Young-Jong Chung

A Conservative Analysis Methodology for a Steam-Line-Break Accident of the SMART-P Plant

Abstract The system-integrated modular advanced reactor (SMART) new phase (SMART-P) with a rated thermal power of 65.5 MW is currently being developed at the Korea Atomic Energy Research Institute. It is an innovative design to achieve a high degree of safety by adopting inherent safety-improving features and passive safety systems. Realistic and conservative calculations and a parameter study for a steam-line pipe break have been carried out by means of the TASS/SMR code. A set of transients for the whole system of SMART-P is investigated from the point of view of fuel integrity. The results of the analyses show that the most conservative initial conditions are thermal design flow, high system pressure, high coolant temperature, and high core power. It is also assumed that off-site power is unavailable and the steam section pipe guillotine break with the least reactive control rod assembly stuck out in the fully withdrawn position is a limiting case under the most moderator density reactivity condition. The SMART-P safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters such as the critical heat flux ratio and the pressure. Natural circulation is well established in the primary and passive residual heat removal systems and is enough to ensure a stable plant shutdown condition after a reactor trips.

Experimental Study of Thermal Mixing in Flow Mixing Header Assembly of SMART

Abstract SMART (System-integrated Modular Advanced ReacTor) is an integral-type nuclear reactor for cogeneration that adopts a flow mixing header assembly (FMHA) to maintain a uniform temperature distribution in the coolant at the core inlet in the case of failure in the steam generator or reactor coolant pump. The SMART FMHA is important for enhancing thermal mixing of the coolant during a transient and even during accidents, so it is essential that the thermal-hydraulic characteristics of flow in the FMHA be understood. Scaling analysis was performed to design the experimental facility for the FMHA test through computational fluid dynamics (CFD) analysis on the SMART prototype and experimental model. The experimental facility was designed by a linear scaling factor 0.18, and the experimental pressure and temperature conditions were 0.1 MPa and 30°C to 60°C, respectively. The experiment was performed in two ways: using FMHAs with large outlet flow hole sizes and FMHAs with small outlet flow hole sizes. In the cases of failure of one or two steam generators, the maximum temperature deviation on the side of the reactor core inlet was measured to be 1°C to 2°C, which demonstrates excellent thermal mixing through the FMHA. In particular, the FMHA with small outlet flow hole sizes tended to have better thermal mixing than the FMHA with large outlet flow hole sizes. The experimental results were comparable to those from CFD analysis.

Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

System-integrated modular advanced reactor (SMART) is a small-sized advanced integral type pressurized water reactor (PWR) with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

Experimental Validation for Specific Thermal Hydraulic Phenomena of SMART

SMART (System-integrated Modular Advanced ReacTor) is an integral light water reactor developed by KAERI for both electricity generation and various applications. SMART can produce 90MWe of electricity and 40,000 tons/day of sea-water desalination, which is deemed suitable for a city population of 100,000. The SMART project aims at standard design approval. The experimental validation for SMART specific thermal hydraulic phenomena is to generate background information required for SMART licensing.The objective of the experimental validation is to demonstrate the safety and performance of the SMART-specific system and component designs, and to provide a database for the validation of the design tools, such as a system analysis code. From the experimental results, it is proved that SMART was designed properly and the data are used to approve the validation of the system analysis code. Also, the experimental results were submitted to a regulatory body as technical background information of the SMART standard design, and they enhance the licensing possibility of this design.Copyright

Best Estimated and Conservative Analyses for a Feedwater Pipe Break Accident of an Integral Type Reactor

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters.