Thermal-hydraulic Phenomena Research Articles

A three-dimensional PWR LBLOCA simulation using the CUPID code

As the need for simulations of multi-dimensional behavior in the safety analysis is increasing, efforts have been devoted to developing a three-dimensional module for the conventional system analysis codes. After having developed a one-dimensional system analysis code, MARS, the Korea Atomic Energy Research Institute (KAERI) began to develop the CUPID code to address the need for a multi-dimensional analysis. Recently, the reactor vessel module, such as a package of interfacial heat and momentum transfer model, wall friction and wall-to-fluid heat transfer model, were implemented in the CUPID and named CUPID-RV module. The final goal of this study was an analysis of the Large Break Loss of Coolant Accident (LBLOCA) at a rod-by-rod level, and the study focused on the validation of LBLOCA thermal-hydraulics phenomena and the application to a three-dimensional LBLOCA simulation. In this paper, we introduce the validations using Edward pipe, UPTF, and FLECHT-SEASET experiments related to the LBLOCA blowdown, refill, and reflood phases and a three-dimensional LBLOCA simulation for the APR1400 PWR reactor pressure vessel using simple boundary conditions at an assembly-by-assembly level as an application.

Sizing of a tube inlet orifice of a once-through steam generator to suppress the parallel channel instability

Sizing the tube inlet orifice of a Once-Through Steam Generator (OTSG) is important to protect the integrity of the tubes from thermal cycling and vibration wear. In this study, a new sizing criterion is proposed for the tube inlet orifice to suppress the parallel channel instability in an OTSG. A perturbation method is used to capture the essential parts of the thermal-hydraulic phenomena of the parallel channel instability. The perturbation model of the heat transfer regime boundaries is identified as a missing part in existing models for sizing the OTSG tube inlet orifice. Limitations and deficiency of the existing models are identified and the reasons for the limitations are explained. The newly proposed model can be utilized to size the tube inlet orifice to suppress the parallel channel instability without excessive engineering margin.

Thermal hydraulic modelling of grating effect for application to 3-dimensional analysis of hydrogen behavior in NPP containment

For three-dimensional analysis of the thermal hydraulic phenomena associated with hydrogen behavior in a containment building, geometry modelling that reflects the three-dimensional features of the containment building is required. A lot of equipment and concrete structures are installed in the containment building, and gratings are also installed on a large floor area. Gratings in the containment can affect gas flow and heat transfer, so modelling of the grating is required for three-dimensional analysis of hydrogen behavior in the containment. As a method of modelling gratings, many computational mesh nodes are required to directly simulate gratings made of steel plates several millimeters thick, so a large amount of computation is required for accident analysis. Therefore, simple modelling of gratings is required to properly control the size of the three-dimensional mesh. In this study, a method of simulating porous media for gratings was adopted, and the model was extended to allow for consideration of heat transfer of the gratings. To verify the porosity-based grating model, analyses of the H2P1_10_2 experiment recently performed through the international joint project OECD/NEA HYMERES-2 were conducted. Through a comparison with the H2P1_10_2 experimental data, it was confirmed that the porous media-based model showed calculation results very similar to those of the direct simulation of the grating, and that the porous media-based model reduced the mesh size by 1/5 and the calculation time by 1/5 compared to those values in the direct simulation of the grating.

An investigation of transition processes from transient gas–liquid plug to slug flow in horizontal pipe: Experiment and Cost-based recurrence analysis

Transient gas–liquid two-phase flows are often encountered in nuclear industries. Investigating the dynamics of the transient gas–liquid flows is of significance for optimizing the reaction processes and predicting the thermal–hydraulic phenomena relevant to the performance of nuclear reactors. In this study, a multi-nozzle gas inlet was designed to mimic the transient gas–liquid flows in a horizontal pipe. A conductance wire-mesh sensor (WMS) was used to visualize the phase distribution at the pipe cross-section. Based on the WMS data, a cost-based recurrence analysis (CBRP) was developed to reveal the dynamics of transient gas–liquid flows. Different transition processes from plug flow to slug flow were observed in the experiment and explored by the determinism (DET) of the CBRP. Note that special flow structures, i.e., wave bridge and pseudo slug, were encountered during the flow pattern transition, and their generation mechanisms were uncovered. The processes of the flow pattern transition were also characterized by the slug frequency and the probability density function of the water holdup, which presented a good agreement with the CBRP results.

A Review of Models for the Sodium Boiling Phenomena in Sodium-Cooled Fast Reactor Subassemblies

Abstract The Euratom Horizon-2020 project ESFR-SMART aims to enhance the safety and performance of the European Sodium-cooled Fast Reactor (ESFR) considering safety objectives envisaged for Generation-IV reactors and the update of European and international safety frameworks, taking into account the Fukushima accident. Further, the project aims to support the development and validation of the computational tools for the situations to be considered in each defense-in-depth level in order to support the safety assessments using data produced in the project as well as selected legacy data. Within this activity, the focus is on the assessment and when needed further development of computer codes for the analysis of the sodium thermal-hydraulics phenomena in SFR subassemblies under operational and accidental conditions including sodium boiling and transitional convection from forced to natural/mixed convection. In support of this activity, a review is performed for the sodium boiling models used in the codes participating in benchmark activities within the project. The objective of the present paper is to summarize the result of the review, which encompasses both the phenomenological and mathematical models implemented in the codes. In particular, the review addresses the physical bases of sodium boiling models, the analytical models in connection with the numerical implementations in the codes and the geometry representations ranging from one-dimensional single channel and finite volume to full three-dimensional CFD and porous media representations. Finally the needs for further developments are discussed.

Safety evaluation of nuclear power plant based on multiscale analyses

A loss-of-coolant accident of nuclear power plant is simulated using the system analysis code. The analytical condition is based on the international standard problem 26 conducted with the integral effect test facility. The experimental analysis is also performed, and the result of plant analysis is compared with the experiment and the experimental analysis. The discharge coefficient of critical flow model is determined so as to obtain the agreement of reactor pressure between the experiment and the experimental analysis, and is used for the plant analysis. The thermal-hydraulic phenomena in the experiment such as core heat up are simulated well by the two analyses, while some problems in the experiment for simulating plant accidents are made clear. Parametric sensitivity analyses of plant accidents are performed as an example of reliable safety evaluation using the validated plant model.

Effects of inclination and flow velocity on steam condensation consisting of air on tube bundle external surfaces

Steam condensation with existence of air is an important thermal hydraulic phenomenon in reactor containment. Previous investigations mostly focused on the surfaces of plates or tube bundles which are in parallel with the gas flow direction, and discussions on heat transfer surfaces with various inclination angles were merely restricted to plates. In addition, there is limited information about the condensation promotion effect caused by enlarged gas flow velocity at different tube inclination angles. To investigate the coupling effects of tube inclination and flow velocity, numerical simulations were carried on a 3 × 3 tube bundle. The local thermohydraulic profiles and heat transfer mechanism of steam condensation were analyzed carefully. The results indicate that the increase of inclination angle and flow velocity can dramatically promote condensation heat transfer. At the velocity of 0.3 m/s, the condensation heat transfer enhanced 327.8% as the tube bundle orientation changes from vertical to horizontal. At the inclination angle of 0°, the condensation heat transfer increased 223.2% as flow velocity increased from 0.3 m/s to 3 m/s. The inhibition effect and stack effect in tube bundle varied by the tube inclination and flow velocity.

Study of the ablation consecutive to jet impingement on a meltable solid – Application to SFR core-catcher

In the scope of safety studies for future sodium cooled fast reactors (SFR), the use of discharge tubes and in-vessel core-catcher are foreseen for severe accident mitigation. During relocation of corium onto the core-catcher, complex thermal hydraulics phenomena can occur. This work focuses on the ablation of the core-catcher by the jet of corium. For that purpose, as experiments using corium and Na are difficult to achieve, we performed experiments using simulant materials: water jet on transparent ice. Transparent ice allows for ablation cavity real-time recording. Two main regimes were identified: a film flowing regime and a pool regime. We also studied the influence of temperature (30, 50, 70 °C) and velocity (1,2.5,5,7.5,10m/s) of the jet. Comparison with existing data have been done and finally, a scaling law in the first regime, for which the ablation rate is constant, has been found.

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

ASTEC code DBA analysis of a passive mitigation strategy on a generic IRIS SMR

Advanced small LWRs, are considered as one of the key design options for the development of a safer nuclear technology considering their inherent safety due to the adoption of passive mitigation strategies and lower nominal power. In the framework of deterministic safety analyses for Small Modular Reactors (SMRs), a model of a generic IRIS SMR was developed by using the severe accident code ASTEC (Study carried out with ASTEC V2, IRSN all rights reserved, [2019]). The ASTEC code thermal-hydraulics modules have been used for the reactor modelling, and the nodalization approach has been described in the present work. The objective of the paper is to analyze the code capability and the consequent applicability to model an integral-type reactor and to simulate the complex thermal–hydraulic phenomena occurring in a passive mitigation strategy. The analysis is based on 2-inches guillotine Direct Vessel Injection (DVI) line break transients.

Computational fluid dynamics modelling of lead natural convection and solidification in a pool type geometry

Lead-cooled fast reactors (LFRs) are being studied by the organisations in the Generation IV International Forum (GIF) due to molten lead’s good thermodynamic properties, nuclear sustainability and safety. The study of lead solidification in a lead-cooled fast reactor is critical for the safety analysis of the reactor. Lead freezing may lead to overheating of the fuel assemblies or other components in the primary circuits. An activity that is focused on the development of the numerical models that deal with lead thermal hydraulics and solidification was ongoing within the H2020 project SESAME. The computational activity was supported by an experimental campaign. The SESAME stand experimental facility was assembled and operated at the Research Centre Rez (CVR) for the collection of thermal-hydraulic data on lead natural convection and solidification in a vessel type geometry. Simultaneously, two computational fluid dynamics (CFD) models of the SESAME stand were developed using ANSYS FLUENT and STAR-CCM + software.The models are benchmarked against the experimental data for both the steady-state and transient regimes. The methodology of the ANSYS FLUENT model has been described in detail, and the results were compared with both the experimental data and the STAR-CCM + model. The capability of the numerical model to deal with the lead thermal–hydraulic phenomena and their shortcomings is discussed. The challenges and the lessons learned from both the experimental and numerical activities are presented to support the development of computational tools for the lead-cooled nuclear reactors and their safety assessment.

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.

Assessment of SIMMER-III code in predicting Water Cooled Lithium Lead Breeding Blanket “in-box-Loss of Coolant Accident”

The in-box Loss of Coolant Accident is a major safety concern for the Water Cooled Lithium Lead Breeding Blanket design. SIMMER-III code has been modified by University of Pisa and ENEA C.R. Brasimone to perform deterministic safety investigation of such accidental scenario. Thanks these modifications, this version of the code has unique features for dealing with the PbLi water reaction related phenomena. The reliability of the code requires a systematic validation activity, carried out applying a standard methodology based on a three-steps procedure and through qualitative and quantitative evaluations. The methodology was applied to the LIFUS5 campaign. Post-test analyses highlighted open issues of test execution and of experimental data, as well as code limitations and capabilities. Nevertheless, the modified version of SIMMER-III code for fusion application is able to predict the relevant thermal-hydraulic phenomena during PbLi/water interaction.

The analysis of density wave instability phenomena of supercritical water in two parallel channels

The density wave instability is a very important thermal hydraulic phenomenon in water reactors design. The supercritical water is a kind of special fluid state which has both the characteristics of single-phase and two-phase fluids. Due to the sharp fluid properties variation near the pseudocritical point, the density wave instability of supercritical water has its own peculiarities. The stability analysis of supercritical fluid has been performed by some researchers. Some experimental results are available now to verify the numerical tools and promote the knowledge of flow instability phenomena at supercritical pressure. In this paper, the system analysis code Relap5 is used as the tool to help explain the mechanism of the density wave instability. The calculation method was verified with the experiments. Parametric effects on flow instability were discussed and stability boundaries were presented. The dimensionless parameters were analyzed which showed good potential to be used for the instability analysis of supercritical water.

CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants

The “dead-leg” in nuclear power plants (NPP) are pipes, which are connected with the primary loop and isolated by two check valves setting at both ends. It is a typical enclosed cavity. As the temperature and pressure in primary side is higher, the coolant in tube is heated by the valve and becoming two phase state, resulting in the corrosion of valves. In present work, the CFD method was employed to study the single phase and two phase thermal hydraulic phenomena in enclosed cavity. The buoyancy was calculated by solving compressible governing equations, and the mixture multiphase model and improved compressible phase change model were used to calculate the boiling process. The models are validated against with the experiment data and the simulation results show great agreement. The coolant natural convection occurs and the significant thermal stratification exists in the vertical direction. The void fraction increases at first and tends to a stable value finally. The vapor accumulates in the upper part of the pipe and the corrosion in the region is more serious. The trend of pressure increasing is consistent with void fraction and self-pressurization phenomenon is caused by the compression of vapor. Increasing the initial pressure could reduce boiling process greatly and it is an effective means to alleviate “dead-leg” phenomenon. This work is meaningful for the deep understanding of thermal hydraulic phenomena in enclosed cavity and provides valuable guidance for engineering.

3D thermal hydraulic characteristics analysis of pool-type upper plenum for lead-cooled fast reactor with multi-scale coupling program

The large space plenum of the lead-cooled reactor has complex three-dimensional thermal hydraulic phenomena such as thermal stratification and coolant mixing, which affects the safety performance of the reactor. In this paper, based on the multi-scale coupling program ATHLET-OpenFOAM, three-dimensional simulation of the hot pool for lead-cooled fast reactor M2LFR-1000 is carried out. Considering the axis symmetry of the hot pool model and reducing the amount of calculation, 1/8 model of the hot pool is selected to build the model by CFD, and the rest is simulated by system program ATHLET. The steady-state operation condition is first calculated by coupling program. Based on the calculation of steady-state condition, the main thermal parameters of M2LFR-1000 primary coolant system and the complex three-dimensional phenomena of the upper plenum under unprotected loss of flow accident (ULOF) are studied, and the results of multi-scale coupling calculation are compared with those of ATHLET alone. The results show that there is an obvious thermal stratification in the inner region and the outer region of the upper plenum in the early period of the accident, which slows down the speed of establishing natural circulation. During ULOF accident, there are obvious vortex flow changes in the inlet outer side and the outlet area of the upper plenum.

Counterpart test for LSTF 1% cold-leg break LOCA (SB-CL-32) utilizing ATLAS test facility

The counterpart test of the OECD-ATLAS project named A5.1 was performed using a thermal–hydraulic integral effect test facility, ATLAS. The target scenario for the A5.1 test was the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action. The test objective is to investigate the thermal hydraulic phenomena during a cold leg SBLOCA such as core heat-up, loop seal clearing (LSC), and the effect of AM actions. To achieve the test objectives, scaling analysis between LSTF and ATLAS was conducted first. The detailed test conditions such as initial and boundary conditions and sequence of events were determined based on the scaling analysis result. The experimental result showed that overall thermal hydraulic behavior which can be expected in this kind of SBLOCA scenario was observed. However, significant differences were found in the A5.1 test result compared with the LSTF SB-CL-32 test result in the view point of the loop seal clearing phenomenon and the maximum heater rod surface temperature behavior. These differences result from the different design between two test facilities which are refer to the different prototype nuclear power plant. Thus, several test condition parameters were analyzed as the sensitivity study utilizing the thermal hydraulic system analysis code to check their effect on the system response. From the sensitivity analysis result, it was confirmed that the A5.1 test was well modeled as a counterpart test of LSTF and realized the thermal hydraulic behavior during SBLOCA transient. This test result can be utilized to evaluate the prediction capability of thermal hydraulic system analysis codes and to validate the scale-up capability of integrated effect test (IET) test data to the nuclear power plant condition.

Numerical Analysis of the CIRCE-HERO PLOFA Scenarios

The present work deals with simulations carried out at the University of Pisa by using the System Thermal Hydraulics code RELAP5/Mod3.3 to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone Research Centre on the CIRColazione Eutettico—Heavy liquid mEtal pRessurized water cOoled tubes (CIRCE-HERO) facility. CIRCE is an integral effect pool type facility dedicated to the study of innovative nuclear systems and cooled by heavy liquid metal, while HERO is a heat exchanger heavy liquid metal/ pressurized cooling water system hosted inside the CIRCE facility. Beside the H2020 project Multi-Purpose Hybrid Research Reactor for High-Tech Applications (MYRRHA) Research and Transmutation Endeavour (MYRTE), a series of experiments were performed with the CIRCE-HERO facility, for both nominal steady-state settings and accidental scenarios. In this framework, the RELAP5/Mod3.3 code was used to simulate the experimental tests assessing the heat losses of the facility and analyzing the thermal hydraulics phenomena occurring during the postulated Protected Loss Of Flow Accident (PLOFA). The modified version Mod. 3.3 of the source code RELAP5 was developed by the University of Pisa to include the updated thermo–physical properties and convective heat transfer correlations suitable for heavy liquid metals. After reproducing the facility through an accurate nodalization, boundary conditions were applied according to the experiments. Then, the PLOFA scenarios were reproduced by implementing the information obtained by the experimental campaign. Sensitivity analyses of the main parameters affecting the thermofluidynamics of the Lead-Bismuth Eutectic (LBE) were carried out. In the simulated scenario, the LBE mass flow rate strongly depends on the injected argon flow time trend. The numerical results are in agreement with the experimental data, however further investigations are planned to analyze the complex phenomena involved.

Integral Effect Test and MARS-KS Calculation with Uncertainty Propagation Analysis for Direct Vessel Injection Line Break Intermediate-Break Loss-of-Coolant Accident

To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect test using the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test facility was conducted as the B3.2 test item of the international cooperation project Organisation for Economic Co-operation and Development (OECD)–ATLAS Project Phase 2 (ATLAS-2) (OECD-ATLAS2). The initial and boundary conditions for the test were determined referring to the Advanced Power Reactor 1400 MW(electric) (APR1400) as a prototype with three-level scaling methodology. A single-failure criterion was applied to the operation of the safety injection pump (SIP), and four safety injection tanks (SITs) were available to cool down the reactor coolant system. In the test result, as the break nozzle was located at the DVI line, the clearance of the upper downcomer could make an effective flow path of the steam toward the break and quench the reactor core. Maximum cladding temperature was measured before clearance of the upper downcomer. Coolant inventory in the reactor pressure vessel was maintained due to the safety injection without any further core heatup. So, it was proved that the current design of the safety systems in APR1400 had a sufficient long-term cooling capability with a single SIP during a DVI line break IBLOCA. The ATLAS test data were utilized to evaluate the prediction capability of the thermal-hydraulic system code Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) for a DVI line break IBLOCA scenario. The calculation result with the uncertainty propagation analysis using the PArallel computing Platform IntegRated for Uncertainty and Sensitivity analysis (PAPIRUS) toolkit proved that major phenomena such as uncovery of the core or intermittent injection of the SIT flow could be reasonably predicted by the MARS-KS code.

Numerical Simulation of the Solid Particle Sedimentation and Bed Formation Behaviors Using a Hybrid Method

In the safety analysis of sodium-cooled fast reactors, numerical simulations of various thermal-hydraulic phenomena with multicomponent and multiphase flows in core disruptive accidents (CDAs) are regarded as particularly difficult. In the material relocation phase of CDAs, core debris settle down on a core support structure and/or an in-vessel retention device and form a debris bed. The bed’s shape is crucial for the subsequent relocation of the molten core and heat removal capability as well as re-criticality. In this study, a hybrid numerical simulation method, coupling the multi-fluid model of the three-dimensional fast reactor safety analysis code SIMMER-IV with the discrete element method (DEM), was applied to analyze the sedimentation and bed formation behaviors of core debris. Three-dimensional simulations were performed and compared with results obtained in a series of particle sedimentation experiments. The present simulation predicts the sedimentation behavior of mixed particles with different properties as well as homogeneous particles. The simulation results on bed shapes and particle distribution in the bed agree well with experimental measurements. They demonstrate the practicality of the present hybrid method to solid particle sedimentation and bed formation behaviors of mixed as well as homogeneous particles.