Passive Condensation Cooling Tank Research Articles

Thermal-hydraulic study of air-cooled passive decay heat removal system for APR+ under extended station blackout

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

A CFD-based design optimization of air-cooled passive decay heat removal system

Abstract The concept of the APDHR (Air-cooled Passive Decay Heat Removal) system was suggested to preserve the safety of a nuclear reactor during accidents. Until 3 days after the reactor shutdown caused by non-LOCA accident, water in the Passive Condensate Cooling Tank (PCCT) was used to condense the steam in secondary side. Since then, the steam was cooled by natural convection of air passively and indefinitely. The focus on this study is on the system performance during the natural convection period. Both, finned and bare heat exchangers (HXs) were considered for the design optimization of the APDHR. As a result, fin height of 4 cm, fin spacing of 4 cm and fin thickness of 0.2 cm was determined as a reference fin geometry regarding heat removal capacity and economic fin installation. Then, the sensitivity of several design parameters of APDHR, such as pitch, height, wall temperature of the HXs and the interval of spacer grids, was checked and determined in a viewpoint of better heat removal capacity and compact construction. The pitch between HXs was determined as 20 cm with the outer diameter of HXs were 5.08 cm, and the height of the HXs was decided as 10 m through the height sensitivity study. Based on the results, the numbers of HXs and PCCTs were analyzed considering the decay heat of 3 days after the shutdown to suggest the overall design of the APDHR. The heat transfer coefficients were 10.07 W/m2K and 15.76 W/m2K in the case of the bare and the finned HXs, respectively. Therefore, the numbers of HXs and PCCTs required can be reduced by using the finned HXs.

Experimental study on the thermal stratification in a pool boiling with a horizontal heat source

Large pools of water near atmospheric pressure have been incorporated into several advanced nuclear reactor designs. These pools provide a heat sink to remove heat from the reactor or containment by natural circulation and are a source of water for core cooling. Thermal stratification is formed in horizontal fluid layers with different temperatures, where the warmer fluid layers are situated above the cooler fluid layers. In the present study, the key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The two-dimensional velocity vector fields that occur as the water temperature increases to the saturation temperature are experimentally investigated in a pool that contains a horizontal heater rod. The detailed velocity measurements using the 2D PIV measurement technique were conducted to investigate single- and/or two-phase natural convection flow and thermal stratification in a pool boiling. The experimental results indicate that a large natural convection flow occurs above the heater rod and that thermal stratification occurs below the heater rod. The flow in the opposite directions to each other was shown in the region between the heater rod and the thermal boundary layer. This flow pattern will contribute to maintain the thermal stratification and retard the water temperature rise and the dissipation of the thermal boundary layer in a pool boiling with a horizontal heat source installed asymmetrically. The current experimental data can be directly used to develop and validate turbulence models in a pool boiling. Furthermore, the present experimental results can help characterize the performance of boiling models on heated surfaces and provide benchmark data to validate the calculation performance of a component scale analysis code and CFD code.

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS).This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS.

ICONE23-2089 INTEGRAL EFFECT TEST ON OPERATIONAL PERFORMANCE OF PAFS (PASSIVE AUXILIARY FEEDWATER SYSTEM) FOR CIV OPENING STROKE

PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). PAFS-CIV-01 test was performed for validating cooling rate of the reactor according to the CIV opening stroke at the FLB accident, which was analyzed as the most severe case in the APR+ SSAR (Standard Safety Analysis Report). With an aim of simulating a FLB+CIV accident of the APR+ as realistically as possible, the three-level scaling methodology was taken into account to determine the test conditions of the steadystate and the transient. From the experiment, major thermalhydraulic phenomena such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate in PAFS were investigated and discussed. With decreasing the flow area of the CIV, the gradient of system pressure and temperature were reduced due to the decrease of the heat removal performance. As the CIV was re-opened, the two-phase natural convection flow in the loop of the PAFS was recovered. From the present experimental result, it could be concluded that the cooling rate of the core was controlled by the adjustment of the CIV opening stroke when the APR+ PAFS was operating.

Experimental investigation on the natural convection flow in pool boiling

In the present study, the key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The volumetric scaling ratio of the test rig is 1/910 the size of the passive auxiliary feedwater system (PAFS) condensing heat removal assessment loop (PASCAL), which is a PAFS performance evaluation test facility. The two-dimensional velocity vector fields that occur as the water level decreases are experimentally investigated in a pool that contains a horizontal heater rod. The 2D particle image velocimetry (PIV) measurement technique is adopted to determine the velocity vector field of the natural convection flow. The experimental results indicate that a large natural convection flow occurs above the heater rod and that thermal stratification occurs below the heater rod. The thermal stratification and the stagnant region begin to disappear when the pool temperature reaches approximately 90°C. The experimental results can provide benchmark data to validate computational fluid dynamics (CFD) calculations of thermal hydraulic phenomena that occur in a pool with a heat source.

Thermal-hydraulic evaluation of passive containment cooling system of improved APR+ during LOCAs

Various reactor concepts and technologies have been devised and evaluated to ensure the integrity of the core and the containment under a prolonged station blackout. After the successful validation of the passive auxiliary feedwater system (PAFS) of Advanced Power Reactor Plus (APR+), Korea is considering an improved APR+ with a passive containment cooling system (PCCS). In a previous paper, we suggested a PCCS design based on APR+ and performed a scoping analysis. We performed a MARS simulation for the thermal hydraulic evaluation of the system behavior, including natural circulation through the Passive Condensation Cooling Tank (PCCT) water pool and inside PCCS tubes as well as steam–air mixture condensation inside the containment. Through a simulation using the MARS system code, we investigated the effect of air holdup tanks (AHTs) on reduction of the air fraction in the containment, the effects of several steam–air mixture condensation models, the role of heat structures, and the flow instability inside the PCCS tubes. We found that the presence of AHT reduced the number of required PCCS tubes by more than half and the heat resistance of the steam–air mixture side is dominant in terms of governing the overall performance of PCCS. The embedded MARS condensation model and Uchida's correlation gave lower heat transfer coefficients than Dehbi's correlation, and heat structures removed more decay heat than PCCS tubes. Finally, intense flow instability inside the PCCS tubes was observed and was mitigated by placing orifice plates at the inlet of the PCCS tubes, increasing the height of the return nozzle, or increasing the tube angle.

Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal-hydraulic code, named CUPID, has been developed. In the present study, the CUPID code was applied for the simulation of the PASCAL test facility constructed with an aim of validating the cooling and operational performance of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor + (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. This paper introduces the simulation results for the passive condensate cooling tank (PCCT) of the PASCAL facility performed with the CUPID code in order to investigate the thermal-hydraulic phenomena in the PCCT. The simulation showed that the important thermal-hydraulic characteristics in the PCCT, such as two-phase natural circulation and boil-off phenomena, can be successfully reproduced by CUPID. Two important validation parameters, collapsed water level and local liquid temperature, were quantitatively well captured in the simulation. This paper presents the description of the PASCAL test facility, the physical models of the CUPID code, and its simulation result for the PCCT.

An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. 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. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance.

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.

MULTI-SCALE THERMAL-HYDRAULIC ANALYSIS OF PWRS USING THE CUPID CODE

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

Design of condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) of APR+ (Advanced Power Reactor Plus)

The APR+ (Advanced Power Reactor Plus), a next generation nuclear power plant in Korea, has adopted the PAFS (Passive Auxiliary Feedwater System) on the secondary system of the steam generator (SG) as an advanced safety feature. It is intended to replace the conventional auxiliary feedwater system, which consists of active components for the SG in a passive way. It removes decay heat from the reactor core by cooling down the secondary system of the SG using a condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank).The objective of this study is to design a condensation heat exchanger for the PAFS and to evaluate the cooling performance for the proposed design using the thermal hydraulic system analysis code, MARS (Multi-dimensional Analysis for Reactor Safety). Requirements such as the heat removal capacity and the prevention of water hammer were preferentially considered to determine the design parameters of the heat exchanger tube. The MARS code analysis result showed that the proposed design of the PAFS heat exchanger is able to cool down the required amount of decay heat. The distribution of a liquid volume fraction and flow regime predicted by the MARS code shows that the proposed design of the heat exchanger excludes the water hammer inside the tube. Estimation of a two-phase flow pressure drop indicates that the pressure drop inside the tube is negligible compared to the total pressure drop in the PAFS. From the MARS code analysis, it is concluded that the proposed design of the condensation heat exchanger in the PAFS satisfies the overall criteria for the performance of the passive heat removal system in APR+.