Thermal-hydraulic Test Facility Research Articles

TRACE assessment of density wave instability onset with void reactivity feedback under natural circulation

Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.

Integral-Effect Test for Loss of Residual Heat Removal System During Mid-Loop Operation in ATLAS Facility

A test called B4.2 in the OECD-ATLAS2 project was performed to simulate loss of the residual heat removal system (RHRS) during mid-loop operation (MLO) using a thermal-hydraulic (T-H) integral-effect test facility: the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). The main purpose of this test was to investigate a T-H transient in the reactor coolant system (RCS) during loss of the RHRS and to evaluate the effectiveness of reflux condensation and the capability of a safety injection tank (SIT) on shutdown coolability. The initial and boundary conditions for the B4.2 test were appropriately determined according to a state of MLO corresponding to 65 h after reactor trip in the Advanced Power Reactor 1400 MW(electric) (APR1400). During the loss of RHRS accident transient simulation, major T-H parameters such as system pressures, temperatures, and collapsed water levels in the RCS were measured, and unique T-H phenomena such as reflux/cocurrent condensations, off-take, countercurrent flow, and countercurrent flow limitation were investigated. In this paper, the overall T-H behavior in the RCS during a simulated loss of the RHRS with SITs is highlighted to provide a better understanding of T-H phenomena regarding coolability with reflux condensation.

Passive Autocatalytic Recombiner Perfomance Assessment Using COCOSYS

The progression of a severe accident in nuclear reactors is composed by a diversity of phenomena that areBeyond Design Basis, such as Direct Containment Heating (DCH), Molten Corium Concrete Interaction(MCCI), hydrogen detonation, and others. Currently, there are several devices and systems that allowmitigating the progression of this events, avoiding the failure of the physical barriers between the nuclearpower plant and the environmental. In this context, the present work aims to reproduce the HR-14experiment carried out at the Thermal-hydraulic, Hydrogen, Aerosols and Iodine (THAI) test facility throughthe Passive Autocatalytic Recombiners (PAR) performance assessment with the COCOSYS code. The analysisof the convergence of the results was performed using the Fast Fourier Transform Based Method (FFTBM)and showed that the results had sufficient accuracy with the experimental data.

Iterative computed tomography reconstruction of void fraction in a fuel bundle on unstructured mesh

The experimental investigation of internal two-phase flow structures in high pressure/high temperature systems is technologically challenging and limited to a few instrumentation techniques. For complex fuel bundle geometries, the use of Computed Tomography (CT) allows for relatively high spatial resolution (0.2 to 1 mm) and non-intrusive measurement of time-averaged void fraction. One of the main issues using standard structured reconstruction mesh is due to the loss of accuracy near the test section boundaries, for instance within the pixels overlapping both the two-phase coolant and the heated rods. In this work, a void fraction reconstruction methodology based on unstructured mesh is developed to accurately bound relevant attenuation regions within the considered test section. This method eliminates any overlapping reconstruction pixels and allows straightforward applications of knowledge-based, low noise, algebraic iterative reconstruction techniques (ART, SIRT, etc.). Within the same consistent framework, CT scan forward projections can be simulated for various selected cases (single-phase, arbitrary-two-phase void fraction distributions, etc.) and reconstructed to assess the performance of both the CT scan setup (e.g., projection data sampling and number of projections) and the reconstruction technique. The method is applied to the void fraction reconstruction of gamma CT data in a 5 × 5 fuel bundle geometry measured at the Westinghouse FRIGG thermal–hydraulic test facility operating under prototypical conditions of Boiling Water Reactor. The unstructured mesh is designed to match the features of the test section (isolation, pressure vessel, water gap, fuel channel, heated rods and coolant) with high-resolution in the domain of interest (coolant) and very low-resolution in all other domains with known attenuation coefficients. As compared to previous void fraction reconstruction attempts for this data set, the new reconstruction methodology allows significant improvements in void fraction resolution and noise reduction.

Deep-learning-based system-scale diagnosis of a nuclear power plant with multiple infrared cameras

Comprehensive condition monitoring of large industry systems such as nuclear power plants (NPPs) is essential for safety and maintenance. In this study, we developed novel system-scale diagnostic technology based on deep-learning and IR thermography that can efficiently and cost-effectively classify system conditions using compact Raspberry Pi and IR sensors. This diagnostic technology can identify the presence of an abnormality or accident in whole system, and when an accident occurs, the type of accident and the location of the abnormality can be identified in real-time. For technology development, the experiment for the thermal image measurement and performance validation of major components at each accident condition of NPPs was conducted using a thermal-hydraulic integral effect test facility with compact infrared sensor modules. These thermal images were used for training of deep-learning model, convolutional neural networks (CNN), which is effective for image processing. As a result, a proposed novel diagnostic was developed that can perform diagnosis of components, whole system and accident classification using thermal images. The optimal model was derived based on the modern CNN model and performed prompt and accurate condition monitoring of component and whole system diagnosis, and accident classification. This diagnostic technology is expected to be applied to comprehensive condition monitoring of nuclear power plants for safety.

Similarity evaluation of the pump simulation loop in STELLA-2 for conservation of mechanical sodium pump characteristics

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

Design of large-scale sodium thermal-hydraulic integral effect test facility, STELLA-2

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multi-dimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

TOPFLOW – a new multipurpose thermalhydraulic test facility for the investigation of steady state and transient two phase flow phenomena

Abstract The Forschungszentrum Rossendorf (FZR) e. V is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for Transient Two Phase Flow Test Facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of Computational Fluid Dynamic (CFD) Codes.

First experiments on visualisation of two-phase high pressure and temperature flows using an ultrasonic mesh sensor

Abstract A novel device for fast visualisation of gas-liquid two-phase flows was developed and tested during loss-off-coolant accident simulations at the thermal hydraulic test facility PSBVVER, a 1 : 300 integral model of the VVER-1000. The device is an ultrasonic mesh sensor. It consists of a metallic frame where transmitter and receiver wave-guides are fixed, that form two grids inside the measurement cross section. Ultrasonic pulses are transmitted into the fluid by the 8 wave-guides of the first plane. A second plane of another 8 wave-guides, that cross the ones of the first plane under an angle of 90 deg, serves as receives. The measurement is based on the acoustic conductivity of the two-phase mixture at the locations where the wave-guides cross. The sampling frequency is 250 frames per second. This allows both void fraction measurements and a fast flow visualisation. The sensor is applicable to high pressures and temperatures. All parts and surfaces that are in contact with the fluid are manufactured from stainless steel. During the tests at PSB-VVER the flow pattern in the hot leg of the primary circuit model was visualised for the first time.

The H2020 McSAFER Project: Main Goals, Technical Work Program, and Status

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.

Experimental study on capability and start-up characteristic of advanced secondary passive residual heat removal system

Safety systems play a key role in the nuclear industry. In the past decades, various safety systems have been designed and built to enhance safety level in nuclear power plants. The advanced secondary passive residual heat removal system (ASP) is one of several passive safety systems designed to ensure the safety of the reactor core. Experimental study on transient and steady-state characteristics as well as relevant influence factors of ASP system were conducted by using the high temperature high pressure thermal-hydraulic test facility. The results showed that the C-type heat exchanger could effectively remove the residual heat, and get stronger with the increase of loop pressure. The results also showed natural circulation could be set up with good stability, and get little influence from movement of isolating valves, relief valve as well as water volume in steam generator. Besides, to certificate the reasonability and reliability of the ASP system directly, an integral test was conducted.

Thermal-hydraulic test facility for nuclear reactor containment: Engineering design methodology and benchmarking

The paper describes detailed thermal and mechanical design procedure of a CONtrolled FLOw (CONFLO) facility and a complimentary Thermal HYdraulic test facility for CONtainment (THYCON), to investigate the post-severe accident scenarios in Nuclear Power Plant (NPP) containments, involving condensation of steam in the presence of non-condensable gases (NCGs), such as air and hydrogen/helium. The former setup allows the study of flow condensation of steam on flat surfaces kept inside a rectangular flow section of 250 mm × 200 mm, under different inlet conditions of mixture concentration, pressure and temperature. The latter is a large cylindrical chamber of total height = 3.6 m and diameter = 0.96 m (vessel volume of 2.7 m3), mimicking a containment, wherein steam condensation and mixing transport of non-condensable gases can be experimentally simulated, as per conditions close to accident scenarios. Parameters of interest are local and average condensation heat transfer coefficient, NCG concentrations, flow Reynolds number, volumetric Richardson number, operating gas/vapour pressure, temperature and inclination angle of condensing surfaces. After due benchmarking of both the setups, the generated experimental data is being used for developing the design equation(s) to model steam condensation in containment environments and for suggesting locations for condensers and passive autocatalytic recombiners essential for containment safety. Ongoing experiments are also providing data for the validation of computer codes. The thermal-hydraulic experiments suggest that the condensation heat transfer inside NPP containments is primarily governed by several interlinked phenomena, such as stratification, mixing, turbulence and condensation.

Major Outcomes through Recent ROSA/LSTF Experiments and Future Plans

Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Recent research activities concerned with the OECD/NEA international joint projects included experimental investigation via the ROSA and ROSA-2 Projects, and counterpart testing with thermal-hydraulic integral test facilities under collaboration of the PKL-2, PKL-3, ATLAS, and ATLAS-2 Projects. Major results of the related integral effect tests (IETs) with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Future separate effect test using the LSTF is planned to simulate loss of emergency core cooling system (ECCS) recirculation functions in a large-break loss-of-coolant accident (LOCA). Key results of the recent IETs utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break LOCA with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from ECCS into cold legs. Also, main outcomes of the LSTF IETs were indicated for wide spectrum LOCA with core uncovery and anticipated transient without scram following small-break LOCA under totally failed high-pressure injection system.

Analysis of the results of thermophysical experiments with multi-rod bundles. Elaboration of the SCP thermalhydraulic test facility design

The present report is devoted to the analysis of the thermophysical properties of supercritical water in the cores of supercritical water-cooled reactors (SCWR) and results of the experiments with multi-rod bundles cooled with supercritical fluid (mainly – water) flows carried out in different countries. The most important results and specific heat transfer regimes observed in the experiments were summarized. On the basis of the experimental data analysis the design of a thermophysical test facility for full-scale experiments for the determination of the heat and mass exchange regimes and supercritical water hydraulic parameters in multi-rod bundles was elaborated. The need for the experimental data for the proper description of the said processes was stated, and the importance of obtaining them was underlined. The list of the controlled parameters and experiments was created.

Study on convective heat transfer correlations for sodium-to-sodium heat exchanger based on STELLA-1 experimental results

Multiple studies have been conducted on the convective heat transfer of a liquid metal flow to present the correlations between the Nusselt and Péclet numbers. However, those correlations are considerably deviated from the experimental data obtained from steady-state tests of a sodium-to-sodium heat exchanger using the large-scale sodium thermal-hydraulic test facility STELLA-1. In this work, a new convective heat transfer correlation acquired through curve fitting with the STELLA-1 experimental results is proposed for a Péclet number range from 68 to 524. The newly proposed correlation was used in the SHXSA design code developed for designing a sodium-to-sodium heat exchanger by the Korea Atomic Energy Research Institute. The adequacy of the correlation was assessed by comparing the calculated results of the SHXSA code with the test results from STELLA-1. By using the proposed correlation, the average absolute deviations of the shell-side outlet temperature, tube-side outlet temperature, and heat transfer rate improved from 0.70% to 0.51%, 0.37% to 0.17%, and 1.61% to 0.82%, respectively. In the assessment with the experimental results of the JOYO sodium-to-sodium heat exchanger, the average absolute deviations of the shell-side outlet temperature, tube-side outlet temperature, and heat transfer rate were determined to be 0.27%, 0.54%, and 2.05%, respectively. The calculated results using the proposed correlation showed a satisfactory agreement with the experimental data from the STELLA-1 and JOYO facilities, and thus the proposed correlation is expected to be practically employed for designing a sodium-to-sodium heat exchanger.

Flow morphology and heat transfer analysis during high-pressure steam condensation in an inclined tube part I: Experimental investigations

In this paper, experimental investigations on the flow morphology and heat transfer in a single steam condenser tube are presented, which were performed at the thermal hydraulic test facility COSMEA (COndensation test rig for flow Morphology and hEAt transfer studies). This facility was setup to study the interrelation of condensation heat transfer with two-phase flow in a single condenser tube that is cooled by forced convection. Studies were performed for elevated pressures up to 65 bar at saturation conditions and for inlet steam mass flow of up to 1 kg/s and different inlet steam qualities. The wall heat flux is measured with distributed heat flux probe and global condensation rates were obtained from integral heat and mass balances. As a unique feature the cross-sectional phase distribution was studied via X-ray computed tomography. The data is going to be used for the validation of numerical simulations with 1D ATHLET and 3D CFD codes as presented in the second part of this paper.

Experimental and theoretical investigations on reverse flow characteristics in vertically inverted U-tube steam generator under transient condition

Under natural circulation conditions, the primary circuit of a steam generator (SG) has a tendency to occur flow instabilities. These instabilities due to reverse flow in some vertically inverted U-tubes will increase the flow resistance coefficient and reduce the natural circulation ability in the primary system. To address the lack of existing research on transient reversed flow, in the present work, the reverse flow characteristics in the U-tubes of vertically inverted U-tube steam generator (UTSG) are investigated experimentally and theoretically. By gradually rising the inlet plenum mass flow of steam generator simulator at a constant fluid temperature, the reverse flow characteristics are observed in a thermal hydraulic test facility. Based on the experiment, the flow and heat transfer model in U-tubes are established to study the effects of change rate of mass flow and flow distribution on the reverse flow inside the UTSG. The experimental and theoretical results demonstrate that during the transient process, some U-tubes get reverse flow with the continuous decrease of mass flow and the reverse flow U-tubes also have a significant impact on other U-tubes, which results in a maximum temperature drop of about 35%. With the decrease of mass flow in U-tubes, the critical mass flow of the UTSG will increase and the reverse flow will occur more easily during the transient process. The results can provide a basis for avoiding the occurrence of reverse flow in actual SGs and improving the ability of natural circulation.

Recent achievement and future prospects of the ATLAS program

As a critical thermal-hydraulic test facility, the ATLAS facility has been widely utilized since its first commission test in 2006 to resolve urgent safety concerns raised by nuclear industries and regulatory body, to verify the performance of new safety systems, to validate safety analysis codes, and to provide a cooperation network for developing safety analysis capabilities. In the first phase of the ATLAS program, major R&D efforts were put into producing code validation data to support the standard design approval of the APR1400 when design basis accident (DBA) scenarios were the main interest of the program. In the following phase, the ATLAS program’s focus shifted to beyond DBA scenarios, and the OECD-ATLAS international joint project started reflecting the safety concerns of many NEA countries following the Fukushima Daiichi disaster in 2011. In this paper, the current status of the ATLAS program, including the OECD-ATLAS Phase-1 and 2 projects, are briefly described together with their major outcomes. The next R&D program based on the ATLAS platform is also briefly discussed.

Elevated temperature design and integrity evaluation of a large-scale sodium test facility, STELLA-2

The construction of a large-scale sodium thermal-hydraulic test facility known as STELLA-2 for use in integral effect tests of a sodium-cooled fast reactor (SFR) at the Korea Atomic Energy Research Institute (KAERI) is scheduled to be completed by the end of 2019. As part of this, elevated temperature design (ETD) evaluations were conducted according to the ETD code of RCC-MRx for all components operating at high temperature in creep ranges whereas for Grade 91 steel components of the intermediate heat exchanger (IHX) and the decay heat exchanger (DHX), ETD evaluations were additionally conducted according to ASME Section III Division 5 Subsection HB as well. Except IHX and DHX, all other components in STELLA-2 are of 316L stainless steel (SS). Since high-temperature material properties for 316L SS are available only in RCC-MRx, mainly RCC-MRx was applied. Comparisons were made between the two ETD codes in terms of chemical compositions, and material properties for austenitic 316 SS and Grade 91 steel. The structural integrity of the sodium test facilities was confirmed based on the design evaluations according to the ETD codes. In addition, design evaluation results from RCC-MRx and ASME Subsection HB were compared for IHX and conservatisms were quantified. Technical aspects that mechanical designers should note associated with application of elevated temperature design rules to actual high temperature components were highlighted.

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.