Experimental Study of Two-Dimensional Thermal-Hydraulic Behavior in Core during Reflood Phase of PWR LOCA

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles due to the nonuniform water accumulation in the upper plenum.     

Effects of ECC Water Injection Rate on Reflood Phase during LOCA

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.

The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI₄ is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.

Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI₄, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.     

Correlation of Quench Phenomena for Bottom Flooding during Loss-of-Coolant Accidents

The three quench modes based on the observation with an outside-heated quartz tube experiment, i.e. (1) liquid column type, (2) dryout type and (3) droplet-rewetting type are discussed quantitatively to provide correlations for a reflood analysis code.

For the quench velocity of liquid column type and dryout type, using the data of PWR-FLECHT experiments, a correlation is obtained. This correlation is compared with experimental data and found that it can predict the inverse quench velocity within ±20% accuracy under the following conditions: pressure 1.0–4.1 kg/cm²-a, quench temperature 370~510°C, local subcooling 0~30°C, flow rate 5~25 cm/sec.

For the quench temperature of droplet-rewetting type, a correlation is derived and the prediction from it gives the qualitative agreement with few available data.     

Effect of Liquid Flow Rate on Film Boiling Heat Transfer during Reflood in Rod Bundle

An experiment was performed in a wide range of core flooding rate to study the effect of liquid flow rate on the film boiling heat transfer during reflood using a 6 × 6 heater rod bundle with a full-height heated length. The flooding rate was in the range of 2–30cm/s. The obtained data were evaluated through assessment and improvement of Murao-Sugimoto heat transfer coefficient correlation, which predicted well the heat transfer coefficient under the core flooding rate up to 10 cm/s, but underestimated the heat transfer coefficient in the range of the flooding rate higher than 10 cm/s. An empirical correction factor for the Murao-Sugimoto correlation was proposed based on the present experimental data. The proposed correction factor was confirmed to be applicable to another large scale tests.     

Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA

The heat transfer in higher power bundles was enhanced in large-scale reflood tests at Japan Atomic Energy Research Institute. The heat transfer enhancement in the core under a radial power distribution is very important to quantify the safety margin in PWR-LOCA. In this study, we analyzed the physical mechanism by numerical simulations with a multi-dimensional two-fluid model code, REFLA/TRAC, using data from the large scale reflood test. The heat transfer enhancement is caused by the increase of local upward liquid velocity resulting from the formation of flow circulation in the core. The flow circulation is generated by a radial difference of waterhead below quench front under a radial power distribution. The upward liquid velocity depends on the bundle power and the cross flow resistance. The higher power and the smaller cross flow resistance give the higher upward liquid velocity, which increases the magnitude of the heat transfer enhancement. Through the present study, some guidelines were obtained for the multi-dimensional analyses to predict the heat transfer enhancement phenomenon with high accuracy.     

Effects of Partial Flow Blockage on Core Heat Transfer in Forced-Feed Reflood Tests

This study investigates the effects of partial flow blockage due to ballooning of fuel cladding on the core heat transfer during reflood phase in a PWR loss-of-coolant accident, in particular, the effects of coolant bypass flow at flow blockage and the effects of major parameters in a wide core with a bundlewise flow blockage.

Forced-feed reflood tests were carried out with the Slab Core Test Facility, in which 8 simulated fuel bundles are arranged in a row with two out of these designed as blockage bundles with about 60 % blockage ratio. The test results which were obtained under most probable coolant injection conditions were investigated with respect to quench and heat transfer coefficients. As the results, the following were revealed, (1)The effects of flow blockage appear only downstream of flow blockage at the flow blockage bundles with promoted cooling of rods. (2)Effects of coolant bypass flow due to flow blockage are insignificant. (3)Flooding velocity by accumulator injection has a predominant effect on promoted cooling of rods downstream of flow blockage.     

Experimental Assessment of Evaluation Model for Safety Analysis on Reflood Phase of PWR-LOCA

A model named Evaluation Model (EM) is used for the safety evaluation analysis on a PWR-LOCA. The model is conservatively constructed to severely evaluate the core cooling during the LOCA.

In order to assess the model for the reflood phase of the LOCA, a test named “EM Test” was conducted with Cylindrical Core Test Facility which modeled a PWR in 1/21 volumetric scale. The test conditions simulated the reflood phase predicted in the safety evaluation analysis. The test results were compared with the blindfold results predicted by EM codes which are programmed based on the EM. The main conclusions are as follows:

(1) The core heat transfer model built in the EM codes gives conservative results.

(2) The system models in the present EM codes predict the flooding rate conservatively in early transient and realistically in later period.

(3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in a test at lower flow rate of emergency core cooling water did not appear in the EM test.     

LOCA下喷放参数对冷却剂喷放特性影响数值研究

冷却剂喷放过程是失水事故（LOCA）的重要过程之一,研究冷却剂喷放过程的热工水力特性对认识LOCA以及预测事故后放射性源项迁移过程有着重要意义。本文利用FLUNET软件建立冷却剂喷放数值计算模型,并对其进行验证。利用模型研究喷口直径、喷放距离和喷放压力等喷放参数对计算域内流场温度、液滴速度和蒸汽流速等特性的影响。研究结果表明：喷口直径的提高使得喷放参数均有提高;随喷放距离的增大,流场温度和液滴速度先上升后下降,而蒸汽流速先上升后趋于平稳;喷放压力越大,喷放参数的最大值离喷放出口越远,液滴速度和蒸汽流速的最大值随喷放压力的增大逐渐上升,而流场温度最大值没有变化。     

Quantitative Evaluation of Heat Transfer Enhancement due to Radial Power Distribution during Reflood Phase of PWR-LOCA

Experimental studies using Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) indicated that the degree of heat transfer enhancement due to the radial power distribution during the reflood phase of a PWR-LOCA was governed mainly by the radial power ratio itself and less dependent on the shape of radial power distribution within the maximum power ratio of 1.36. The experimental condition covering the wide ranges of the reflood phase and the scale of core radius from 1/4.6 to 1/1 had little effect on the two- dimensional heat transfer behavior. The heat transfer coefficient under nonuniform radial power distribution was expressed as a sum of the heat transfer coefficient obtained under a complete mixing condition and an additional value given by an empirical correlation based on the SCTF results. The temperature rise at the peak power rod calculated with this expression tended to be lower than that calculated with the complete mixing model used in a reflood analysis code REFLA. That is, the complete mixing model was proved to give a conservative result under a nonuniform radial power distribution condition in a full size core.     

Numerical Study on Influence of Blowdown Parameter on Coolant Blowdown Characteristic in LOCA

The coolant blowdown process is one of the important processes of the loss of coolant accident (LOCA). It is of great significance to study the thermal hydraulic characteristics of coolant blowdown process for understanding LOCA and predicting the migration process of radioactive source term after accident. The numerical simulation model of coolant blowdown was established by FLUENT software and verified. The model was used to study the effects of blowdown parameters such as diameter of nozzle, blowdown distance and blowdown pressure on flow field temperature, droplet velocity and vapor velocity. The results show that the increase of diameter of nozzle increases the blowdown parameters. As the blowdown distance increases, the flow field temperature and the droplet velocity increase first and then decrease, while the vapor velocity first rises and then stabilizes. The greater the blowdown pressure is, the farther the blowdown parameter is from the blowdown outlet. The maximum values of droplet velocity and vapor velocity increase gradually with the blowdown pressure, while the maximum value of the flow field temperature does not change.     

Void Fraction in Simulated PWR Fuel Bundle during Reflood Phase

In order to evaluate void fraction in a bundle geometry during the reflood phase, reflooding experiment with a 4×4 simulated fuel array was conducted.

As the result, it was found that the effects of the clad temperature and the power of the heater rods are small and the effects of the pressure and the inlet flow rate are large on the relationship between the superficial steam velocity and the void fraction in a bundle geometry during the reflood phase. It was, also, found that there is no distinct difference of the void fractions caused by the different flow patterns in the wet clad region and in the dry clad region in a bundle geometry during the reflood phase, when compared at the same superficial steam velocities.

Furthermore, the applicability of Cunningham-Yeh's void fraction correlation was investigated under a wide range of conditions anticipated during the reflood phase. The range of conditions under which Cunningham-Yeh's correlation predicts the void fraction within an error band of ±20% were made clear.     

RELAP4 Modeling for Reverse Flow Property of BWR Jet Pump under Large Break LOCA Condition

It is very important to identify the reverse loss coefficient of BWR jet pump in the evaluation of core inlet flow at the beginning phase of BWR LBLOCA (Large Break Loss-of-Coolant Accident) analyses. Hence, the reverse flow property of jet pump was investigated in relation between the momentum equation, pressure loss coefficient and RELAP4 noding, and a new modeling has been proposed. In the proposed modeling, an equivalent pressure loss coefficient is used to take into account of the effect of accellerating pressure loss by the continuous flow area reduction from the tale pipe to the throat. The effectiveness of this model was studied by analyses for the LOFT 1/6 scale jet pump experiment and typical BWR LBLOCA. It has been, consequently, shown that this proposed model gives better jet pump property than a previous model which is used in the WREM sample problem and which gives very conservative result in core inlet flow and in the peak cladding temperature through whole transient.     

Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model

Abstract

In the initial stage of reflood phase of PWR-LOCA, quasi-stable surface of water is established in a core. Spurious pressure spikes are often encountered when two-fluid model codes are used to analyze two-phase flow dynamics in the initial stage of reflood phase. These pressure spikes are not observed in experiments. Since these pressure spikes affect other variables such as void fraction, it is important to eliminate these pressure spikes to get physically reasonable results with two-fluid model codes. In the present study, it is quantitatively clarified that these pressure spikes result from numerical acceleration loss of liquid above the surface of water where liquid does not exist. Furthermore, a method is developed to mitigate the acceleration loss of liquid above the surface of water. It is confirmed that this method is effective to eliminate these pressure spikes without losing benefits of the present two-fluid model codes.     

Application of REFLA Core Model to Safety Evaluation Code

In the analysis of the core thermal-hydraulic behavior during the reflood phase of a PWR-LOCA, current safety evaluation codes like WREM code system are usually limited to use in narrow region where the employed empirical correlations are validated. In order to make a safety evaluation code more flexibly applicable, the empirical correlations in TOODEE2 code in the WREM code system was replaced with the core model built in the REFLA code. By changing the multiplication factor for the calculated heat transfer coefficient for the region above the quench front, the predicted clad surface temperatures were compared with those measured in Cylindrical Core Test Facility (CCTF) tests.

It was found that the multiplication factor 0.9 gives always a conservative prediction against CCTF data.     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Multi-Channel Thermal Hydraulic Model for LOCA Analysis of Heterogeneous BWR Core

A multi-channel thermal hydraulic model for LOCA analysis of a heterogeneous core such as a HCBWR has been developed. This model solves integral formulations for basic equations based on a one-dimensional drift flux model. The core region is divided into several fuel channel groups which differ in their thermal power or geometry. The various flow patterns in the core are determined by calculating the redistribution of vapor generated in the lower plenum into the fuel channel groups. In order to verify the multi-channel model, a computer program FLORA was developed based on the multi-channel model and large and small break LOCA experiments conducted in the Two Bundle Loop (TBL) facility were analyzed by the FLORA program. As a result, the difference in thermal hydraulic behavior between two bundles with different power in the various break LOCA experiments were well simulated.