Experimental Study of a Stoppage Natural Circulation during a Nuclear Heating Reactor LOCA

The 5MW nuclear heating reactor is an integral natural circulation reactor. The rupture of the coolant penetrating tube is a typical accident causing coolant loss. When the water level drops down to the upper edge of the inlet of the heat exchanger, the natural circulation stops. This influences the core cooling and the stability of the main loop. A series of tests showed that there is a stable drop of pressure, and the heated element temperature is not too high to cause burnout. But the backward flow or flow oscillation in the primary coolant circuit occurs when the flow breaks completely in the end. The whole flow process is described and the mechanism is discussed.

Research and Development of Nuclear Heating Reactors in China

The research and development (R & D) of nuclear heating reactors (NHRs) have been conducted as one of the national key projects in science and technology in China since the 1980s. This paper presents the development status. main design featur and safety concepts of the NHR.

Thermal-hydraulic Stability Analysis of Nuclear Heating Reactors

The two-phase flow instability that can occur in a natural circulation system is of importance in the design of nuclear heating reactors. The time domain code RETRAN-02 and the frequency domain code NUFREQ were applied to estimate the instability boundary and the effects of such parameters as pressure, inlet resistance and riser height in NHR-5 and an experimental loop. The results of the calculations and the experiments are in good agreement. Nonlinear density wave oscillations were analyzed using the RETRAN-02 code. The theory of nonequilibrium thermodynamics was used to find an explicit criterion to estimate the threshold of the stability. Experimental simulation of the nuclear feedback density wave instability was also carried out in a test loop using. computer controlled electric power.

Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss－of－Coolant Accidents

The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. UP to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as folios: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

Loss of Coolant Experiments for the Test Nuclear Heating Reactor

A series of tests were completed for three types of loss of coolant accidents (LOCAs) (pipe break in the gas plenum. near the liquid level and submerged under water) in the test nuclear heating reactor (NHR). Experiments show that the three cases of LOCAs (loss of coolant accidents) have different patterns. In the case of a pipe break connected to the gas plenum, the quantity of water lost is independent of the diameter of the broken pipe. In the case of a pipe located near the liquid level. the quantity of water lost depends on the location of the pipe. In the case of a pipe break below the water level. all the water above the break will be discharged. The discharge patterns for all three cases are analyzed in detail.

Dynamic Programming Method to Optimize Control Rod Positions in NHR-200

A code. OPTROD. based on dynamic programming method is developed and utilized to optimize control rod positions throughout the fuel cycle of the 200 MW nuclear heating reactor (NHR-200). A fuel cycle is divided into many steps or stages. Optimization of the multistage process is solved iteratively in the forward direction throughout a fuel cycle. The dynamic programming method is much more efficient than the normal nonlinear programming method. Convergence is obtained even if poor initial control rod positions are given.

Fuel Assembly Arrangement Optimization for NHR-200

This study considers optimization of the fuel assembly arrangement in the initial core loading of the 200 MW nuclear heating reactor (NHR-200). The enrichment of the fuel assemblies is used as the control variable with the objective to minimize the power peaking factor. The optimization methods are applied indirectly because it is difficult to directly relate the control variable and the object function in a single equation. Therefore, the solution uses linearized functons which are solved with linear programming. The corrected simplex method is used to solve the optimal problem. Useful engineer software has been designed and used in reactor physics design.