The stable operating conditions for an open loop passive containment heat removal system were identified through testing conducted with the PASI test facility, a half-height wall condenser model at LUT University, Finland. Previous tests have shown that open loop systems tend to operate in a quasi-steady oscillatory mode characterized by geysering and flashing. The cessation of flow oscillations depends on the sparger structure. When flooding of the riser pipeline is prevented, the oscillation fade-out and steady two-phase natural circulation is reached quickly after the system reaches saturation conditions. Conversely, if flooding is allowed, the oscillations disappear only at heating power large enough to meet the countercurrent flow limitation (CCFL) criterion in the riser. The impact of gravity head on the system behavior was also examined. The amplitude of two-phase flow oscillations decreased along the lowering of the pool water level. When the water level decreased below the pressure balancing hole, the flow behavior changed since the riser flooding ended. Additionally, the riser boil-out was tested. The results show that the open-loop natural circulation system can effectively remove heat as long as there is water inventory inside the loop, even if the pool is empty of water. The containment pressure rises only when boiling initiates in the heat exchanger.

Thermal hydraulic experiments with the modular integral test facility, MOTEL, were performed as a part of the European McSAFER research project. The facility models an integral pressure water small modular reactor (SMR) with a helical steam generator and a core with separate heater rod groups, in which power can be individually controlled. Different asymmetric and ring-shaped radial core power distributions were imposed in the experiments to provoke cross flows in the buoyancy-driven coolant flow. The purpose of the experiments was to produce new SMR-relevant data for the validation of computational fluid dynamic (CFD) and thermal-hydraulic subchannel codes. The experimental measurements revealed cross flow mixing effects, mainly in the top part of the core. Obtaining visible differences in the fluid temperature measurements between different heater regions required significant power gradients between the regions. CFD simulations were performed using ANSYS CFX with a detailed model comprising the whole primary side of the facility, and additional investigations were conducted with a stand-alone model of the heat exchanger. Good agreement with the measurements was obtained with the CFD simulations, which also revealed further details of the core flow characteristics in an asymmetric heating case. Furthermore, simulations with the subchannel codes, CTF and VIPRE-01, were performed. The simulations with CTF highlighted the code's capability to handle flow rates typical to natural circulation driven SMRs, as the results agreed well with the experiments and were able to predict the correct axial temperature profiles in the different regions of the core. VIPRE-01 solution stability was found to be highly sensitive to the flow rate, the power level, and the axial nodalization. Simulations with VIPRE-01 ended unsuccessfully due to convergence issues, and it was concluded that the conditions of the experiments are beyond the current capabilities of the code.

Several nuclear reactor designs rely on passive containment cooling systems. The so-called containment wall condenser relies on natural circulation loops to extract heat from high-temperature steam in the containment to a water tank at ambient pressure. In such passive systems, phase changes can happen and cause flow instabilities in the cooling loop. The flashing-induced instability occurs when the heated fluid in the riser suddenly vaporizes due to a hydrostatic pressure decrease. This instability causes periodic flow peaks, which are of major concern but whose characteristics have not been studied quantitatively.This paper presents two analytical models that predict the flashing frequency and a maximum flow amplitude from geometry and basic operating parameters such as power level and reservoir temperature. The expressions are derived from a physical analysis and do not involve any calibration constants. The flashing frequency appears to be driven by the power level, the inlet temperature and the riser pipe geometry. For the amplitude, the maximum flow rate can be expressed in a Froude number that depends only on the total pressure losses. These models are validated against PASI experiments and system-scale simulations with the CATHARE 3 code, both performed as part of the European Commission funded PASTELS project. Additional data from numerous experimental studies in the literature are used to extend the validity range of the frequency model.Successfully validated against experimental data and additional simulations, these models provide an explicit relationship between oscillations characteristics and design parameters, making them valuable tools for nuclear engineers.

Small modular reactors (SMRs) are under extensive development globally. Some SMR concepts have design features that are rare in traditional pressurized water reactors (PWRs). One such feature is a helical coil steam generator which differs from traditional horizontal and vertical inverted U-tube steam generators in several ways. The helical coil steam generator is a once-through design where the primary side flow runs in the shell side and the secondary side flow runs inside the tubes and can generate superheated steam. Boiling instabilities in helically coiled tubes are a crucial research question due to their potentially negative impact on steady plant operation. The MOTEL (MOdular TEst Loop) facility at LUT University is a model of an integral SMR with a helical coil steam generator representing an integral pressurized water reactor. A large variation of core power and feedwater flow values were tested to map MOTEL operating conditions in which the steam production in the helical coil steam generator is stable.

Post-accident heat removal from nuclear containments is an important safety function in nuclear reactors. Passive systems have been proposed to increase the reliability of the function, but owing to small driving forces, concerns exist on the performance and testability of such systems. The basic functioning of an open loop gravity-driven containment passive heat removal system was studied with the PASI test facility, a half-height wall condenser model at LUT University, Finland. The quasi-steady-state behavior of the natural circulation flow was observed in both single-phase and two-phase flow modes. A wide range of heating powers was tested, and robust heat transfer performance was observed throughout. Two-phase flow was found intermittent with flashing-induced sharp flow peaks separated by slow steady flow. Piping dynamic loads due to flow peaks were found to be small, although one weak indication of a possible flashing instability induced water hammer (FIIWH) was observed. In addition to accident operating conditions, moderate conditions that could be reproduced during an actual power plant outage were also tested, showing that periodic testing of these passive systems is practicable.

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today’s nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020–February 2024), funded by the European Commission “Euratom H2020” programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome’s PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New “system/CFD” coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

The Lappeenranta-Lahti University of Technology LUT has been the custodian of Finnish nuclear reactor safety research, mainly thermal-hydraulic and regulations-related, from the dawn of the nuclear era in Finland in the mid-1970 s. This paper provides a concise summary of research carried out at LUT University since 2000, presenting work done on large Light-Water Reactors, Small Modular Reactors, gas-cooled reactors, advanced measurement techniques, and nuclear regulations. Independent of the size or shape of the application, one common theme repeats through the decades of effort: a fully representative picture of Nature's behaviour can only be obtained by a balanced combination of experiments, scaling analyses and analytical modelling. Performance shortcomings have been forestalled and resolution of emerging safety issues supported by experiments at LUT University.

This paper describes the main objectives, technical content, and status of the H2020 project entitled “High-performance advanced methods and experimental investigations for the safety evaluation of generic Small Modular Reactors (McSAFER)”. The main pillars of this project are the combination of safety-relevant thermal hydraulic experiments and numerical simulations of different approaches for safety evaluations of light water-cooled Small Modular Reactors (SMR). It describes the goals, the consortium, and the involved thermal hydraulic test facilities, e.g., the COSMOS-H (KIT), HWAT (KTH), and MOTEL (LUT), including the experimental programs. It also outlines the different safety assessment methodologies applied to four different SMR-designs, namely the CAREM (CNEA), SMART (KAERI), F-SMR (CEA), and NuScale. These methodologies are multiscale thermal hydraulics, conventional, low order, and high fidelity neutron physical methods used to demonstrate the inherent safety features of SMR-core designs under postulated design-basis-accident conditions. Finally, the status of the investigations is shortly discussed followed by the dissemination activities and an outlook.

The PASI test facility has been designed and constructed at LUT University for the thermal-hydraulic studies of an open-loop passive containment heat removal system for nuclear reactors. The PASI facility consists of a pressure vessel that simulates containment conditions and a loop comprising of a heat exchanger, interconnecting pipelines, and a water reservoir. The research work with the PASI test facility aims to study the system performance characteristics and detect issues disturbing or preventing the designed operation of an open naturally circulating system. The PASI facility offers flexible structural features, traditional measurements, and the possibility to utilize advanced measurement techniques. It provides a new perspective for phenomena studies and experimental data useful for code validation. The PASI facility was taken into operation in 2018 with the characterizing experiments. The facility worked as expected, exhibiting two typical operational modes: stable and oscillatory. This paper presents the background for the chosen design of the PASI facility, summarizing the scaling and practical constraints considered in the design process. The details about the construction of the PASI facility with characteristic parameters and figures are presented. In addition, the characterizing experiments and the future research activities utilizing the PASI facility are discussed.

Various Wire-Mesh Sensor (WMS) designs have been developed and applied for the single-and two-phase flow studies over the recent years. The design and the construction of the WMS can significantly vary depending on the purpose of the research. Typically, the WMSs installed in pipes and ducts have been located across the flow, to obtain instantaneous cross-sections of the flow area. The traditional approach would not be ideal if the scope of the study is to measure the development or the dynamics of the flow in the axial direction. Hence, the axially-aligned conductivity Wire-Mesh Sensor (named AXE) was designed and constructed to be located in the centerline of the 50 mm pipe in the current study. The paper presents briefly the design of the axial sensor and the HIPE test loop that was used in the experiments. The sensor enables the study of the two-phase flow in high resolution (10,000 frames/s, 3 mm x 3 mm resolution). In this paper, the main emphasis is on the study of the different methods that can be applied for the estimation of the velocity fields from the axial wire-mesh sensor data. Two methods have been tested: traditional Time-of-Flight (ToF) estimation and optical flow method. The Time-of-Flight estimation relies on the cross-correlation of the time series of the void fraction values from the different locations. The velocities can be calculated from the Time-of-Flight estimates and the distances between the particular locations. The optical flow methods can be used to evaluate the motion of the objects between the two sequential images. The estimation of the velocity fields from the axial sensor data is analogous to the calculation of the velocity fields from the Particle Image Velocimetry (PIV) images. Now, the same methods are applied to process the axial WMS data. The paper discusses the pros and cons of the different approaches and gives some general recommendations on how and when they should be applied for the WMS data. The velocity fields from the axial sensor experiments are compared to the experiments conducted with the two cross-sectional WMS (32 x 32 sensors). One aim of the studies on axial sensor is to utilize the sensor in research dealing with swirling flows generated by different mechanisms. (C) 2017 Elsevier B.V. All rights reserved.