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The Kori-1 nuclear power plant (NPP), a Westinghouse 2-loop pressurized water reactor in Korea, permanently ceased operations on June 18, 2017. Before the main decommissioning activities, full chemical decontamination of primary systems will be performed to secure the safety of workers and minimize the quantity of radioactive waste. In this study, the steady and transient flow characteristics for the chemical decontamination operations were analyzed via a simulation using the Multidimensional Analysis of Reactor Safety, KINS Standard version. Then, based on the accumulated flow mass released from the reactant cooling systems during transient events, the radioactivity levels were estimated using the oxide layer removal ratio obtained from the chemical decontamination experiment. The loss of residual heat removal, steam generator tube rupture (SGTR), power-operated relief valve open (PORV Open), and combination of SGTR and PORV Open simulations were investigated in terms of the transient events. Finally, a strategy for managing transient events during the full-system chemical decontamination operation of the Kori-1 NPP is proposed. [ABSTRACT FROM AUTHOR]

Nuclear power is a major source of electricity in the international community. However, a significant problem with nuclear power is that, if a severe nuclear accident occurs, radiation may leak and cause great damage. As such, research on nuclear safety has become increasingly popular worldwide. In this paper, the structural integrity of a reactor cavity during a steam explosion—one kind of the aforementioned severe nuclear accidents—was evaluated. Steam explosions are primarily caused by fuel–coolant interactions (FCI), and result from issues in the cooling system that discharges the melt from the reactor core to the outside. A steam explosion can damage the nuclear power plant, and radiation leakage, the greatest concern, may occur. In the Chernobyl or Fukushima Daiichi accidents, significant radiation leakages resulted in damages extending beyond the country of origin. In this paper, a steam explosion was simulated using values given by the transient analysis code for explosive reactions (TRACER-II)—the only steam explosion code in Korea. The walls of the reactor cavity were modeled after the APR-1400 currently operating in Korea. The integrity of the concrete, rebars, and liner plate in the reactor cavity during a steam explosion was evaluated in terms of stress and ductile failure strain limits. [ABSTRACT FROM AUTHOR]

• Transient analysis for block-type VHTR core was developed and tested. • Time-dependent neutron diffusion analysis using FEM was implemented in CAPP code. • A new method (iPRN) was introduced for reducing the control rod cusping effect. • A neutronics solver and a simple thermal-fluid analysis solver were coupled in CAPP. • CAPP gave appropriate results for control rod ejection and withdrawal accidents. The very high-temperature gas-cooled reactor (VHTR) has attracted interest owing to its high passive safety. A reactor analysis tool is required to analyze the safety characteristics of a VHTR in detail, and such a tool should be capable of a transient calculation for a reactivity insertion accident such as control rod ejection or withdrawal. This work presents a transient analysis of a block-type VHTR core. Time-dependent neutron diffusion equation is solved by the finite element method. A simplified thermal fluid analysis tool is also implemented to consider thermal feedback. In addition, a new method is introduced to resolve the control rod cusping effect without additional computation or mesh reconstruction. The above methods are applied to a reactor physics code CAPP, developed at Korea Atomic Energy Research Institute (KAERI). The methods are tested on three-dimensional VHTR problems that include control rods and show encouraging results. [ABSTRACT FROM AUTHOR]