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Abstract: Bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in nuclear reactors. Liquid droplet impingement (LDI) erosion could be regarded as one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plant safety. In this paper a computational methodology is established for simulation of LDI erosion using computational fluid dynamics (CFD) simulation and theoretical calculation. Two-phase flow numerical simulations are conducted for standard elbow geometry, typically with the pipe diameter of 170mm. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ɛ turbulence model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The turbulence damping in vapor–droplets flow is theoretically analyzed by a damping function on the energy spectrum basis of single phase flow. Locally, a droplet impact angle function is employed to determine the overall erosion rate. Finally, the overall and local investigations are combined to purpose a general methodology of LDI erosion prediction procedure, which has been complemented into CFD code. Based on our more physical computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena. [Copyright &y& Elsevier]

Abstract: Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air–water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam–water annular flow conditions. In each experiment, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using the liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant [Sawant, P.H., Ishii, M., Mori, M., 2008. Droplet entrainment correlation in vertical upward co-current annular two-phase flow. Nucl. Eng. Des. 238 (6), 1342–1352] for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air–water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air–water, helium–water and air–genklene experimental data measured by Willetts [Willetts, I.P., 1987. Non-aqueous annular two-phase flow. D.Phil. Thesis, University of Oxford]. However, comparison of the correlations with the steam–water data available in literature showed significant discrepancies. It is proposed that these discrepancies might have been caused due to the inadequacy of the liquid film extraction method used to measure the entrainment fraction or due to the change in mechanism of entrainment under high liquid flow conditions. [Copyright &y& Elsevier]

• Evaluation of decay heat removal systems of FHR during thermal-hydraulic transient was performed. • Heat removal performances of FHR systems are investigated. • FHR's decay heat removal system shows sufficient heat removal capability to avoid catastrophic accidents. The Fluoride-salt-cooled High-temperature Reactor (FHR) is one of the Generation IV nuclear reactor concepts being investigated mainly in the U.S. and China. Liquid salt with high boiling point of over 1400 °C is used as a coolant and its baseline fuel is the graphite-matrix coated-particle fuel developed for high-temperature gas-cooled reactors. To establish advanced safety systems in FHR, passive decay heat removal system which utilizes fluid's natural circulation is considered. Three systems are proposed: Direct Reactor Air Cooling System (DRACS), Reactor Vessel Air Cooling System (RVACS), and Silo Cooling System (SCS). Although these three safety systems were previously developed for other types of reactors, their heat removal performances need to be evaluated for the FHR plant. The current study provides evaluation of decay heat removal systems of FHR during thermal-hydraulic transient. In addition, FHR's grace period to start the reactor cooling was determined under the condition where no operable heat removal systems are available. The current study shows that FHR's decay heat removal systems possess sufficient heat removal capability to avoid catastrophic accidents. The construction of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently being planned and its development is a key step to realize the commercialization in near future. This study provides performance evaluation of decay heat systems for FHTR under thermal-hydraulic transient, and also suggests the adequate operation procedure suitable for FHTR to avoid a severe accident. [ABSTRACT FROM AUTHOR]

In this paper, experimental investigation of the molten metal jet's colliding and spreading behaviors on the flat steel surface covered with water layer was carried out. High-frequency induction heating system was utilized to produce the molten metal sample and it was released to the wet surface from a fixed elevation. As the molten metal collides against the surface, it rapidly goes through solidification while spreading on the wet surface. High-speed thermo-camera was utilized to measure the molten metal's surface temperature during the spreading transient. Once the molten metal completely solidifies, molten metal's spread area and thickness were measured. From the obtained database, a dimensional analysis was conducted to investigate the key parameters responsible for the molten metal spreading on the wet surface. Based on the key non-dimensional parameters identified in the current analysis, the new empirical correlation was proposed. Its predictive capability was found to be 18.9% in mean absolute relative deviation. [ABSTRACT FROM AUTHOR]

The Great East Japan Earthquake occurred on March 11, 2011 fatally damaged the Fukushima Daiichi Nuclear Power Plant (NPP), caused prolonged station blackout (SBO). Following the SBO, the reactor water level gradually dropped due to the increase in steam discharge from safety relief valves and eventually led to nuclear fuel melt down. Almost four years have passed since the accident, and official reports by the Japanese regulatory have given the general description of causes and progressions of the fatal accident. Even after the Fukushima accident, more than 430 nuclear power plants are currently operating and over 80 units are under construction worldwide. From this fact alone, it is extremely important to learn from the Fukushima accident and enhance the safety culture of the reactor operation to completely eliminate the possibilities of catastrophic accidents seen in 2011. In this study, the best estimate transient thermal-hydraulics code, RETRAN-03/MOD04 was utilized to focus on the effectiveness of the Isolation Condenser (IC) installed on Advanced Boiling Water Reactors (ABWR). The ICs turned out to be one of the very few operable safety systems during the Fukushima accident, and this simple yet reliable safety system should be utilized to secure ABWR from possible reactor core damages. In the present paper, several case studies conducted utilizing the ICs are presented and methods of countermeasures to improve light water reactor safety level in design and operation features are proposed. [ABSTRACT FROM AUTHOR]