Byong-Jo Yun

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.

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3 × 3 Rod Bundles

In this paper, the interfacial flow structure of subcooled water boiling flow in a subchannel of 3 × 3 rod bundles is presented. The 9 rods are positioned in a quadrangular assembly with a rod diameter of 8.2mm and a pitch distance of 16.6 mm. Local void fraction, interfacial area concentration, interfacial velocity, Sauter mean diameter, and liquid velocity have been measured using a conductivity probe and a Pitot tube in 20 locations inside one of the subchannels. A total of 53 flow conditions have been considered in the experimental dataset at atmospheric pressure conditions with a mass flow rate, heat flux, inlet temperature, and subcooled temperature ranges of 250–522 kg/m s, 25–185 kW/m2, 96.6–104.9°C, and 2–11 K, respectively. The dataset has been used to analyze the effect of the heat flux and mass flow rate on the local flow parameters. In addition, the area-averaged data integrated over the whole subchannel have been used to validate some of the distribution parameter and drift velocity constitutive equations and interfacial area concentration correlations most used in the literature.

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

Abstract Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focused on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.

Development of the five-sensor conductivity probe method for the measurement of the interfacial area concentration

The interfacial area concentration (IAC) is one of the most important parameters in the two-fluid model for two-phase flow analysis. The purpose of this study is to develop the local IAC measuring method by using the five-sensor conductivity probe. In this paper, the mathematical approach of the five-sensor conductivity probe method for measuring the local time averaged IAC is described, and numerical simulations are carried out for cap bubbles and two types of ellipsoidal bubbles, as well as spherical bubbles, in order to evaluate the new method. The sizes and locations of the bubbles are determined by using random number generators. To investigate the probe size effect on the accuracy of IAC measurement, three cases for length scales of the probe are applied to the simulations. The simulations show that the five-sensor conductivity probe method proposed in this paper produces better results than the four-sensor method.

Investigation of the Downcomer Boiling Phenomena During the Reflood Phase of a Postulated Large-Break LOCA in the APR1400

Hydraulic phenomena in the downcomer of a conventional pressurized water reactor have an important effect on the transient evaluations of a postulated large-break loss-of-coolant accident (LBLOCA). In particular, safety analyses using best-estimate codes show that downcomer boiling is one of the important phenomena in the postulated LBLOCA because it can degrade the hydraulic head in the downcomer and consequently affect the reflood flow rate for core cooling. To experimentally identify the thermal-hydraulic behavior in the downcomer, a downcomer-boiling test facility was constructed for simulating downcomer boiling in the reflood phase of a postulated LBLOCA. The test facility was designed by adopting a full-pressure, full-height, and full-size downcomer-gap approach but with the circumferential length reduced 47.08-fold. The test was divided into two phases: (a) visual observation and acquisition of the global two-phase flow parameters and (b) measurement of the local two-phase flow parameters. This paper presents the test results from Phase I. The major measured parameters were the axial void fraction and the fluid temperatures and pressures in the test section. The measured data were used to evaluate a safety analysis code, MARS 2.1b, to investigate its modeling accuracy and identify weaknesses of the thermal-hydraulic models therein.

ASSESSMENT OF CONDENSATION HEAT TRANSFER MODEL TO EVALUATE PERFORMANCE OF THE PASSIVE AUXILIARY FEEDWATER SYSTEM

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

Effect of the yaw injection angle on the ECC bypass in comparison with the horizontal DVI

Abstract The comparison tests for the direct emergency core cooling (ECC) bypass fraction were experimentally performed with a typical direct vessel injection (DVI) nozzle and an ECC column nozzle having a yaw injection angle to the gravity axis. The ECC yaw injection nozzle is newly introduced to make an ECC water column in the downcomer region. The yaw injection angle of the ECC water relative to the gravity axis is varied from 0 to (±)90° stepped by 45°. The tests are performed in the air–water separate effect test facility (direct injection visualization and analysis (DIVA)), which is a 1/7.07 linearly scaled-down model of the APR1400 nuclear reactor. The test results show that (1) if the ECC water column is injected into the wake region which is induced by the hot leg blunt body in the downcomer annulus, the ECC bypass fraction is greatly reduced compared with the typical horizontal ECC injection which makes ECC film on the downcomer wall. At the same time, the ECC penetration toward the lower downcomer region becomes larger than those of a typical horizontal type of direct vessel injection on the downcomer wall vertically. (2) If the ECC water column is injected near the broken cold leg, the ECC water is directly bypassed. Thus, the ECC penetration fraction is greatly reduced compared with a typical film type of the horizontal ECC injection. (3) In order to minimize the ECC bypass fraction, the ECC water should be injected toward the wake region of the hot leg blunt bodies.

Experimental Investigation into the Effect of the Passive Condensation Cooling Tank Water Level in the Thermal Performance of the Passive Auxiliary Feedwater System

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+) and is designed to completely replace a conventional, active auxiliary feedwater system. With the aim of validating the cooling and operational performance of the PAFS, a separate effect test facility, the PAFS Condensing heat removal Assessment Loop (PASCAL), was constructed by simulating a single passive condensation heat exchanger (PCHX) tube submerged in the passive condensation cooling tank (PCCT) according to the volumetric scaling methodology. During heat removal of the PAFS, the pool water in the PCCT plays a role in the ultimate heat sink of a decay heat. In this study, the effect of the PCCT water level on the cooling performance of the PAFS was experimentally investigated with the PASCAL facility. Quasi-steady-state and PCCT level decrease test cases were sequentially performed by varying the steam generator heater power from 300 to 750 kW to investigate the thermal-hydraulic behavior during the decrease of the PCCT water level. From the experimental results, it was found that the decrease of the PCCT water level enhanced evaporative heat transfer at the outer wall of the PCHX tube by reducing the degree of subcooling around the PCHX. That induced an increase of the heat removal rate by the PCHX during the transient. Thus, it can be concluded that the current design of the PCHX in the PAFS has sufficient capacity to cool down the decay heat during the whole transient of the PCCT water level decrease.

Local Flow Structure of Subcooled Boiling Flow of Water in a Heated Annulus

Local measurements of flow parameters were performed for vertical upward subcooled boiling flows in an internally heated annulus. The annulus channel consisted of an inner heater rod with a diameter of 19.0 mm and an outer round tube with an inner diameter of 37.5 mm, and the hydraulic equivalent diameter was 18.5 mm. The double-sensor conductivity probe method was used for measuring the local void fraction, interfacial area concentration, bubble Sauter mean diameter and gas velocity, whereas the miniature Pitot tube was used for measuring the local liquid velocity. A total of 32 data sets were acquired consisting of various combinations of heat flux, 88.1–350.9 kW/m2 , mass flux, 469.7–1061.4 kg/(m2 s) and inlet liquid temperature, 83.8–100.5 °C. Six existing drift-flux models and six existing correlations of the interfacial area concentration were evaluated by the data obtained in the experiment.Copyright

Characteristics of Downcomer Boiling Phenomena During the Reflood Phase of a Postulated Large Break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor have an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling is being progressed in the reflood phase of a postulated LBLOCA. For this, the test facility was designed as a one side heated rectangular test section which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II) measurement of the local two-phase flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and parts of Phase-II were introduced.Copyright