O2 supply due to the increase of temperature, the decrease of gas flow rate, etc. Here, we proposed a method to characterize thermal-oxidation behaviors of nuclear graphite by combining O2 supply and micro surface area of graphite. The proposed method improved the linearity and reduced the standard error of Arrhenius plots of oxidized graphite IG-110 (10 L/min reactant gas) and ET-10 (0.2 L/min reactant gas). The value of activation energy of graphite IG-110 oxidized under ASTM D7542 condition is calculated as 220u2009kJ/mol by this method echoing the results of previous studies with sufficient O2 supply. For the conditions with less O2 supply at low gas flow rate and/or high temperature, the change of microstructure of oxidized graphite should be obtained as an important factor influencing oxidation rate of graphite.

Thermal Mixing Performance of Mixing Structure at Reactor Outlet of HTR-PM under Complex Conditions

In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the Thermal Mixing Performance (TMP) of the thermal mixing structure at Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) core outlet. The experiments on Test Facility - Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation. Simulation calculations are conducted for the mixing structure of TF-HGM and HTR-PM. According to the results of experiments and CFD simulation, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, under all considered conditions. In addition, the TMP obtained by the temperature difference between main flow and bypass flow is suitable for the experiment and the simulation results while the new definition of TMP integrating specific heat, flow rate, inlet position and thermal uniformity needs further improvement.