Thermal-hydraulic Code Research Articles

Validation of three-dimensional simulation method for two-phase flow in triangular-pitch tube bundle in secondary side of steam generators on porous two-fluid model

ABSTRACT In the design process of a steam generator, a porous-media-approach computational code is utilized for simulating three-dimensional two-phase flow behavior in the secondary side of steam generators. For the design of next-generation steam generators, the advanced thermal-hydraulic analysis code based on the two-fluid model has been developed by implementing constitutive equations into the ANSYS fluent. For code validation, experiments using two-phase simulant fluids were performed to collect data in a simulated steam generator with a triangular tube array. The simulant two-phase flow system selected in the experiment was an adiabatic sulfur hexafluoride gas and liquid ethanol system, which allowed us to achieve the density ratio of prototypic steam-water flow with low pressure (=0.68 MPa). In the experiment, void fraction and gas–liquid interfacial velocity distributions along U-tubes were measured. The code validation was conducted by analyzing code predictions for measured distributions of void fraction and interfacial velocity under prototypic full load and partial load conditions, i.e., 50–80% flow rate conditions. Their bias and random errors were evaluated. The reasonable random errors demonstrated the validity of the newly developed code based on the two-fluid model in terms of predictions of void fraction and interfacial velocity in the secondary side of steam generators.

Development of interfacial velocity correlation for benchmarking porous media based three-dimensional code for steam generator with a square rod array

ABSTRACT A steam generator is one of the essential components in Pressurized Water Reactors. One of its critical design issues is to prevent heat transfer tube damage caused by corrosion and flow-induced vibration. When designing a U-tube bundle of steam generators, a thermal hydraulic simulation code is used to predict velocity and density distribution along tubes, especially porous-media-based code, from the viewpoint of computational cost. The code first calculates a control-volume-averaged velocity. A mass conservation equation converts it to the area-averaged mixture volumetric flux at the smallest gap between two tubes. Then, the area-averaged mixture volumetric flux is converted to the local interfacial velocity at the center of the sub-channel using an interfacial velocity correlation. Comparison between measured and calculated interfacial velocities ensures the validity of the simulation code. The validation process for the code requires an interfacial velocity correlation. This study shows the newly developed interfacial velocity correlation applicable to two-phase flow in a tube bundle with a square array. The prediction uncertainty of the correlation is +6.7% (bias) ± 24.2% (random error). The correlation based on an analytical equation suggests that the correlation has good scalability to apply to a wide range of flow and geometrical parameters.

One-step coupled calculations (Serpent-OpenFOAM) for a fuel rod of the IPR-R1 triga reactor

In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels. The aim of this paper is to present a practical approach to perform coupled calculations between neutronic (Monte Carlo) and thermal-hydraulic (CFD) codes. For this purpose, is necessary to evaluate the influence of the water thermal-physical properties temperature variations on keff parameter. Besides that, Serpent Nuclear Code was used for the neutronics evaluation, while OpenFOAM was used for thermal-hydraulics. OpenFOAM si- mula-tions were made by using a modified chtMultiRegionFoam solver, developed to read Serpent output correctly. The neutronic code was used without any modifications. The results shows that this coupled calculations were consistent and that leads to encouraging further methodology development and its use for full core simulation. Also, the results shows good agreement with calculations performed using other version of OpenFOAM and Milonga as neutronic code.

CSPACE for a simulation of core damage progression during severe accidents

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

Development of Subchannel Void Sensor for Wide Pressure and Temperature Ranges and Its Application to Boiling Flow Dynamics in a Heated Rod Bundle

A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes.

Thermal-hydraulic two-phase modeling of reactivity-initiated transients with CATHARE2 – Application to SPERT-IV simulation

Reactivity-Initiated Accidents (RIA) in nuclear reactor cores are very complex multiphysic transients. They consist in a very fast power excursion in the core, leading to wall temperature excursions. Many experiments have shown that the heat exchanges coefficients during fast transients differ from those in steady-states. The thermal-hydraulic phenomenology of such transients remains complex and some experimental studies allowed to better understand and quantify the heat exchanges between the wall and the fluid, for both single and two-phase flows.The current study lies in a continuous effort of developing and validating a coupled neutronic and thermal-hydraulic code for the analysis of protected and unprotected transient behavior of reactors. More specifically, this study deals with the extension of the thermal hydraulic model of the CATHARE2 code to fast transient configurations. The extended CATHARE2 model is validated against experimental results from SPERT-IV. This reactor was driven mainly by the coolant density reactivity feedback. Thanks to that, many flow regimes can be observed in the core during the transient. That allows a more accurate evaluation of the adequacy of available transient two-phase flow models and correlations. The results, that represent the state of modeling, allow concluding that introduced models in CATHARE2 are able to simulate RIA tests such as SPERT-IV involving fast transient boiling.

Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems

Despite system thermal–hydraulic codes were extensively validated for transient simulations of LWR, several activities highlighted limited capabilities of these tools to model heat transfer within in-pool passive power removal system. Discrepancies with experimental results were related to the underestimate of pool boiling and film condensation heat transfer coefficients. Thus, the DIAEE of “Sapienza” University of Rome developed a modified version of RELAP5/Mod3.3, able to handle fundamental heat exchange phenomena involved in passive in-pool safety systems. A primary validation procedure has been performed for separated and integral effects. Dealing with nucleate boiling, the Root Mean Squared Relative Error (RMSRE) of wall superheat has been reduced from 1.290 to 0.182. Concerning film condensation, wall temperature RMSRE has been reduced from 0.192 to 0.058. The integral effect assessment has involved an experimental test of the PERSEO facility. The qualitative comparison between experiments and calculations has highlighted significant improvements of the modified RELAP5/Mod3.3.

Application of the coupled code ATHLET-ANSYS CFX for the simulation of the flow mixing inside the ROCOM test facility

System thermal-hydraulic (STH) codes are based on 1D lumped parameter approach, which has limited capabilities to model 3D flow mixing phenomena. To close this gap, STH programs are coupled with modern computational fluid dynamics (CFD) tools. The development and validation of coupled 1D-3D numerical codes received recently strong attention in numerous international projects like SESAME, THINS, NURISP, etc. GRS coupled its own STH code ATHLET with the CFD programs ANSYS CFX and OpenFOAM. Within the validation process of ATHLET-ANSYS CFX the ROCOM PKLIIIT1.1 experiment was simulated. Ethanol-water is injected in the water-filled primary ROCOM circuit, leading to complex flow mixing phenomena in the facility. In this work PKLIIIT1.1 is analyzed and the ATHLET-ANSYS CFX multiscale results are compared with experimental data for validation purposes.

Validation of Serpent-SUBCHANFLOW-TRANSURANUS pin-by-pin burnup calculations using experimental data from the Temelín II VVER-1000 reactor

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

Multiphysics Coupling Methods for Molten Salt Reactor Modeling and Simulation in VERA

New multiphysics coupling capabilities for molten salt reactor (MSR) analysis have been developed in the Virtual Environment for Reactor Applications (VERA). This development consisted of two main efforts. First, a generic species transport module was added in the CTF code, which is the thermal-hydraulics (TH) code for VERA. This module uses the velocity fields for which CTF solves during the TH calculation to transport species through the core and around the primary loop. Additionally, a gas sparging model has been added to CTF to model the movement of certain species, namely, fission products such as xenon gas, to transport between the molten salt and gas bubbles present in the salt. The second effort in this development was coupling this capability to VERA’s neutron transport code MPACT. This effort focused on coupling the detailed TH transport models in CTF to MPACT to account for feedback effects in the neutron transport calculations. Finally, the thermochemistry code Thermochimica has also been coupled to VERA. Thermochimica performs pointwise calculations for chemical potential and Gibbs free energy and determines what phases are produced by the temperature, pressure, and elemental concentrations at different locations in the primary loop. These capabilities are demonstrated using a model of the Molten Salt Reactor Experiment (MSRE). This reactor operated at Oak Ridge National Laboratory in the 1960s, providing sources of experimental data that were used to develop the model. Various combinations of species were modeled using VERA’s new multiphysics coupling capabilities. Species distributions and reactivity effects behaved as expected for the MSRE model, demonstrating that the coupling is behaving correctly and causing appropriate feedback. The results of these calculations show the potential for VERA to be used for a wide variety of MSR analyses.

Extension and validation of the SubChanFlow code for the thermo-hydraulic analysis of MTR cores with plate-type fuel assemblies

Many research reactors with plate-type fuel are operated worldwide and the thermal hydraulic core analysis is done with dedicated codes or with 1D system thermal hydraulic codes. At the Karlsruhe Institute of Technology (KIT), investigations are underway to develop advanced coupled neutronic/thermal hydraulic simulation tools for the Safety analysis of research reactors with plate-type fuel. For this purpose, the subchannel thermal hydraulic code SubChanFlow was extended and validated using relevant tests. The SubChanFlow extension consists of the addition of a heat conduction module for a thin plate and the implementation of heat transfer correlations for rectangular channels. For the validation of the modified SubChanFlow code, temperature data from RA-6 experiment obtained in a Reynolds range of 7×103<Re<1.4×104 were used.The RA-6 experiment was carried out at the Bariloche Atomic Center to support the development of MTR research reactors. Different heat transfer correlations e.g., the Colburn, Dittus-Boelter, Gnielinski, Sieder-Tate, and Y-Sudo were used to simulate the RA-6 tests. The comparison of the plate temperatures measured along the active height with the predicted values using the extended SubChanFlow code shows a deviation ranging from ± 4 °C to ± 18 °C for the Colburn and Y-Sudo correlations, respectively, which is promising compared to results of other codes.

Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space.In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation.The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses.

Numerical analysis of the equalization capability of submerged perforated sheets for wwer-1500 horizontal steam generator

The results of a study on the equalization of a non-uniform steam load using different submerged perforated sheets (SPSs) are presented for a full-scale horizontal WWER-1500 steam generator (SG). The calculations performed with the STEG code are used. A short description of the 3D thermal-hydraulic code STEG is presented. A new model of gravitational separation based on empirical correlations for the size distribution of water droplets and for the modal droplet diameter is implemented in the STEG code. Three variants of the distribution of an SPS perforation degree are analyzed. Various characteristics are compared: the distributions of the superficial velocity on the SPS and evaporation surface, the fractions of the steam flow rate with high velocity in the steam region and the steam moisture at the outlet from the SG. The SPS perforation degrees, which significantly improve the equalization of the non-uniform steam load, are found using the STEG code.

Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre.A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid.

Multi-physics evaluation of the steady-state operation of an Aqueous Homogeneous Reactor for producing Mo-99 for the Brazilian demand

The studies summarized in this paper aims to predict the steady state operation of a low-enriched uranium fuel ARGUS type aqueous homogeneous reactor for producing 99Mo to meet the domestic demand of Brazil through a coupled multi-physics (Neutronics + Thermal-hydraulics) evaluation. The coupled multi-physics evaluation included aspects related to the neutronic behavior such as fission induced energy deposition profile, medical isotopes production; and the thermal-hydraulic behavior such as temperature, velocities and gas volume fraction profiles. The methodology followed for the multi-physics and multi-scale coupling of the neutronic and thermal-hydraulic codes (MCNP + ANSYS-CFX), discussed in detail in this paper, represent one of the main outcomes of the current study. The methodology was tested for two different operating configurations of the ARGUS reactor, the original high-enriched uranium configuration used since 1981, and the new low-enriched uranium configuration after the conversion process during 2012-2014. The calculations carried out showed that the reactor, in the studied configuration, is able to produce 246.5 six days Curie of 99Mo in operation cycles of five days. Which is equivalent to more than a third of the estimated Brazilian demand for 2025.

Development of a multi-physics coupling system based on ICoCo interface and its validation on NEA-OECD core transient benchmark

This paper describes the development of a multi-physics coupling system of Thermal-Hydraulics (T-H) code CORTH and reactor dynamics diffusion code CORCA-K for nuclear core analysis. The coupling system is based on a typical server-client coupling system ICoCo (Interface for Code Coupling), where each code is first wrapped by the ICoCo interface and then compiled into a shared library. And a 3rd-party library named MEDCoupling is adopted to handle the mesh interpolation and field mapping between CORTH and CORCA-K.The prediction capabilities of CORCA-K/CORTH are proved by the analysis of the NEA-OECD rod ejection transient benchmark, which proves that the developed coupling system CORCA-K/CORTH has good accuracy in nuclear reactor core transient calculation.

Total loss of flow benchmark in CIRCE-HERO integral test facility

In the framework of the HORIZON2020 SESAME European project, an experimental campaign has been carried out on the large Lead-Bismuth Eutectic pool integral effect CIRCE facility at ENEA Brasimone Research Center, implementing the HERO test section. This test section is aimed at supporting the development of the ALFRED design. Within the SESAME Project, three protected loss of flow accident tests have been designed and executed, identified as SE-Test1, SE-Test2, and SE-Test3. The third test (SE-Test3) has been selected by the project participants for a benchmark activity involving system thermal-hydraulic codes and coupled system thermalhydraulic/ CFD. This activity has been divided into two phases: a blind phase, and a post-test phase. The present paper illustrates the results achieved during the post-test phase of the CIRCE-HERO benchmark activity. Four participants simulated the experimental test. Three (ENEA, SCKCEN and UNIROMA1) used different versions of RELAP5 code, i.e. RELAP5-3D and RELAP5Mod3.3, and one (NRG) used a coupled approach based on an in-house system thermal-hydraulic code (SPECTRA) and the commercial code ANSYS CFX. The benchmark activity hereafter presented is a relevant exercise for evaluating and comparing the predictive capabilities of system thermal-hydraulic codes, CFD codes and coupled technics, in relation to phenomena occurring during the transition between forced and natural circulation (e.g. loss of flow) in a heavy liquid metals Generation IV system. The results of the numerical exercise showed an adequate capability of the codes to reproduce the relevant phenomena involved during the experiment. The SYS-TH codes have been more accurate than SYS-TH/CFD in predicting the trend of the main parameters during the transient, reproducing them with quite satisfactory accuracy. SYS-TH/CFD simulation, instead, provided a more satisfactory representation of the pool thermal stratification. The results highlighted that the planning and the execution of experiments fully devoted for the code V&V process is needed for the further development of such numerical tools, in particular for SYS-TH/CFD coupled approaches, whose use is more recent respect to SYS-TH codes.

An analysis of air-water flow phenomena due to a pipe break under sub-atmospheric pressures using TRACE

For advanced research reactor designs that have a high core power with downward coolant flow in the core, it is important to analyze a unique postulated pipe break scenario under sub-atmospheric pressures. In order to benchmark the capability of thermal–hydraulic codes such as TRACE to predict flow behavior under these conditions, we must first compare TRACE analyses to reference experimental data for air–water flow near atmospheric pressure and temperature conditions. Given a successful comparison between predictions and test data, we then use TRACE to analyze air–water two-phase flow phenomena during a postulated pipe break accident with the initial conditions prototypic of a particular research reactor design. The normalized flow rate, pressure, and void fractions were predicted to describe the overall behavior of the pipe break phenomena. These results indicate that the amount of air ingested from the environment into the reactor coolant system through the pipe break has the dominant effect on the downstream void fraction as well as flow and pressure response, and thereby the time available (delay time) to take compensatory actions before deleterious effects are felt by the reactor cooling system. This defined delay time is the figure of merit that can be used for design of the system. Since the two-phase flow under sub-atmospheric pressure in the system can be affected by various design conditions, as well as initial and boundary conditions, the sensitivity analyses were performed to better understand their effect. A set of parameters were chosen based on their quantitative effect on the overall transient behavior; i.e., pool inventory, system design flow rate, and pipe break size. The various quantitative characteristics of the sensitivity analysis results showed that TRACE was capable of analyzing this unique air–water two-phase flow phenomena under sub-atmospheric pressure and will be utilized in the safety analysis for this research reactor system.

Development and verification of the coupled thermal–hydraulic code – TRACE/SCF based on the ICoCo interface and the SALOME platform

The system thermal–hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) and the sub-channel code SubChanFlow (SCF) have been coupled together based on the Interface for Code Coupling (ICoCo) and the SALOME platform. It is a server-client system where TRACE and SCF are computational engines, ICoCo transfers the coupling data, and SALOME supervises the two clients’ synchronization. TRACE and SCF run serially. The coupled code adopts the domain-overlapping (assisted by a Dynamic Implicit Additional Source (DIAS) method) and the Operator Split (OS) explicit coupling approaches. The MEDCoupling library manages the field mapping between different meshes of TRACE and SCF. The YACS module in SALOME provides developers with useful tools that can implement numerical codes in SALOME. It provides users with Graphic User Interfaces (GUI) that can construct and execute the coupling simulations as well. Users can conveniently build/optimize/execute the coupling schemes in SALOME. The coupled code has been verified with an AP1000 feedwater-temperature-drop transient.

Break location influence in pressure vessel SBLOCA scenarios

The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused on simulating different hypothetical accidental scenarios in the PV using the thermal-hydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios.