低过冷流场中蒸汽水下喷注直接接触冷凝声学特性研究

低过冷流场中蒸汽水下喷注直接接触冷凝（DCC）是核动力舰船冷凝器热井鼓泡除氧的热力基础,该过程涉及强湍动、相变、多相流等复杂行为,可诱发强烈的流场压力脉动与辐射噪声,是影响舰船声隐身性能的重要因素。针对该问题,本文通过实验对流场过冷度2.8~14.7 ℃范围内蒸汽水下喷注DCC过程的汽羽流型演化和声学特性进行研究。结果表明,蒸汽质量流速和流场过冷度对汽羽流型演化具有不同的作用机制和影响规律。低过冷流场中,蒸汽水下喷注冷凝噪声具有宽频特征;流场过冷度极低时,低频噪声占辐射噪声的主导地位,随流场过冷度的增大,高频噪声的贡献逐渐提升。     

Application of Simplified Condensation Model to PWR LBLOCA Transient Analysis with TRAC-PF1 Code

In the system analyses of a large-break loss-of-coolant accident (LBLOCA) of pressurized water reactors (PWRs), the TRAC-PF1 code predicted an unrealistic depressurization and required much computational time due to the problem of the condensation model. To eliminate the unrealistic depressurization, the TRAC-PF1 code was improved using a simplified condensation model that determined the total condensation rate at cold leg. Through the assessment calculations for CCTF, UPTF and LOFT tests, it was confirmed that the simplified model could eliminate the unrealistic depressurization and reduce the computational time. It was also confirmed that the model could improve the accuracy of the system calculation for the core inlet flow rate and clad temperature as the result of the elimination of the unrealistic depressurization. It has been concluded that the simplified condensation model is useful for the system calculation of the PWR LBLOCA.     

Experimental Study of Counter-Current Two-Phase Flow in Horizontal Tube Connected to Inclined Riser

In order to determine the Counter-Current Flow Limitation (CCFL) in hot legs of PWRS, CCFL characteristics of air-water and saturated steam-water flow were experimentally investigated in a modeled flow path of a horizontal tube connected to an inclined riser. The ranges of dimensions of experimental tubes were as follows: diameter D 0.026–0.076 m, length of horizontal tube H 0.01–0.4 m, length of inclined riser I 0.038–0.6 m and inclination angle of inclined riser θ 40° or 45°.

Wallis-type correlation (J _g ^*1/2+mJ _i ^*1/2=C) was applicable to the data during a steady separated flow. An analysis based on envelope theory showed that the constant C should be a function of H/D and I. A function of C with those parameters was empirically determined by using data obtained in this study. The developed function correlated well with the results of Richter et al. (D = 0.203 m, H = 1.26 m, I = 0.5 m and θ = 45°). The constant m in the Wallis-type correlation was almost constant 0.75. The problems were discussed, which should be made clear to apply the correlation obtained in this study to an actual PWR hot leg.     

Heat Transfer Coefficients of Steam Condensation on Containment Vessel Wall during Blowdown

Heat transfer coefficients of steam condensation on the containment vessel wall at a LOCA are studied. As to the steady state heat transfer coefficients, though Sagawa's data are a little smaller than Uchida's data, they are very close to the analytical solutions by Mori-Hijikata. Transient heat transfer coefficients are represented by the steady state heat transfer coefficients multiplied by a factor. The factor expresses the agitation effect weakening with time during blowdown. Values of parameters in the factor are determined so as the heat transfer coefficients to fit Sagawa's data. These heat transfer coefficients are applied to the analyses of the experiment with the simulation apparatus of an integrated type marine water reactor. Values of the parameters are also determined so that the temperature transients on the containment vessel wall by analyses fit them of experiment. The differences of the values of parameters by the analyses and by Sagawa's data are discussed.     

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.     

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.     

Numerical Analysis of Two-Phase Jet Impingement by Homogeneous Flow Model

A two-dimensional homogeneous equilibrium model is developed in order to evaluate impingement load caused by discharge of a two-phase mixture in postulated pipe rupture accidents of light water reactors. The present analysis differs from previous studies mainly in that a backward expansion around the pipe exit is taken into account. As a result: (1) it is confirmed that the backward expansion occurs around the pipe exit in a super-sonic two-phase flow; (2) when the dimensionless position of an impingement wall z/D is larger than 2.0, the present calculations predict the pressure distribution on the impingement wall within an error of 10%, while the previous calculations, which did not take the backward expansion into account, overestimated the pressure by 25%; (3) existence of jet core and occurrence of shock waves in the two-phase jet are obtained and (4) a supersonic state of the jet is illustrated by comparing the velocity with the sonic velocity.     

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.     

Two-Phase Flow Phenomena in Broken Recirculation Line of BWR

When a primary recirculation line of BWR is ruptured, a primary recirculation pump may be subjected to very high velocity two-phase flow and its speed may be accelerated by this flow. It is important for safety evaluation to estimate the pump behavior during blowdown. There are two problems involved in analyzing this behavior. One problem concerns the pump characteristics under two-phase flow. The other involves the two-phase conditions at the pump inlet. If the rupture occurs at a suction side of the pump, choking is considered to occur at a broken jet pump nozzle. Then, a void fraction becomes larger downstream from the jet pump nozzle and volumetric flow through the pump will be very high. However, there is little experimental data available on two-phase flow downstream from a choking plane. Blowdown tests were performed using a simulated broken recirculation line and measured data were analyzed by TRAC-PIA. Analytical results agreed with measured data.     

Relaxation Process of Thermal Non-Equilibrium in Boiling Two-Phase Flow, (I)

The thermo-hydraulics in a thermal non-equilibrium state, which may appear during a postulated loss-of-coolant accident, is one of phenomena which have never been understood enough. In this paper, accelerating boiling two-phase flow is dealt with which realizes thermal non- equilibrium conditions in a steady state in order to clarify the relaxation process of superheated liquid. The following insight is clarified from the results; Jones' model cannot explain the underpressure and relaxation time unless an empirical constant is changed. Therefore, the effect of interfacial area density on the macroscopic relaxation inception is experimentally investigated by adding artificially bubbles in accelerating boiling two-phase flow in order to offer a fundamental data base for modeling the relaxation process of superheated liquid. It is made clear that the macroscopic relaxation inception can be evaluated by a relationship between the integration of liquid superheating by the travelling time and the initial interfacial area density.     

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.     

Restrictive Effect of Ascending Steam on Falling Water during Top Spray Emergency Core Cooling

Water spraying experiments were conducted to find out a flow rate of falling water overcoming ascending steam during top spray emergency cooling with an 8×8 type simulated fuel rod bundle of real size. The bundle consisted of 64 rods, each with a diameter of 12.5 mm, arranged in the form of square lattice with a pitch of 16.3 mm. In the experiments the simulated fuel rods were not heated. Instead, steam was injected into the lower plenum vessel simulating bundle-generated steam. As the results, (1) a criterion was proposed to determine the region where the restrictive effect of ascending steam on falling water appears, considering the decrease of a flow rate of ascending steam due to condensation by a spray of subcooled water, (2) the restrictive effect was independent of water head on the upper tie plate and water injection methods, and (3) an analytical model based on the pressure balance at the upper tie plate was proposed to calculate a flow rate of falling water overcoming ascending steam.     

Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility

An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top.     

Development of Interfacial Friction Model for Two-Fluid Model Code against Countercurrent Gas-Liquid Flow Limitation in PWR Hot Leg

An interfacial friction model for two-fluid model code has been developed for the counter- current gas-liquid flow limitation at hot leg in a pressurized water reactor. Firstly, using a typical two-fluid model code TRAC-PF1/MOD1, we analyzed whether the interfacial friction model under countercurrent stratified flow by Ohnuki et al., which has been verified with an envelope model at steady state, functions well for the dynamic calculation with the two-fluid model code or not. It was found from the analyses that the model by Ohnuki et al. should be combined with a suitable interfacial friction model for the slug flow regime in hot leg. Based on flow observation in a small scale air-water experiment, models at the bend of hot leg and in the roll wave regime in the horizontal flow path of hot leg were newly developed as the model in the slug flow regime and the slug flow model was combined with the model by Ohnuki et al., The validity of the present model was confirmed with the data under various conditions of scales, pressures and fluid combinations (inner diameter: 0.025~0.75m, pressure: 0.1~7.1 MPa and air-water or steam-water).