The Experimental Thermal Hydraulics for Analysis, Research and Innovations in Nuclear Safety (called ETHARINUS) project developed in the frame of the OECD Nuclear Energy Agency (NEA), serves the objectives of investigating complex thermal-hydraulics phenomena, issues related to the performance of passive heat removal systems and Design Extension Conditions (DEC) scenarios. The simulation of such events is of high importance to ensure relevant understanding of key thermal-hydraulic phenomena, to perform adequate safety analysis, to assess the efficiency of the adopted accident management procedures and to optimize operator training.ETHARINUS project is highly relevant for the improvement and validation of thermal-hydraulic (TH) system codes and their use, to maintain a high level of competence and expertise in the field of system thermal hydraulics. It is also a way to gather the actors within the area (safety authorities, operators, experimental facilities operators, universities, R&D institutes, Technical and Scientific Support Organization (TSO), etc.) to develop knowledge and common culture about key safety issues, for the operating fleet and for innovative designs.The objective of this paper is to describe the capabilities of the thermal-hydraulic research facilities employed for the activities, to outline the associated tests programme, and to highlight their significance regarding key nuclear safety issues. Such tests programme is conducted in an international context, to share costs and expertise, and to promote quicker and deeper international consensus on safety issues. Recommendations are finally proposed regarding how to address the loss of critical research infrastructure (i.e. facilities, capabilities and expertise).

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.

The presence of non-condensable gases in the reactor cooling system can significantly influence the operation of several safety systems within a nuclear power plant. The pressurized nitrogen volume at the top of the accumulator tank is the driving force for injecting accumulator water. In certain nuclear power plants, the release of gaseous nitrogen to the primary side is inhibited by an automatic closure of the accumulator injection line at the end of the discharge. However, if this automatic closure system fails, nitrogen will inadvertently flow into the reactor cooling system once the accumulators have been depleted. Furthermore, it is important also to note that the water within the accumulator is saturated with dissolved nitrogen, resulting in the injection of some nitrogen into the primary system alongside every accumulator discharge. The impact of nitrogen on core cooling during loss of coolant accidents (LOCAs) has been investigated experimentally using the PWR PACTEL facility. The main observations were that when a break occurs in the hot leg, the injection of nitrogen from an accumulator can effectively prevent depressurization of the primary side. Consequently, the core can experience a heat-up at primary pressure somewhat above the typical low-pressure safety injection (LPSI) shut-off head. The decoupling of primary and secondary side pressures depends on the amount of nitrogen released from the accumulator, how much of it accumulates into the U-shaped steam generator heat exchange tubes, decreasing the condensation heat transfer, and the number of steam generators participating in the secondary side depressurization. Furthermore, the size of the break significantly affects the volume of nitrogen escaping through the break, which in turn influences the nitrogen levels within the system. Specifically, larger breaks permit a greater flow of nitrogen, thereby reducing the likelihood of disrupting heat transfer between the primary and secondary sides while also mitigating the depressurization of the primary side.

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.

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.

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.

The hydroaccumulators in pressurized water reactors can inject nitrogen into the reactor system. In the primary system, nitrogen affects core cooling and accident management, both adversely and beneficially. The PWR PACTEL experiment NCG-13 have shown that during a hot leg SB LOCA, nitrogen in the primary side can block the primary to secondary heat transfer and thereby prevent the primary depressurization to the point needed for the long-term accident management. This paper presents the APROS and TRACE calculations of the PWR PACTEL NCG-13 experiment. Both codes calculate the transient progression and the timing of the main events satisfactorily, once the suitable options are selected and adjustments made. There is, however, one big discrepancy between the code simulations and the experiment: in the simulations, much more nitrogen is needed to get qualitatively the same behaviour. The difference is a factor of 2.5 by mass for stopping the depressurization and 4-6 to cause a core heat-up. This is of concern regarding the confidence in the codes, as the simulations underestimate the adverse effect of nitrogen on the core coolability.

Noncondensable gases, if present in the reactor cooling system, affect the coolability of the nuclear reactor core. In Loss-Of-Coolant Accidents (LOCA), nitrogen from hydroaccumulators will enter the reactor systems, and can temporarily alter the water level in the core by a piston effect. The piston effect on the downcomer side can increase water level in core and improve core cooling (Damerell et al., 1993). The effect of nitrogen on core cooling in LOCA situations was studied experimentally in the PWR PACTEL facility. The main goal of this testing was to independently verify whether the claimed positive effect of nitrogen on the core cooling can be reproduced and to generate data for the development and validation of thermal-hydraulic system codes. Four experiments with an accumulator injection to a cold leg were performed with the PWR PACTEL facility. In two of the experiments, the break was in the other cold leg and in the other two in the hot leg. The cold leg injection experiments confirmed that nitrogen injection had a small positive impact on the core cooling, but not by shortly redistributing water masses in the vessel as expected. The presence of nitrogen reduced the break water flow rate, delaying the core level depletion and thus postponing core heatup by few minutes. In contrast, with the hot leg break, the injection of the accumulator nitrogen into the primary side had a negative impact on the core cooling. The nitrogen accumulated in the steam generator tubes, blocking the primary side depressurization and causing a core heatup.

A computational procedure suitable for modeling both vertical and horizontal steam generators is presented. The procedure consists of a one-dimensional model for the whole steam generator and a CFD model for the secondary side. The three-dimensional two-phase CFD simulation of the secondary side utilizes the porous medium approximation and the mixture k - epsilon turbulence model. One-way coupling between the models is used, in which the surface temperatures of the heat transfer tubes are considered as a boundary condition for the CFD simulation. A stationary state and two pressure transients, a pressure rise and a pressure reduction, for the PWR PACTEL steam generator are calculated and the internal circulation on the secondary side is analyzed. PWR PACTEL is a test facility, located at Lappeenranta University of technology, in Finland, which was constructed for research activities associated with the European Pressurized Reactor (EPR). (C) 2016 Elsevier B.V. All rights reserved.

Water seal formation in the loop seal in pressurized water reactors can occur during a small or intermediate break loss-of-coolant accident, causing temporary fuel overheating. Quantification of the accuracy of overheating prediction is of interest in the best-estimate safety analyses, even though the peak cladding temperatures due to the water seal formation in the loop seal seldom approach acceptance criteria as such. The aim of this study was to test and evaluate the accuracy with which the thermal-hydraulic system code nodalizations of the PWR PACTEL predict loop seal clearing in a small break loss-of-coolant-accident test performed with the PWR PACTEL facility. PWR PACTEL is a thermal-hydraulic test facility with two loops and vertical inverted U-tube steam generators. Post-test simulations were performed with the TRACE and AFROS system codes. In the post-test simulations, the main events of the transient such as the decrease in the core water level, depressurization of the primary circuit, and the behavior of the water seal formation and clearing in the loop seal were predicted satisfactorily by both codes. However, discrepancies with the experiment results were observed in the analyses with both codes, for example the core temperature excursions were halted too early and the peak temperature predictions were too low. The core water level increase caused by loop seal clearing was overestimated with both codes, and the pressure and temperature were overestimated on the secondary side of the steam generators. Loop Seal 2 was evidently cleared out while Loop Seal 1 remained closed, just like in the experiment. It was noticed that, which of the loop seals clears in the simulations, was sensitive on the nodalization structure and initial conditions. The simulations indicate that the PWR PACTEL nodalizations are capable of simulating loop seal clearing quantitatively well. A quantitative error in the non-conservative direction was observed in peak cladding temperatures in both codes. (C) 2015 Elsevier Ltd. All rights reserved.

The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues.The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released.According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code.The benchmark offered a unique opportunity to test the best practices and solutions in modeling and analyzing tasks as well as a possibility to increase knowledge about the interpretation of test results. The benchmark exercise served as a practical and rewarding forum to discuss the needs, problems and possibilities in the analysis and in producing useful data with an experiment facility. The workshops provided an advantageous site for interaction of the code users and the experimenters. (C) 2013 Elsevier Ltd. All rights reserved.

ABSTRACT The PWR PACTEL test facility has recently been designed to support the safety studies of EPR type nuclear reactor thermal-hydraulics. The facility is located at the Lappeenranta University of Technology (LUT) in Finland. It is essentially important to understand the system behavior under natural circulation conditions during a Loss of Coolant Accident (LOCA). With this objective in mind, an international benchmark transient was conducted at the LUT in 2010-2011. The SBL-50 test was a SB-LOCA with a 1 mm break in the cold leg. The continuous inventory loss led to core dry-out. This project gave unique opportunities for several organizations to build and validate their models for the PWR PACTEL, as well as to simulate the transient by using various computer codes. Chalmers University of Technology participated with RELAP5/Mod3.3 calculations both in the pre-test and post-test phases of the project. The pre-test simulation included a simplified steam generator model, with the description of the heat exchange by a single characteristic U-tube. This coarse nodalization resulted in reasonably good agreement with the measured data. As the test data became known in the post-test, minor modifications contributed to achievement of better results. The changes were related to the upper plenum nodalization and the critical discharge flow parameters at the break assembly. Even if the post-test model provided better agreement for most of the parameters, it still had difficulties to predict the temperatures in the longest tubes in the steam generators (SGs). A careful examination of the measured data indicated that discrepancies might originate from a flow reversal in the SG primary side. Thus, an advanced SG model was prepared with application of a multi-channel system. Altogether 51 heat exchanger tubes are arranged into 5 bundles in the PWR PACTEL SG, according to 5 different lengths. These bundles were individually modeled in the refined input. The most recent multi-channel SG model has not only confirmed the presence of reverse flow and internal circulation in the SG primary, but it has also contributed to a much better prediction of the fluid temperature distribution.

This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work) of the EC project "Improved Accident Management of VVER nuclear power plants" (IMPAM-VVER). The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model), represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

Regardless of the large number of thermal hydraulic experiments conducted with many different facilities, the need for good quality data from integral test facilities has not yet reached saturation. The parallel channel test loop (PACTEL) facility is one of the largest facilities of its kind. It was originally designed to model the thermal-hydraulic behavior of VVER-440 type pressurized water reactors (PWRs) currently used in Finland. Nevertheless, the PACTEL facility has served also in many other purposes than for VVERs only. The facility has been modified on a case-by-case basis according to the needs in configuration and positioning of auxiliary equipment. The newest plan is to modify PACTEL to be the PWR PACTEL, a facility with vertical steam generators for EPR applications topical in Finland. This paper describes the versatile use of the PACTEL facility for a large spectrum of thermal hydraulic research. The PACTEL facility is ideal for investigating planned recovery procedures during accidents and operational transients. For this purpose experimental series among others on small break loss-of-coolant accidents (SBLOCA), primary-to-secondary leakages (PRISE), and on anticipated transients without scram (ATWS) have been carried out. The PACTEL natural circulation experiment with stepwise coolant inventory reduction formed the basis for the OECD International Standard Problem (ISP-33). In addition, many other one-phase and two-phase natural circulation tests have been executed.

The PACTEL is a test facility designed to model the thermal-hydraulic behavioursof the Soviet-designed VVER-440 pressurized water reactors currently in use inFinland. These reactors have unique features that differ from other PWR designs.The PACTEL simulates the major components and systems of the reference PWR,making it possible to examine postulated small- and medium-break LOCA's andoperational transients.The PACTEL is a volume-scaled model (1:305). To ensure that gravitationalforces...