Sang-Ki Moon

KAERI Integral Effect Test Program and the ATLAS Design

Abstract The thermal-hydraulic integral effect test (IET) program is being progressed by the Korea Atomic Energy Research Institute. This paper presents an overview of the IET program; the scientific design characteristics of the IET facility; ATLAS, which is under construction; and the experimental and analytical validation works. The ATLAS facility has the following characteristics: (a) a 1/2-height, 1/288-volume, full-pressure simulation of the APR1400, (b) geometrical similarity with the APR1400, including 2 (hot legs) × 4 (cold legs) reactor coolant loops, a direct vessel injection (DVI), an integrated annular downcomer, etc., (c) incorporation of the specific design characteristics of the 1000-MW(electric) class Korean Standard Nuclear Power Plant, such as a cold-leg injection and the low-pressure injection pumps, (d) a maximum 8% of the scaled nominal core power, and (e) simulation capability of broad scenarios, including the reflood phase of the large-break loss-of-coolant accidents (LOCAs), small-break LOCA scenarios including the DVI line breaks, steam generator tube ruptures, main steam line breaks, midloop operation, etc. The scientific design of the ATLAS was accomplished rigorously from the viewpoints of both a global and local scaling based on the three-level scaling methodology of Ishii et al. The validation works showed that the scientific design of the ATLAS test facility is sound.

Effect of pressure on critical heat flux in uniformly heated vertical annulus under low flow conditions

Critical heat flux (CHF) experiments have been carried out in a wide range of pressure for an internally heated vertical annulus. The experimental conditions covered a range of pressure from 0.57 to 15.01 MPa, mass fluxes of 0 kg m−2 s−1 and from 200 to 650 kg m−2 s−1, and inlet subcoolings from 85 to 413 kJ kg−1. Most of the CHFs were identified to the dryout of the liquid film in the annular-mist flow. For the mass fluxes of 550 and 650 kg m−2 s−1, the CHFs had a maximum value at a pressure of 2–3 MPa, and the pressure at the maximum CHF values had a trend moving toward the pressure at the peak value of pool boiling CHF as the mass flux decreased. The CHF data under a zero mass flux condition indicate that both the effects of pressure and inlet subcooling on the CHF were smaller, compared with those for the CHF with a net water upflow. The Doerffer correlation using the 1995 CHF look-up table and the Bowring correlation show a good prediction capability for the present CHF data.

Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.

Critical heat flux under zero flow conditions in vertical annulus with uniformly and non-uniformly heated sections

Abstract The experimental study of water CHF (critical heat flux) under zero flow conditions has been carried out in an annulus flow channel with uniformly and non-uniformly heated sections over a pressure range of 0.52–14.96 MPa. In the present boiling system, the CHFs occur in the upper region of the heated section, in contrast to the results in the experiments for boiling tubes conducted by several investigators. The general trend of the CHF with pressure is that the CHF increases up to a medium pressure of about 6–8 MPa and decreases as the pressure is further increased. A comparison of the present data with the existing flooding CHF correlations shows that the correlations depend greatly on the effect of the heat flux distribution. When the correction terms with the density ratio and the effect of the heat flux distribution proposed in the present work are used with the CHF correlation based on the Wallis flooding correlation, it predicts the measured flooding CHF within an RMS error of 9.0%.

Spacer Grid Effects during a Reflood in an Annulus Flow Channel

An experimental study has been performed to investigate the effects of a spacer grid in an annulus flow channel with a uniform power shaped single rod during a bottom-reflood phase. The ranges of the experimental parameters are 2-8 cm/s for the flooding velocity, 20-80°C for the inlet subcooling temperature, and 500-700°C for the initial wall temperature. Two types of spacer grids, i.e., a swirl-vane type grid and a straight egg-crate type spacer grid, have been tested to compare the differences in their thermal hydraulic behavior through the spacer grids. Flow patterns and a rewetting front behavior near a spacer grid are remarkably altered with the types of spacer grids used. In the case of a low flooding rate and a high wall temperature condition, the cooling capacity of the swirl-vane spacer grid is better than that of the straight egg-crate type grid. Rewetting velocities through the swirl-vane spacer grids are faster than those through the other types of grids. The cladding temperature of the heater rod near a spacer grid shows a different pattern with the types of spacer grid used.

Effect of Flow Blockage on the Coolability during Reflood in a 2 2 Rod Bundle

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

An Experimental Study of Critical Heat Flux in Non-uniformly Heated Vertical Annulus under Low Flow Conditions

An experimental study on critical heat flux (CHF) has been performed in an internally heated vertical annulus with non-uniform heating. The CHF data for the chopped cosine heat flux have been compared with those for uniform heat flux obtained from the previous study of the authors, in order to investigate the effect of axial heat flux distribution on CHF. The local CHF with the parameters such as mass flux and critical quality shows an irregular behavior. However, thetotal critical power with mass flux and theaverage CHF with critical quality are represented by a unique curve without the irregularity. The effect of the heat flux distribution on CHF is large at low pressure conditions but becomes rapidly smaller as the pressure increases. The relationship between the critical quality and the boiling length is represented by a single curve, independent of the axial heat flux distribution. For non-uniform axial heat flux distribution, the prediction results from Doerffer et al.’s and Bowring’s CHF correlations have considerably large errors, compared to the prediction for uniform heat flux distribution. KeyWords : Critical Heat Flux, Heated Vertical Annulus, Low Mass Flux, Wide Range Pressure, Non-uniform Heating, Effect of Axial Heat Flux Distribution, Boiling Length

Effect of the Axial Heat Flux on the Critical Heat Flux in Low Flow Conditions with Vertical Annuli

An experimental study on the critical heat flux (CHF) has been performed for a water flow in vertical annuli with uniform and chopped cosine axial heat flux distributions under low flow and a wide range of pressure conditions. The effect of an axial heat flux distribution on the critical power is large at low-pressure conditions, but the effect decreases rapidly as the pressure increases. The ‘overall power’ hypothesis becomes more accurate as the pressure and critical quality increase, and the ‘local conditions’ hypothesis is not valid at high quality conditions. For a fixed mass flux and pressure, the relationship between the critical quality and the boiling length is represented by a single curve for the whole range of the present experimental conditions, regardless of the axial heat flux distributions. Smolin et al.s F-factor and Bowrings Y parameter show good prediction results for the present CHF data.

Critical Heat Flux Tests for an Application of the Three-Pin Fuel Test Loop in HANARO

Critical heat flux (CHF) tests in a three-rod bundle were carried out, and a CHF database was established in order to obtain an operational license for the three-pin fuel test loop (FTL) in HANARO (High-flux Advanced Neutron Application Reactor). Two kinds of CHF tests are currently being performed for the CANDU and PWR-type fuel test loop. In this study, an experimental work on the CHF tests for the CANDU-type fuel test loop is explained, and a new CHF prediction methodology is proposed. In all, 108 experimental data points have been obtained, and the data are analyzed and compared with the available CHF correlations for a bundle or an annulus geometry. The 1986 AECL look-up table with a bundle correction factor and the Doerffers correlation for annuli are compared with the present CHF data. It is found that the 1986 AECL look-up table with a bundle correction factor results in a better prediction than the Doerffers correlation for annuli geometry. A three-pin correction factor is developed to account for the geometric effects of the three-rod bundle and to improve the prediction accuracy. It is concluded that the best estimate thermal hydraulic system code, MARS 3.0, which uses the same look-up table for a CHF prediction, can be used for a safety analysis of the CANDU-type three-pin FTL to obtain a license if it is corrected by the developed three-pin correction factor.