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Session Chair: Grant Hawkes, Idaho National Laboratory, United States of America Session Chair: Changho Lee, Argonne National Laboratory-ANL, United States of America
Thermal Model Heat Rate Predictions of the AGR-5/6/7 Experiment
Grant Hawkes, James Sterbentz
Idaho National Laboratory
The AGR-5/6/7 experiment is currently being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory and is approximately 50% complete. Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 is testing low-enriched uranium oxycarbide TRISO fuel. The AGR-5/6/7 test train has 5 capsules with thermocouples and independent gas control mixtures.
Unique to this paper is the ability to use a curve fit of each fuel compact’s heat generation rate (HGR) from previous ATR operating cycles in order to predict HGRs for future operating cycles. Typically, the physics calculations use daily as-run conditions for completed ATR operating cycles. This method bypasses waiting for the detailed daily as-run calculations and predicts the HGRs during the current or future ATR cycle. These predicted HGRs are used in the thermal model to calculate the helium/neon gas fraction necessary to achieve the target fuel temperatures. This method of calculating and predicting fuel and capsule temperatures is especially useful for capsules with no functioning thermocouples (TCs). The as-run HGRs are calculated using complicated physics models based on Monte Carlo techniques. These HGRs are then used to generate curve fits and project HGR data for future ATR cycles.
Temperature Measuring Ball Sorting Techniques during the Maximum Temperature Measurement Test of HTR-10
Ling Liu, Feng Xie, Yucheng Wang, Dongmei Ding, Liqiang Wei, Tao Ma, Xiaoming Chen, Zhihui Li
During the maximum temperature test of HTR-10, a kind of measuring balls with metal block inside are loaded into the reactor core to measure the maximum temperature. The general parameters of measuring ball are similar with the fuel elements and graphite balls in the core. It very difficult to sorting the measuring ball among these balls. Several ingenious techniques are developed to solved this problem--- dose rate comparison method，neutron activation & γ spectrum analysis, and X ray radioscopy.
Steady-state thermal fluids analysis for the HTR-PM equilibrium core
Fubing Chen, Zihao Han
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 gas-cooled 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. Among various aspects of the reactor design, the thermal hydraulics design is of extreme importance for the evaluation of the reactor performance under normal operation. 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 level 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 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.
Safety features evaluation of Stirling integrated gas cooled reactor
1Beijing Institute of Spacecraft Environment Engineering; Science and Technology on Reliability and Environmental Engineering Laboratory; 2Tsinghua University; Key laboratory of Advanced Reactor Engineering and Safety, Ministry of Education
The space nuclear reactors are effective and promising energy sources for the planetary and deep space exploration. The Stirling integrated space nuclear reactor (autonomous circulation miniature integrated reactor, ACMIR) integrates the reactor core and the energy conversion system in a pressure vessel, which has the advantages of high specific power and efficiency, satisfactory performance and high reliability. In this paper, the safety feature of AMCIR is evaluated based on the neutronic characteristics study of a 5 kWe ACMIR reactor core. The reactivity temperature coefficient, control drum capability, criticality safety during launch accident, neutron flux and power density distribution are calculated by the Monte Carlo code. The results show the safety features of ACMIR, which supports the detailed design and development of the space nuclear reactor.