Han-Ok Kang

RESEARCH ACTIVITIES ON A SUPERCRITICAL PRESSURE WATER REACTOR IN KOREA

This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical , an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.

A SIMPLE ANALYTICAL METHOD FOR NONLINEAR DENSITY WAVE TWO-PHASE INSTABILITY IN A SODIUM-HEATED AND HELICALLY COILED STEAM GENERATOR

A simple model to analyze non-linear density-wave instability in a sodium-cooled helically coiled steam generator is developed. The model is formulated with three regions with moving boundaries. The homogeneous equilibrium flow model is used for the two-phase region and the shell-side energy conservation is also considered for the heat flux variation in each region. The proposed model is applied to the analysis of two-phase instability in a JAEA (Japan Atomic Energy Agency) 50MWt No.2 steam generator. The steady state results show that the proposed model accurately predicts the six cases of operating temperatures on the primary and secondary sides. The sizes of three regions, the secondary side pressure drop according to the flow rate, and the temperature variation in the vertical direction are also predicted well. The temporal variations of the inlet flow rate according to the throttling coefficient, the boiling and superheating boundaries and the pressure drop in the two-phase and superheating regions are obtained from the unsteady analysis.

Structural Integrity Confirmation of a Once-through Steam Generator from the Viewpoint of Flow Instability

Helically-coiled once-through steam generators have been utilized for an integral type reactor showing several benefits such as high quality steam generation, geometric compactness, and compensation for a thermal expansion. Steam generator operations with unstable two-phase flow conditions on the tube-side may cause degradation of the tube materials and curtail the lifetime of the component. Based on existing experimental results for a once-through steam generator, its structural integrity was confirmed from the viewpoint of flow instability. The work was composed of three items, the prevention of static instability between the module steam/feedwater pipes, tube inlet orifice sizing against a dynamic instability between the heated coils, and a thermal-cyclic stress analysis for an overall component lifetime evaluation. The static thermo-hydraulic calculation for the steam generator cassette showed that while the prevention of the static instability was satisfied for the power operational mode, special care should be taken during the startup/cooling operational modes. The tube inlet orifice size was determined based on the orifice coefficient concept and existing experimental data for once-through steam generators. The thermal-cyclic stress evaluation for the heated tube revealed that the maximum alternating stress intensity was lower than the allowable fatigue limit value of the tube material.

DEVELOPMENT OF REACTOR POWER CONTROL LOGIC FOR THE POWER MANEUVERING OF KALIMER-600

We developed an achievable control logic for the reactor power level during a power maneuvering event and set up some constraints for the control of the reactor power in a conceptual sodium-cooled fast reactor (KALIMER-600) that was developed at KAERI. For simulating the dynamic behaviors of the plant, we developed a fast-running performance analysis code. Through various simulations of the power maneuvering event, we evaluated some suggested control logic for the reactor power and found an achievable control logic. The objective of the control logic is to search for the position of the control rods that would keep the average temperature of the primary pool constant and, concurrently, minimize the power deviation between the reactor and the BOP cycle during the power maneuvering. In addition, the flow rates of the primary pool and the intermediate loop should be changed according to the power level in order to not violate the constraints set up in this study. Also, we evaluated some movement speeds of the control rods and found that a fast movement of the control rods might cause the power to fluctuate during the power maneuvering event. We suggested a reasonable movement speed of the control rods for the developed control logic.

Development of a thermal-hydraulic analysis code for once-through steam generators using straight tubes for SMRs

A thermal-hydraulic design and performance analysis computer code for a once-through steam generator using straight tubes is developed. To benchmark the developed physical models and computer code, an once-through steam generator developed by other designer is simulated and the calculated results are compared with the design data. Also, the same steam generator is analyzed with the best-estimate thermal-hydraulic system code, MARS, for the code-to-code validation. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by the developed code show general agreements with the published design data as well as the analysis results of MARS. It is demonstrated that the developed code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

LINEAR INSTABILITY ANALYSIS OF A WATER SHEET TRAILING FROM A WET SPACER GRID IN A ROD BUNDLE

The reflood test data from the rod bundle heat transfer (RBHT) test facility showed that the grids in the upper portion of the rod bundle could become wet well before the arrival of the quench front and that the sizes of liquid droplets downstream of a wet grid could not be predicted by the droplet breakup models for a dry grid. To investigate the water droplet generation from a wet grid spacer, a viscous linear temporal instability model of the water sheet issuing from the trailing edge of the grid with the surrounding steam up-flow is developed in this study. The Orr-Sommerfeld equations along with appropriate boundary conditions for the flow are solved using Chebyshev series expansions and the Tau-Galerkin projection method. The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.

Simple Nonlinear Methods for Predicting Two-Phase Instabilities in a Helically Coiled Steam Generator

A simple model to analyze the non-linear density-wave instability in a sodium cooled, helically coiled steam generator is developed. The model is formulated with three regions with moving boundaries. The homogeneous equilibrium flow model is used for the two-phase region and the shell-side energy conservation is also considered for the heat flux variation in each region. The proposed model is applied to the analysis of two-phase instability in a JAEA (Japan Atomic Energy Agency) 50MWt No.2 steam generator. The steady state results show that the proposed model accurately predicts the six cases of the operating temperatures in the primary and secondary sides. The sizes of the three regions and the secondary side pressure drop according to the flow rate, and the temperature variation in the vertical direction are also predicted well. The temporal variations of the inlet flow rate according to the throttling coefficient, the boiling and superheating boundaries and the pressure drop in the two-phase and superheating regions are obtained from the unsteady analysis.Copyright

Aperiodic Instability of a Once-Through Steam Generator with a Feedwater Line

Aperiodic (static) flow instability is an instability related to the change of a flow direction in individual steam generating U-shaped channels operating at given pressure difference. The nature of an aperiodic instability is close to a Ledinegg instability [1] related to the presence of multiple flows at the full hydraulic curve of a U-shaped channel. In this paper, the conditions for a reverse flow for a once-through steam generator (OTSG) with U-shaped modular feedwater line (MFL) are studied. From the results of the studies, it is revealed that the change of a flow direction in the MFL is due to the boiling of the feedwater in the downcomer branch of the U-shaped MFL and that multiple flows start in an area of the extremes corresponding to the minimum pressure difference of the hydraulic curves. Calculation models for predicting a threshold of an aperiodic instability for the OTSG of interest is proposed and the analysis results are compared with the experimental data.Copyright