In-containment Refueling Water Storage Tank Research Articles

Numerical Investigation of Nanofluid’s Heat Transfer Performance in Passive Residual Heat Removing System of AP1000 Nuclear Reactor

The Passive Heat Removal system (PHRS) is designed to remove the residual heat from the core in case of a station blackout, failure of emergency core cooling system, or failure of feedwater supply through the Passive Residual Heat Removal Heat Exchanger (PRHR HX). PRHR HX consists of a C-shaped tube bundle as a heat exchanger and the In-Containment Refueling Water Storage Tank (IRWST) as a heat sink. A temperature distribution of this passive heat removal system of an AP1000 Reactor is generated using COMSOL Multiphysics and the heat transfer coefficient is calculated to illustrate the effectiveness of the PHRS. A comparison of the heat transfer coefficient between the IRWST filled with water and nanofluid has been generated using the PRHR HX design. Thermophysical properties of nanofluids have been calculated in the process of calculating the heat transfer coefficient. Numerical results show the difference in temperature reduction of Al2O3, TiO2, and Ag as opposed to water in the IRWST. Time-dependent heat conduction of water and nanofluid results contribute to the effective analysis of passive heat removal systems and provide information for the safe operation of AP1000 reactors. By the end of 2024/2025, two VVER-1200 power stations with a combined capacity of 2400 MW will be operating in Bangladesh. For safety and licensing reasons, heat transfer simulation of VVER-1200 can be performed using COMSOL software.

Experiment on the Excitation of Vapor Behavior and the Vibration by Thermal Flow Under C-Tube Pool Boiling Model

Abstract As an important safety equipment in nuclear system, the passive residual heat removal heat exchanger (PRHR HX) with C-tube is responsible for timely removing heat from the reactor core in accident. A C-tube heating device was setup to investigate the flow characteristics of the in-containment refueling water storage tank (IRWST), which is of the PRHR HX shell-side flow field and the vibration characteristics of the heat exchanger tube. The temperature of the flow field and the acceleration signal of the C-tube were measured. The C-tube frequency in the pool boiling condition is 0–0.660 Hz. There is no resonance with the first natural frequency of 8.98 Hz. The IRWST temperature field can be divided into slow-heating period, fast-heating period, and boiling period. Steam vapor exists in three periods, condensation and collapse of vapor concentrate on horizontal tube. The fast-heating period exists the subcooling degree ΔTsub for the most intense vapor excitation, ΔTsub is in the range of 13–16 °C. In term of thermal flow excitation vibration, the existence of the cross flow with C-plane direction is in the transition period from the slow-heating period to the fast-heating period, which excites the maximum root mean squares (RMS) amplitude 0.039 mm and 0.034 mm at the midpoint of upper horizontal tube and vertical tube. When the fluid enters fast-heating period, the direction of flow changes into the normal C-plane. Normal RMS amplitude reached maximum 0.02 mm at upper 1/4 of the vertical tube in boiling period. The above research provides important data support for the safety and reliability of PRHR HX.

Overview of separate effect and integral system tests on the passive containment cooling system of SMART100

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

Probabilistic safety assessment has been widely used to evaluate nuclear power plant safety systems. It couples fault tree (FT) analysis and event tree (ET) analysis. FT analysis is to develop failure scenarios, to quantify the failure probability and to identify critical components. ET analysis is to develop event sequences (ESs) leading to reactor core damage and quantify core damage frequency. Currently, there is no approach in probabilistic safety assessment that can be used to identify minimal combinations of safety system failures leading to reactor core damage. This study proposes an event sequence based fault tree analysis, which integrates FT model with ET model. The ET model is to identify the ESs leading to the reactor core damage. Meanwhile, the FT model is to identify minimal combinations of those ESs found in the ET model. The motivation of this study is on how to identify critical safety systems in nuclear power plants to mitigate disturbances caused by a group of postulated initiating events to avoid core damage. To confirm the feasibility and the applicability of the proposed approach, the performance of the AP1000 passive core cooling system is evaluated. It is found that if in-containment refueling water storage tank, automatic depressurization system - full, accumulator, core make-up tank, and passive residual heat removal work properly, the reactor core will intact. These results confirmed that the proposed approach can be applicable to identify minimal combinations of safety systems to keep the reactor core intact.

Preliminary design of the suppressive containment system based on HPR1000

The current HPR1000 has a high construction cost and weak economic competitiveness due to its complex system and large containment. This study proposes a conceptual design of an advanced containment system based on HPR1000, which aims at reducing the construction cost of nuclear power plants without reducing their safety. In the design, the In-containment Refueling Water Storage Tank (IRWST) and spray system were discarded, and the space of the double containment vessel was used to create an annular cavity pool (ACP). The transient pressurization in the containment caused in the early stages of the accident can be quickly suppressed by ACP instead of the traditional large-volume suppression concept. And due to the significant condensation of steam within the ACP water, the transportation of air into the ACP results in a reduction in the air mass fraction in the vicinity of the passive containment heat removal system (PCHR) located within the containment structure. This phenomenon subsequently leads to an enhancement in heat transfer performance. In addition, ACP also serves as the water source for the safety injection system, the reactor cavity water injection system, and the core replacement system. Compared with HPR1000, the system of the preliminary design has been greatly simplified, and the size of the containment vessel has been reduced by nearly 47%. The simulation of the large-break loss-of-coolant accident (LBLOCA) scenario has been conducted to examine the response of the containment system, specifically focusing on the performance of the safety injection system and suppression system. The results show that the integrity of the containment and core was ensured throughout the accident, and there was a safety margin.

Three-dimensional numerical research on coupling heat transfer characteristics of PRHR HX under different water levels based on the experimental model

The passive residual heat removal system (PRHRs) is an important heat transfer system to ensure the removal of residual heat in non-LOCA conditions occurring in advanced Generation III nuclear reactors. The water level of the In-containment Refueling Water Storage Tank (IRWST) decreases due to long-term water evaporation under accident conditions and the later stages of PRHRs operation. Additionally, the water level also decreases during the water leakage accidents. The reduction of water level contributes to a decrease in the heat transfer area and the passive residual heat removal ability of the core. In order to investigate the coupling heat transfer characteristics of both primary and secondary side in the heat exchanger under different water levels, Three-dimensional numerical simulations of the PRHR HX's operation under different water levels was carried out based on the scaled-down IRWST&PRHR HX experimental model. The fluid flow and heat transfer processes inside and outside the C-type tubes were analyzed in detail. The results indicate that when heat transfer area is reduced by 50%, the overall heat transfer rate is reduced by about 41%, meanwhile the average heat transfer coefficient is about 37.4% higher than that under the full water level condition. It indicates that PRHR HX can maintain high passive residual heat removal capacity even under half water level, which provides an important reference for the design and safe operation of the PRHR HX.

Experimental and theoretical research on heat transfer characteristics of the scaled-down C-type PRHR HX model under different cooling water levels

Passive Residual Heat Removal heat exchanger (PRHR HX) is the key equipment of the passive residual heat removal system in AP1000 reactor. It is designed in special C-type heat exchange tube bundle and locates in the In-containment Refueling Water Storage Tank (IRWST). In some accidental events, the cooling water in the IRWST may leak, evaporate, or passively inject into the primary loop, leading to the drop of water level. The corresponding temperature, flow, and heat transfer effects may change under different water level conditions. In the present work, the heat transfer performance of PRHR HX model under different cooling water levels conditions are studied based on the modified scaled-down PRHR HX&IRWST experimental facility. The temperature distributions in both primary and secondary side distributions are obtained. Meanwhile, key parameters including the heat transfer quantity, heat flux, and heat transfer coefficient are analyzed in depth. The experimental results indicate that though the total heat transfer quantity decreases as the water level of the water tank drops, but due to the increase of average heat flux, the residual heat removal effect can also be maintained at a relatively high level. Furthermore, combined correlations under both nucleate boiling and natural convection are employed to depict the heat transfer mechanisms and predict the heat transfer coefficient of different PRHR HX tube bundle sections. Both the experimental and theoretical results provide important references for the operation and safe analysis of the prototype PRHR HX.

Multi-Scale study of an innovative safety system for Pressurized water reactors

After the FUKUSHIMA accident, passive systems become an important issue for the new projects of Pressurized Water Reactor (PWR). In response to the demands of its partners, the CEA (French Commission for the Atomic Energy and Alternative Energy) has proposed suitable systems to deal with different accidental sequences. Then, it has dimensioned their components and modelled their integration in the reactor using the CATHARE system code. In particular, the CEA has worked on the ADS system (Automatic Depressurization System) devised to tackle with the well-known LOCA accident (Loss of Coolant Accident), resulting from a breach in the pipes of the primary circuit.Concretely, the ADS system is made up of a tank called IRWST (In-Containment Refueling Water Storage Tank), placed high in the containment, which has the role of achieving a safety passive injection to replace the conventional one using low-pressure pumps, during a LOCA accident. The ADS is in fact already present on the AP1000 reactor, the US PWR designed to be totally passive. The CEA has proposed to improve it by adding a LPWT tank (Low-Pressure Water Tank) pressurisable by steam issued from the pressurizer on the last line of the ADS system. It should allow a better core cooling by the additional water injection in the reactor core, anticipated in time compared to that of the IRWST.The CFD application in reactor safety studies, due to its high-level of accuracy, is now widespread. The CFD calculation is used to predict local fluid temperature in the different regions of a nuclear reactor in nominal conditions and transient situations. Local CFD calculation coupled with a component system code for the whole domain of the reactor is also a current approach (Guelfi et al.).The study presented here exclusively focuses on the LPWT water pressurization phase prior to its injection into the reactor core. Three approaches are exposed:- An analytical approach, in which one considers the pressurization of the volume of gas present in the tank without taking into account the liquid phase;- A system approach with the CATHARE thermal–hydraulic code which allows to describe at the same time the volume of gas and the volume of liquid contained in the tank;- A CFD approach using the NEPTUNE_CFD code to determine the local evolution of the quantities that characterize the system. This approach will notably make it possible to describe the movement of the interface and the velocity profile of the water vapor. It will also help to bring a better understanding of the physical phenomena involved during the LPWT water pressurization phase.At the end, these different approaches show that the increase in pressure by the steam injection is an interesting option to pressurize the water tank. On the other hand, the results of the pressure evolution obtained by the system approach are very close to those obtained analytically, in the absence of liquid phase. However, they are overestimated compared to the CFD approach, which, thanks to a better modeling of the condensation of the injected vapor, leads to a more realistic evaluation.

Analysis of Post-Accident Debris Transport Performance in In-Containment Refueling Water Storage Tank in Nuclear Power Plants

To avoid the blockage of strainer in the in-containment refueling water storage tank (IRWST) and guarantee the safe operation of the pump downstream the IRWST after accident occurrence, the debris transport performance in the IRWST must be evaluated carefully. For the double-loop pool-type IRWST of a nuclear power plant (NPP), the Computational Fluid Dynamics (CFD) method is adopted to simulate the flow field of this IRWST, and the volume ratio of the high velocity region and high turbulent kinetic region is used to quantitatively evaluate the debris transport performance of the IRWST. The results indicate that the debris transport ratio of IRWST does not exceed the design value of the strainer under various post-accident conditions, which ensure the safety of the strainer and its connected systems after accident occurrence; the strainer load reaches the maximum value when only the strainer A in the inner loop is put into operation; and the post-accident debris transport performance is mainly affected by the high velocity region in the flow field. As a result, this paper proposes the optimization scheme, namely, increasing the diameter of the mixing pipeline in the outer loop, for the existing IRWST layout. This scheme can reduce greatly the debris transport ratio in the IRWST after accident occurrence and thus improve the post-accident safety of the NPP.

Experimental Study on Sensitivity of PRHR Pipeline Break Location on ACME Test Facility

To study the safety performance of the advanced passive (AP) nuclear power plant (NPP) under passive system failure condition, the experimental study on the loss of coolant from the passive residual heat removal (PRHR) pipeline break is performed by the advanced core-cooling mechanism experiment (ACME) facility, during which the effect of main test sequences and break location on the key parameters in different phases of the accident is analyzed. As demonstrated by the study results, the ACME PRHR pipeline break test sequences, basically same as those for the small-break loss of coolant accident (SBLOCA) of the cold leg, reproduce the safety characteristics in the natural circulation phase of the passive NPP, blowout phase of the automatic depressurization system (ADS) and safety injection (SI) phase of the in-containment refueling water storage tank (IRWST); in the test at different break locations, the passive core cooling system (PXS) can always ensure the water makeup for core and the core active region remains below the mixed liquid level; and the break locations have notable effect on the ACME LOCA accident sequence, initial depressurization rate of reactor coolant system (RCS), PRHR heat exchanger (HX) flow, blowout flow, core level, IRWST SI flow and other parameters, and have little impact on the SI flow of the core makeup tank (CMT) and accumulator (ACC).

Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600

Abstract The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.

Numerical Study of the Impingement of Water Film on a Small Attached Bulging Plate on a Vertical Plane

In the passive containment cooling system of AP1000, the condensed water is expected to flow down on the inner surface of the steel wall of the containment, and recover to the in-containment refueling water storage tank (IRWST), therefore, to maintain the long-term coolability of the passive residual heat removal system. However, there are attached bulging plates on the inner surface for various engineering needs, such as supporting, and the impingement of condensed water film on these bulging plates can reduce the amount of the recovered water. In this article, a 3-D Eulerian wall film model in FLUENT was used to study a series of flow behaviors when a water film impinged on the bulging plate on a plane surface. The loss ratio of falling film impinging on attached plates of different sizes under different flow rates were calculated and in good agreement with the experiment results. Four stages of the film behavior during the impingement were identified and analyzed; in addition, the influences of the bulging height of attached plate and flow rate were studied. And a correlation between the loss ratio of impinging water film, the bulging height of the attached plate, and the Weber number was obtained.

An Extended Thermosyphon Cooling System for APR-1400 Reactor Design

Advanced Power Reactor (APR-1400) is a Generation III + Pressurized Water Reactor (PWRs) and has gained popularity among energy mix community. APR-1400 features enhanced safety limits to prevent a “Fukushima-type” accident scenario for a duration up to 8 h. This work proposes a system that enhances APR 1400 passive cooling (Thermosyphon Cooling System, TCS) capabilities to guarantee at least double the coping time before core meltdown. TCS provides enough cooling to the reactor core in the case of loss of offsite power, known as Station Black-Out (SBO) accident. This work invites to use the In-containment Refueling Water Storage Tank (IRWST) to further extend the APR1400 capability to passively cool the reactor core after an SBO accident. This work is designed and tested by 3Keymaster simulator platform developed by Western Service Corporation. Simulation has modified the existing model of APR1400 in the simulator by adding pipes, valves and heat exchanger. Results show the system will cool the reactor core for almost 18 h after the SBO accident. The system works initially in parallel with steam generators then it continues cooling the reactor core for extra 18 h resulting in providing more time for restoring the on-site power sources and preventing core meltdown.

Development of severe accident mitigation technology and analysis for SMART

Severe accident mitigation technology was developed and evaluated for a small integral reactor of SMART. The containment pressure and hydrogen behavior were analyzed using MELCOR computer code (Sandia National Laboratory, 2018) in the SBO (Station Black Out) sequence, which was selected from PSA (Probabilistic Safety Assessment) results for the SMART. The severe accident mitigation technology to improve the SMART safety include reactor vessel depressurization using the ADS (Automatic Depressurization System) to prevent DCH (Direct Containment Heating) in case of a reactor vessel failure, a reactor cavity flooding using the CFS (Cavity Flooding System) with the IRWST (In-containment Refueling Water Storage Tank) for the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) to prevent the reactor vessel failure, and a hydrogen control system of PARs (Passive Autocatalytic Recombiners) to remove hazards from hydrogen combustion considering the amount of hydrogen to be generated by 100% fuel cladding oxidation. The MELCOR results showed that the containment pressures during the SBO sequence was below the design pressure, which meant that the containment integrity is maintained during a severe accident in the SMART. The hydrogen mole fraction of the containment was much lower than the hydrogen safety criteria of 10 vol%, which meant that the possibility of the hydrogen burn in the containment was negligible. For this reason, it could be concluded that the hydrogen was controlled by the hydrogen moving path and the hydrogen control system of PARs in the SBO sequence of the SMART.

AP1000 IRWST numerical analysis with GOTHIC

The AP1000 Passive Residual Heat Removal (PRHR) system plays a significant role as it helps to remove the core decay heat, using the In-containment Refueling Water Storage Tank (IRWST) as a heat sink. The IRWST is located above the core, promoting natural circulation and allowing to be the mid-term heat-sink for the reactor core. The thermo-hydraulic pool behavior during an accident has an influence in the containment pressure and temperature, as happens in the pressure suppression pool in BWR reactors.Due to the importance of the IRWST performance on the AP1000 containment, a numerical analysis of a scaled experiment has been done using the GOTHIC code. Firstly, an IRWST scaled-down model is created, a mesh sensitivity analysis is performed, and the results obtained are compared against experimental results available in the literature. It is found the importance of the mesh discretization for the proper thermal-stratification modeling. Then, the reference model setup and mesh are applied in a full-scale prototypic model. The mesh influence on the thermal stratification is analyzed, as well as its impact on the AP1000 containment pressure. The mesh and setup for the full-scale model are selected to be implemented in a full containment 3D model in future works.

Investigation on transient thermal hydraulics of reduced scale passive residual heat removal heat exchanger in tank

In advanced nuclear power plant AP1000, the passive residual heat removal heat exchanger (PRHR HX) is a key part of the passive safety system. It can transfer decay heat from the reactor core to the water in the in-containment refueling water storage tank (IRWST) during the accidents. The secure operation of the C-shaped PRHR HX has a great impact on the reactor safety. In the present work, the heat transfer performance of a reduced scale PRHR HX was investigated by experiments and numerical simulations. A system including a 6 × 7 C-shaped tube bundle and a rectangular tank was studied. A three-dimensional numerical model using porous media method was developed to explore the transient single phase thermal hydraulic characteristics of the PRHR HX. The distributed resistances caused by the C-shaped tube bundle were calculated by empirical correlations. Both the heat transfer coefficients from the hot fluid in the tube to the inside surface of tube and that from the outer tube surface to the water in the IRWST were calculated by empirical correlations published in literatures. Tests were performed on the 6 × 7 C-shaped tube bundle installed in the middle of the tank. The outer wall temperature and bulk water temperature in the tank were measured by thermocouples. The numerical methods were validated by the experimental results. The calculated results of the local temperature distribution and heat transfer rate were both well consistent with the experimental values. The deviation of the heat transfer rate was smaller than 5%.

Experimental study of the steam condensate dripping behavior on the containment dome

Under accident conditions of CAP1400, recycling of condensate film from the inner containment wall is a significant aspect for the water level maintenance of the in-containment refueling water storage tank, whilst dripping from the containment dome is the dominant factor that causing the condensate loss. Therefore, experimental investigations on the dripping phenomena were carried out in this study. A pressure vessel was set to simulate the condensing atmosphere in the containment, in which a 1.5 × 0.6 m2 rotatable test section was suspended with similar surface condition as CAP1400. Experimental results show that the condensate flow patterns could be divided into four types. It was found that dripping was triggered by the condensate mass flow flux exceeding the critical value on an unobstructed condensing surface. Meanwhile, the dripping fraction increases with the difference between condensate mass flow flux and the critical value. Besides, the effects of inclination, bulk pressure, air concentration etc. on dripping were experimental analyzed. In general, this study hopes to provide data support and theoretical guidance for the further studies of the condensate loss under accident conditions.

Experimental investigation on natural convection and thermal stratification of IRWST using PIV measurement

Passive Residual Heat Removal Heat Exchanger (PRHR HX) immerged in the In-containment Refueling Water Storage Tank (IRWST) plays an important role in removing the core decay heat under non-Loss of Coolant Accident (LOCA) accident. In the initial stage of the Passive Residual Heat Removal System (PRHRS) activating, the natural convection is the main way to exchange heat between the secondary side of the PRHR HX with IRWST. However, it is almost impossible to acquire the thermohydraulic performance in IRWST since the influence of the natural convection and thermal stratification in IRWST happens simultaneously. In this paper, Hierarchical Two-Tiered Scaling (H2TS) method is employed to get scaling criteria based on control equations. Anda visualized water tank with five C-shape electrical heating tubes was scaled-down and applied to investigate the thermohydraulic performance for IRWST of CAP1400 Nuclear Power Plants (NPPs). Particle Image Velocimetry (PIV) measurement was used to investigate the evolution of the flow field of total twenty-five surfaces in three axial directions under three steady- and two variable-heating conditions. Meanwhile, three thermocouple bundles were adopted to acquire temperature data in three regions. The experimental results show that the temperature deviation in the same height of the IRWST is flattened by plenty of vortexes. As a result, local circulation and the temperature difference is less than 1 °C. While the thermal stratification along the vertical direction is obvious, the distribution of flow field and temperature demonstrate that a “dead zone” exists in bottom of the IRWST. Besides, a “thermal interface” region forms in the lower middle region as hot and cold fluid mixing. An obvious upwelling flow is observed near the PRHR HX heating rod bundle from flow fields of three axial directions. At the same time, the maximum height of the upwelling flow decreases with time increasing. On the one hand, the natural convection is induced by the thermal stratification. What is more, the thermal stratification will impair the scale and strength of natural convection, which will intensify the thermal stratification conversely.

Scaling and integral effect test on transition from depressurization to long-term gravity injection

The passive safety commercial pressurized water reactors (PWRs), AP1000 and CAP1400 are equipped with the automatic depressurization system (ADS). The fourth stage of ADS (ADS-4) has the largest discharging capacity among the all ADS stages. During a small break loss-of-coolant accident (SBLOCA), only when the system is fully depressurized by the ADS-4 to near the containment atmospheric pressure, the in-containment refueling water storage tank (IRWST) can deliver the large coolant inventory for the long-term core cooling by its water head. Before the IRWST injection, the core reaches its minimum inventory. So the transition from the ADS-4 depressurization to the IRWST injection is a very challenging period in a SBLOCA. Therefore, to evaluate the passive system performance by the integral effect test (IET), the transition must be simulated properly. This work conducted the scaling analysis for the ADS-4 depressurization and the IRWST injection for designing the related components of the ACME IET facility. A uniform scaling law based on the existing T-branch liquid entrainment models was proposed. For scaling the onset of the IRWST injection, the requirement for reproducing the synergetic effect in an IET was introduced in order to preserve the important event/process occurrence time with the proper conditions, and the scaling laws for the full pressure IRWST injection were obtained. For investigating the transition thermal hydraulic behaviors and the ADS-4 configuration influences on the core safety, one cold leg break test and three double-ended direct-vessel-injection (DEDVI) line break tests with the different ADS-4 failure modes were selected from the ACME SBLOCA test matrix, and the main test data related to the core safety were analyzed and compared. It was found that the ADS-4 actuation significantly accelerated the system depressurization but worsened the core inventory depletion at the same time. In addition, the break condition and the ADS-4 failure mode could differently affect the system depressurization and the core level. The ACC injection during the ADS-4 depressurization in DEDVI was critical for the core level makeup. The insufficient ADS-4 venting capacity delayed IRWST injection, but also mitigated the core inventory depletion. It suggests the ADS-4 design can be potentially optimized in the passive safety system design.

CFD simulation of secondary side fluid flow and heat transfer of the passive residual heat removal heat exchanger

Abstract Passive residual heat removal heat exchanger (PRHR HX) is a key equipment in advanced passive safety pressurized water reactors (PWRs), such as AP1000. Immerged in the In-containment refueling water storage tank (IRWST), the C-shape PRHR HX removes the residual heat using natural convection and boiling heat transfer during postulated accidents. Therefore, the safety operation of the PRHR HX is very important for the nuclear power plant (NPP). In this paper, three dimensional CFD simulation of the secondary side fluid flow and heat transfer of the PRHR HX is performed. The drift flux model is used to simulate the two phase flow phenomenon in the IRWST. The tube region is modeled by the porous media approach. The flow resistance in the tube region is calculated by empirical pressure drop correlation and two phase flow multiplier. The heat transfer rate from the primary side to the secondary side fluid is also evaluated by widely used heat transfer empirical correlations. Additional source terms corresponding to the flow resistance and removed heat in the tube region are added to the momentum and energy equation, respectively. The governing equations are solved by the commercial CFD package FLUENT. Three dimensional distributions of the fluid velocity, temperature and void fraction are obtained. The heat transfer characteristics of the tube bundle is analyzed.