Passive Residual Heat Removal Research Articles

Analysis of transient characteristics and design improvement of the passive residual heat removal system of NHR-200-II

NHR-200-II is a small integrated pressurized water reactor with 200 MW core thermal power. The core heat is transferred to two independent intermediate circuits via fourteen in-vessel primary heat exchangers (PHE), and the heat in the intermediate circuits is transferred to feedwater by two steam generators (SG) in the two intermediate circuits respectively. A passive residual heat removal (PRHR) branch is connected to each intermediate circuit to remove core decay heat under postulated accidents. During normal operation, PRHR branches are isolated by valves while SG branches in intermediate circuits are open. The valves in PRHR branches will be opened and the isolation valves of SG branches will be closed during decay heat removal scenarios. The decay heat removal capacity of NHR-200-II PRHRS could be seriously deteriorated once the isolation valves for SG branches fail to close, which was confirmed in a scaled integral test loop previously. Current understanding of PRHRS’s thermal-hydraulic characteristics with possible isolation failure in SG branches is limited. In this paper, the NHR-200-II PRHRS is modeled with RELAP5 considering the case of success and fail to isolate SG branches. A series of numerical simulations are carried out to study the impact of various parameters, such as the initial temperature, the size of the intermediate circuits’ header, and the initial flow direction in the intermediate circuits. Oscillatory flow is found when SG branches fail to be isolated under certain parameters combinations. An improved PRHRS design is purposed to eliminate possible flow oscillations, and the purposed improved design are tested by numerical simulations.

Analysis of a parallel-channel natural circulation loop in the passive residual heat removal system of the NHR200-II reactor

The passive residual heat removal (PRHR) system of the NHR200-II nuclear heating reactor is a parallel-channel single-phase natural circulation loop. The reverse-flow phenomenon in parallel-channel natural circulation loops will cause insufficient heat removal capacity in the PRHR system. In this study, a simplified mathematical model was established for the PRHR system of NHR200-II. The mechanism of the reverse flow phenomenon in the PRHR system was analyzed using a pressure drop-flow rate diagram, and the effects of cooler elevation and loop resistance on the number of reverse flow branches were discussed. The analysis reveals that the reverse flow phenomenon of the PRHR system can be suppressed by increasing the elevation of the air cooler. Furthermore, there exists a critical elevation of the air cooler. It is not feasible to suppress the reverse flow by changing the friction coefficients of either the PHE branches or the air cooler branch when the air cooler elevation is less than the critical value. Oppositely, the number of reverse flow branches can be reduced or even eliminated by increasing the resistance of the PHE branches or reducing the resistance of the air cooler branch when the air cooler elevation is greater than the critical value.

Numerical simulation of the operational characteristics of air-cooled PRHR under SBO accidents

In this paper, the operational characteristics of air-cooled type passive residual heat removal (PRHR) system of the nuclear heating reactor are studied with RELAP5 code. The PRHR uses the air natural circulation which can transfer heat directly from the reactor vessel to the air, and its heat transfer process is particularly important for the reactor safety. However, the current thermal–hydraulic code lacks the relevant air cooler model. A method how to model an air cooler is proposed based on the conservation equation and can be widely applied to different situations. Moreover, the operating characteristics of air cooling tower and the transient thermal–hydraulic characteristics of station blackout (SBO) event have been analyzed. The results show that the heat transfer coefficient of air cooler increases with the heat transfer power, and the inlet temperature has little influence on the heat transfer coefficient. The air cooling type PRHRS can transfer the core residual heat timely and continuously. There is a time delay of the heat exchanger power between the main heat exchanger and air cooler caused by the temperature distribution.

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).

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 THE START-UP FEATURES OF SECONDARY-SIDE PASSIVE RESIDUAL HEAT REMOVAL SYSTEM

In order to study the start-up characteristics of passive residual heat removal (PRHR) system, a numerical model is established to calculate the transient characteristics of secondary-side PRHR system under station blackout (SBO) accident. The results show that two-phase flow oscillation occurs during the establishment of natural circulation. This is caused by the mismatch between the variation of reactor residual heat power and the PRHR system operation power. In this paper, a thermal energy storage device is introduced to remove the high residual heat power of the reactor at the initial stage of shutdown. The latent thermal storage device using phase change materials (PCMs) as heat storage medium has the advantage of high potential heat storage density and almost constant phase change temperature. This new configuration can eliminate the flow oscillation phenomenon and reduce the maximum pressure of the steam generator.

CFD investigation on thermal hydraulics of the passive residual heat removal heat exchanger (PRHR HX)

Passive residual heat removal (PRHR) system is a very important component of the passive safety systems in advanced passive safety pressurizer water reactors (PWRs) such as AP1000. The passive residual heat removal heat exchanger (PRHR HX) is a C-shape tube bundle heat exchanger immerged in the In-containment refueling water storage tank (IRWST) which removes the core decay heat during the accident transients. The performance of the PRHR HX is significant for the safety of the nuclear power plant (NPP). In this paper, the thermal hydraulics characteristics of the PRHR HX in the IRWST is analyzed using computational fluid dynamics (CFD). The tube region is modeled by the porous media approach along with the distributed resistance method. Heat transfer from the primary side fluid inside the tube to the secondary side fluid in the IRWST is considered. The simulation is carried out by the commercial CFD package FLUENT. The calculation of the flow resistance and heat transfer in the tube region is implemented using the User Defined Functions (UDF) in FLUENT based on the local flow conditions. Three dimensional distributions of the fluid velocity and temperature in the IRWST are obtained and thermal stratification is observed. The PRHR HX heat transfer capacity and the primary side fluid temperature distribution inside tubes are analyzed.

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no-load conditions

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

AP1000® SBLOCA simulations with TRACE code

The AP1000® is an advanced pressurized water reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The performing of such systems must be evaluated through the performance of experiments and simulations with a variety of thermal–hydraulic codes. This paper presents the results which has been obtained for different SBLOCA break sizes with the best estimate TRACE V5.0 patch 2 thermal–hydraulic code and their comparison with those obtained by Westinghouse with NOTRUMP code. The main results show that TRACE code predicts a similar trend in all sequences with some differences that are expected to be an issue of the more conservative models and hypothesis assumed in the SBLOCA licensing analysis performed with NOTRUMP. Some particular characteristics of this reactor are also shown in this paper such as the importance of core makeup tank (CMT) and passive residual heat removal (PRHR) system for core cooling in case of small break sizes and the behavior of the plant in case of availability of half of the total passive safety injection systems which is the case of the double-ended direct vessel injection line break (DEDVI).

Experimental Study on the Performance of the Passive Safety Injection in an IIST

A passive core cooling system (PCCS) has been installed at the Institute of Nuclear Energy Research Integral System Test facility. It includes three core makeup tanks (CMTs), three accumulators (ACCs), a four-stage automatic depressurization system (ADS), a passive residual heat removal (PRHR) heat exchanger submerged into an in-containment refueling water storage tank (IRWST). The purpose of this research is to study the performance of the PCCS with passive injection during either a pressure balance line (PBL) break or a direct vessel injection (DVI) line break. Five experiments were performed simulating break area ratios of 0.5 to 2.0% (1.88 to 3.77 mm) at either a PBL or a DVI line. The general system response and the interactions of CMT, ACC, PRHR, and IRWST to the effect of core heat removal are observed and discussed. The experimental results show long-term core cooling can be reached for the cases of the PBL break and the DVI-line break following the PCCS actuation procedures.

Heat Transfer Analysis of the Passive Residual Heat Removal System in ROSA/AP600 Experiments

The passive residual heat removal (PRHR) system in the Westinghouse AP600 advanced passive reactor design is a natural-circulation-driven heat exchanger cooled by the water in the in-containment refueling water storage tank (IRWST). During the experiments, which simulated small-break loss-of-coolant accidents in the AP600 reactor using the ROSA-V Large-Scale Test Facility (LSTF), the PRHR system heat removal rates well exceeded the core decay power soon after the actuation of the PRHR. This resulted in continuous cooldown and depressurization of the primary side. The PRHR heat transfer performance in these experiments was analyzed by applying heat transfer correlations available in literature to the PRHR heat exchanger tube bundle. Also, the three-dimensional natural circulation in the IRWST was simulated numerically using the FLUENT code. The total heat transfer rate of the PRHR was predicted within 5% of the measured value. The fluid temperature distribution in the IRWST was also predicted well except that the elevation of the thermally stratified region was underpredicted. The calculated flow pattern in the IRWST suggests that the atypical IRWST geometry in the LSTF may have affected the PRHR heat transfer performance during the experiments only a little.

Evaluation on Driving Force of Natural Circulation in Downcomer for Passive Residual Heat Removal System in JAERI Passive Safety Reactor JPSR

The driving-force of the natural circulation in the residual heat removal (RHR) system for the JPSR (JAERI Passive Safety Reactor) is given as a gravity force of the density difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The amount of density difference and time to achieve the enough density difference for the RHR system change directly dependent on the thermal fluid flow pattern in downcomer of annulus flow pass. The purposes of the present study are to investigate the possibilities of the followings by evaluating the three-dimensional thermal fluid flow in downcomer by numerical analysis using the STREAM code; 1) promotion of making the flow pattern uniform in downcomer by installing a baffle, 2) achievement of an enough driving-force of the natural circulation, 3) validity of one-point assumption, that is, complete mixing down-flow assumption for the three-dimensional thermal fluid flow in downcomer to evaluate the function of the passive RHR system.The following conclusions were obtained: (1) The effect of baffle on the thermal fluid flow and driving-force is little, (2) The driving-force required for natural circulation cooling can be obtained in wide range of inlet velocity even if the flow is multi-dimensional, (3) Both in initial transient stage and in steady-state, the one-point assumption can be applied to evaluate the driving-force of natural circulation in the passive RHR system.