Design and Performance of Hybrid Control Rod For Passive IN-core Cooling System

Abstract

Protection of the public and the environment from undue radiation hazards is a definition of nuclear safety. Although there are various safety systems in nuclear power plants to achieve the nuclear safety, Fukushima-Daiichi accident showed the vulnerabilities of the installed safety systems. After the Fukushima accident, various passive safety systems and strategies are under development to cope with the postulated accidents. The majority of passive safety systems concentrated to inject emergency core coolant (ECC) or feedwater with the circuits comprise many pipelines and valves. In station blackout condition, the pressure of reactor vessel would be higher than the ECC injection pressure resulting in failure of ECC supply and eventually causing core damage. The reliability issues about the performance of passive safety systems have been discussed owing to their high uncertainties, low performance, and lack of experience in operation compared to active safety systems. In aspect of probabilistic safety, complex circuits which comprise many valves and pipelines have possibilities of single failure and common cause failure. Development of innovative passive safety system having differentiated working principle, significant performance, and low possibility of failure can enhance reactor safety providing solutions for the aforementioned problems. Based on these requirements, hybrid control rod which combines the functions of control rod and heat pipe was proposed for the development of passive in-core cooling system (PINCs). The control rods drop to the core using gravity and shutdown the reactor by neutron absorption. The thermosyphon heat pipe is a passive heat transfer device using phase change and convection of working fluid in a closed metal container having two different temperature interfaces (evaporator and condenser). The combination of thermosyphon and control rod, hybrid control rod can achieve reactor shutdown and decay heat removal simultaneously at accident conditions. Hybrid control rod was designed considering the aspects of neutronics (reactivity worth) and mechanical integrity. Most of the nuclear reactors operate at high temperature and high pressure environment with high power density. Thus, pressure control strategies of the hybrid control rod using non-condensable gas and expansion of the working fluid were established to achieve high decay heat removal capacity and operating conditions. The designed hybrid control rods were equipped on the experimental facility and their thermal performances were studied under various amount of working fluid, amount of non-condensable gas, and operating pressures of the test section. The experimental results showed relations between heat transfer characteristics and controlled parameters. Controlling operating condition of hybrid control rod in high pressure worked successfully, and the proportionality between maximum heat removal capacity and operating pressure of hybrid control rod design has been proven. Measured maximum heat transfer rate of single hybrid control rod was 6 kW at 20 bar. Simulations of multi-dimensional analysis for reactor safety (MARS) code were also performed to validate the experimental results and evaluate the prediction capability of the code on the hybrid control rod. The simulation results showed the limits of heat transfer models in the code analyzing the hybrid control rod in which the boiling and condensation heat transfer occurs simultaneously in a manner of countercurrent flow. The experimental results were compared with several models associated with boiling heat transfer, condensation heat transfer, and critical heat flux (CHF) of thermosyphon for the development or the selection of optimal models. The selected models could be implemented to system analysis codes in the purpose of deterministic safety assessment of PINCs against design basis accidents. Imura’s correlation, which was developed in two-phase natural convection condition and validated with experiments in wide range, was selected as boiling heat transfer model of pressurized hybrid control rod. The existing condensation models were based on Nusselt’s film condensation theory. Hence, the effect of non-condensable gas and perturbation between upward vapor flow and downward liquid film flow were not considered at the same time. The change of effective heat transfer length due to presence of non-condensable gas and effect of fluid inertia were considered for the derivation of new condensation model. The main thermal-hydraulic phenomenon which induces CHF of thermosyphon is flooding. The flooding-based CHF models for thermosyphon were derived with theories on instability of the liquid film or maximum liquid film flow rate in countercurrent flow condition. The limited prediction capabilities of the models were attributed to difference between hydraulic diameter and heated diameter as well as high operating pressure. Consequently, new model regarding the CHF of hybrid control rod was suggested to explain its unique characteristics. The hybrid control rod could be equipped on spent fuel dry storage casks for the extension of their thermal margins. The mock-up was designed to be scaled-down to 1/10 of metal dry storage cask developed by NAC. The effect of hybrid control rod on thermal margins of the cask was experimentally studied. The equipment of hybrid control rod with installation of heat sink lid reduced the temperature distributions inside the cask at equal power density condition. Application of hybrid control rod could extend the thermal margin up to 30 %. Feasibility of PINCs based on experimentally and analytically studied hybrid control rods were discussed according to commercial reactors. A number of nuclear facilities has been built to supply and manage energy. The nuclear fuels generate decay heat even in shutdown condition by fission products. Management of the decay heat is important to satisfy demand for nuclear safety. Therefore, new conceptual safety system is required to supplement the issues on existing safety systems. Passive in-core cooling system based on hybrid control rod is the effective way to be applied on extensive nuclear facilities containing nuclear fuels. Pressurized hybrid control rod could meet the operating conditions of application objects with significant decay heat removal capacity.