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European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems
Nuclear Engineering and Design 290 (2015) 2–12
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Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems X. Cheng a,∗ , A. Batta a , G. Bandini b , F. Roelofs c , K. Van Tichelen d , A. Gerschenfeld e , M. Prasser f , A. Papukchiev g , U. Hampel h , W.M. Ma i a
Karlsruhe Institute of Technology (KIT), Germany Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy c Nuclear Research and Consultancy Group (NRG), The Netherlands d Studiecentrum voor Kernenergie – Centre d’étude de l’Energie Nucléaire (SCK·CEN), Belgium e Commissariat à l’Energie Atomique (CEA), France f Paul Scherrer Institute (PSI), Switzerland g Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Germany h Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR), Germany i Kungliga Tekniska Högskolan (KTH), Sweden b
h i g h l i g h t s • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized.
a r t i c l e
i n f o
Article history: Received 27 October 2014 Accepted 6 November 2014
a b s t r a c t Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are:
• Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena.
• Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena.
• Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue. © 2014 Elsevier B.V. All rights reserved.
Abbreviations: ADS, accelerator-driven subcritical nuclear systems; CFD, computational fluid dynamics; DHR, decay heat removal; DNS, direct numerical simulation; GFR, gas-cooled fast reactor; GIF, Gen-IV International Forum; HLM, heavy liquid metal; HTR, high temperature reactor; LBE, lead bismuth eutectic; LDA, laser Doppler anemometry; LES, large eddy simulation; LFR, lead-cooled fast reactor; LMR, liquid metal reactors; MRI, matching refractive index; MSR, molten salt reactor; PTV, particle tracking velocimetry; RANS, Reynolds-averaged Navier–Stokes equations; SC, supercritical; SCWR, supercritical water cooled reactor; SFR, sodium-cooled fast reactor; THINS, thermal-hydraulics of innovative nuclear systems; VHTR, very high temperature reactor. ∗ Corresponding author. E-mail address: [email protected] (X. Cheng). http://dx.doi.org/10.1016/j.nucengdes.2014.11.007 0029-5493/© 2014 Elsevier B.V. All rights reserved.
X. Cheng et al. / Nuclear Engineering and Design 290 (2015) 2–12
1. Introduction For the long-term development of nuclear power, reactors of generation IV (Gen-IV) with enhanced safety, economics, sustainability and non-proliferation features need to be developed. The Gen-IV International Forum (GIF) recommended six innovative nuclear energy system for meeting future energy challenges and proposed a technology roadmap for Gen-IV nuclear energy systems (USDOE, 2002). These six innovative nuclear energy systems are very high temperature reactor (VHTR), gas-cooled fast reactor (GFR), sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), supercritical water cooled reactor (SCWR) and molten salt reactor (MSR). One aspect of sustainability is the reduction of nuclear waste. Nowadays, management of nuclear waste becomes a key issue in the public acceptance of nuclear energy. Extensive studies in the last years showed that partitioning and transmutation is a promising approach of the waste management. Transmutation using accelerator-driven subcritical nuclear systems (ADS) has attracted strong attention worldwide due to its claimed favorable safety features and its high incineration rate of nuclear waste (IAEA, 2003). Thermal-hydraulics is recognized as a key scientific subject in the development of the different innovative reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by coolants, flow channel structures formed by different fuel lattice arrangements and primary circuit lay-out (pool vs. loop). They result in different micro- and macroscopic behavior of flow and heat transfer and require specific models and advanced analysis tools. On the other hand, many common thermal-hydraulic issues are identified among various innovative nuclear systems (THIRS, 2008). In 2010, the EU project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) was launched (www.ifrt.kit.edu/thins/), which focuses on several crosscutting issues and synergizes infrastructure for the thermal-hydraulic research of the innovative nuclear systems in Europe. The overall objectives of the THINS project are the development and validation of new physical models and qualification of numerical analysis tools. Specific objectives are:
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based on advanced physical models and numerical methodology. Coupling of the code solutions at various scales and the qualification of the coupled calculations extend the applicability and ensure the reliability of the numerical platform. The THINS project has a duration of five years and will terminate in January 2015. In January 2014 the International Workshop on Thermal-Hydraulics of Innovative Nuclear systems, THINS-2014, took place at the University of Modena, Italy (http://www.thins2014.unimore.it) with the aim to review the project progress, to disseminate the main achievements of the project among the international nuclear community and to foster the exchange of the newest scientific and technical results between researchers and experts. In this paper, the main objectives, technical tasks, scientific methodology and some main results are presented, in order to give an introductory overview of the THINS project and provide guidance to this special journal issue. 2. Technical tasks The THINS project is structured with the following six work packages (WPs): • • • • • •
WP1: Advanced reactor core thermal-hydraulics. WP2: Single phase mixed convection. WP3: Single phase turbulence. WP4: Multi-phase flow. WP5: Code coupling and qualification. WP6: Education and training.
The first five WPs are devoted to the individual crosscutting issues. Experiments are foreseen to provide experimental evidence and fundamental test data base. New physical models are developed and further employed to improve codes. The last WP is devoted to the education and training of young nuclear engineers and researchers. 2.1. Advanced reactor core thermal-hydraulics
• Establishment of a data base for the development and validation of new physical models and numerical codes for a more accurate description of the selected crosscutting thermal-hydraulic phenomena. Generic experiments are performed in the THINS project to produce a comprehensive data base for the validation purpose. In addition, direct numerical simulation (DNS) will be performed to provide numerical data base, which is of crucial importance for the development of turbulence models. With this project, a data base will be established for the fundamental thermal-hydraulic issues occurring in the innovative nuclear systems. • Establishment of an experimental platform for the thermalhydraulic research of the innovative nuclear systems. In the past, also in the frame of the previous European projects, experimental facilities, specifically for the thermal-hydraulics of innovative nuclear systems, have been built and operated at various institutions (http://vella.brasimone.enea.it/na.htm). The THINS project will make the optimum utilization of the available European experimental facilities and expertise, combine the resources available and establish a European experimental platform. • Establishment of a numerical platform for the design analysis of the innovative nuclear systems. Numerical codes for nuclear thermal-hydraulics cover various classes of spatial scales, i.e. system analysis based mainly on lumped parameter approach, sub-channel analysis specifically for fuel assembly and reactor core and CFD codes for detailed local flow behavior. With the THINS project, more reliable and validated codes will be proposed
Design of innovative reactor cores requires detailed analysis of the thermal-hydraulics within the fuel assemblies with high resolution inside individual sub-channels. The goal of this work package is to provide validated and verified simulation tools of the coolant flow within the reactor core components for typical states encountered in liquid metal cooled fuel assembly. The THINS project considers different numerical approaches applicable to the design analysis of innovative reactor cores. The first approach is based on sub-channel analysis codes. Efforts are made to introduce advanced numerical methods, e.g. coarse grid simulation, and to improve the physical models suitable for specific conditions of innovative reactor cores. The purpose of the second approach is to develop CFD simulation tools with more advanced models than the current state-of-the-art RANS modeling and improved accuracy. The complex nature of heat transfer in pebble bed requires the development of new algorithm techniques to describe the core, as well as robust methods capable to treat convective–conductive–radiative heat transfer with fluid flow across complex structure of pebble bed. In the THINS project a method is proposed based on macroscopic modeling of flow phenomena in pebble bed reactors. The macroscopic properties are obtained by LES simulation technique applied to a geometry consisting of spheres in an array with irregular structure. Two experiments are considered in this WP to provide test data for the validation purpose, as shown in Figs. 1 and 2. Test data in fuel assemblies with liquid metal for mixed and natural convection
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Fig. 1. Rod bundle tests with LBE in KIT.
conditions will be produced with the rod bundle test section, see Fig. 1, performed in the KALLA laboratory of KIT. Heat transfer tests on single rod with textured surface, see Fig. 2, are performed with local temperature measurement in the gas and on the heated rod surface along with LDA to provide experimental data on integral and on the sub-texture scale. The third test data set was obtained in a pebble bed configuration simulating a “fuel compact” in the HTR/VHTR (Hassan and Dominguez-Ontiveros, 2008). Velocity fields inside the packed bed are measured using Particle Tracking Velocimetry (PTV) combined with the Matching Refractive Index (MRI) methods. 2.2. Single phase mixed convection The use of the natural convection to remove residual heat from the core is widely adopted in the Gen-IV reactor concepts that implement passive safety features to enhance reliability. Correct prediction of the onset and stability of the natural convection is essential for the design of the decay heat removal system and for the transient analysis. In addition, the pool-type configuration proposed for most of the innovative Liquid Metal Reactors (LMR) to compact the primary system and to limit the risk of coolant leakage requires numerical tools able to simulate local 3-D phenomena, such as thermal stratification, for the characterization of the
Fig. 3. The PHENIX reactor in France.
thermal loads on the primary vessel and components as well as for the design of in-pool heat exchangers. In this WP, simulation of dynamic behavior of the innovative nuclear systems is mainly based on system analysis codes originally developed for LWR. The applicability of these codes to support the design process and safety analysis of LMR will be extended and validated. Main extension is focused on the implementation of closure relationships suitable for innovative cooling fluids and for a large spectrum of flow conditions, from forced convection to natural convection. For the simulation of local 3-D phenomena, CFD approach is applied. The qualification of the CFD models on the separate effect tests and the validation of the system analysis codes on integral tests represent one of the main tasks. To achieve this target, four experiments are planned, one with sodium and the other three with LBE. One of the four test facilities ESCAPE is being constructed in the frame of the THINS project. However, its operation will start after the termination of the THINS project. The other three test facilities will provide test data and are presented in Figs. 3–5. The PHENIX end-of-life tests (Gauthé et al., 2012) represent a real reactor configuration and contain complex effects of prototypical systems, see Fig. 3. Due to the limitation in measurement instrumentation, however, data available for validation are restricted, e.g. to system dynamic behavior. In the CIRCE (Fig. 4) and TALL-3D (Fig. 5) test facilities, more sophisticated measurements are applicable. Test data for the qualification for both system codes and 3-D CFD codes are expected. Main focus of all three experiments is on flows at strong buoyancy effect, i.e. mixed convection conditions.
2.3. Single phase turbulence
Fig. 2. L-STAR tests with air in KIT.
For the coolants envisaged in the innovative nuclear reactors, usually experiments are very expensive and accurate measurements are challenging or even impossible. Therefore, application of CFD for prediction of various flow characteristics becomes an
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Fig. 6. DeLight facility with SC Freon in DUT.
Fig. 4. CIRCE facility with LBE in ENEA.
attractive and complementary practice used in the design and evaluation process of innovative nuclear reactors. It is well agreed (Roelofs et al., 2014) that one of the key issues ensuring a reliable CFD simulation is the modeling of turbulence. For innovative nuclear systems, two features are important and need to be considered in the turbulence modeling.
(i) The coolants envisaged for advanced reactor systems cover a wide range of fluids with various physical properties, e.g. the molecular Prandtl number varies from the order of 10−3 –103 . This implies the specific behavior and prediction of turbulence and represents a challenging task. (ii) At normal operating conditions, the fluids in all innovative nuclear systems considered are at single-phase flow conditions. Therefore, this WP is devoted to single-phase turbulence with the main objective to improve and to develop turbulence models for non-unity Prandtl number flows, their implementation in engineering tools and application to liquid metal and supercritical flows. Related to innovative nuclear reactors with operation at elevated temperature, this WP focuses on two main issues, i.e. the strong buoyancy effect on turbulence and the occurrence of temperature fluctuation which may lead to thermal fatigue. A large number of CFD codes are applied (see Section 3). For validation of turbulence models and qualification of CFD codes, three experiments are conducted, as illustrated in Figs. 6–8. Both Delight (Fig. 6) and SCMix (Fig. 7) test facilities uses supercritical (SC) Freon as working fluids. In DeLight facility heat transfer at SC Freon will be investigated with the main emphasis on the investigation of mechanism of heat transfer deterioration. The main purpose of SCMix experiments is the mixing behavior of supercritical fluids under strong density variation and buoyancy effect. The HOMER facility (Fig. 8) is designed for study mixing process of two gas components with different densities. 2.4. Multi-phase flow
Fig. 5. TALL facility with LBE in KTH.
Multi-phase flows are encountered in several innovative reactor systems. A sample of these flows forms the basis of the work in this WP. Free surface flows are present in the pool-type liquid metal cooled reactors (LMR). They are also key phenomena in the accelerator driven systems (ADS) with windowless spallation targets. Bubbly flows occur from the possible incidental interaction between water and heavy liquid metal (HLM) in lead alloy cooled reactor systems. In HTR/VHTR, graphite dust is generated by abrasion and transported in the coolant loop.
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Fig. 7. SCMix facility with SC Freon in DUT.
In this WP advanced free surface modeling methods, such as the ALE-Moving Mesh Algorithm (Maciocco, 2002), will be further developed and validated which are capable of capturing sharp fluid-vacuum interfaces and simulating near-interface subscale turbulence. Concerning the HLM/water interaction, mainly the SIMMER-III code is used to investigate the related phenomena. Complete simulation of the entire HLM/water interaction procedure requires the implementation of additional modules predicting energy release and fluid–structure interaction. For the safety analysis of HTR/VHTR, transportation behavior of contaminated graphite dust generated by abrasion of fuel plays an important role. Development of appropriate CFD models for dust transport simulations is then one of the main tasks of this WP. To provide test data for model development and numerical codes qualification, three experiments are implemented in this WP, as summarized in Figs. 9–11. The free surface test at KIT using LBE investigates the characteristics of free surface and the effect of various parameters. The LIFUS facility at ENEA is devoted to study the interaction of LBE with water. It focuses on the measurement of energy release during the LBE/water interaction. At the GPLoop facility at HZDR deposition and resuspension of particles at various flow and surface conditions will be investigated.
Fig. 9. Free surface tests with LBE in KIT.
qualification towards two scenarios entirely dedicated to innovative nuclear systems, i.e. LMR (sodium cooled, lead alloy cooled) with buoyancy effects and transition from forced convection to natural convection. Two complementary experiments are taken into consideration in the THINS project to provide test data for the qualification
2.5. Code coupling and qualification The overall objective of the work in this WP is the development and qualification of new multi-scale computational solutions applicable to innovative nuclear systems. A coupled multi-scale solution will combine the merits of various classes of codes and provide design tools with feasible efficiency and high accuracy. This task is devoted to the coupling of system codes with CFD codes and its
Fig. 8. HOMER facility with gas mixture in ETH.
Fig. 10. LIFUS facility in ENEA.
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3. Methodology applied 3.1. Technical approach
Fig. 11. GPLoop facility for particle deposition and resuspension in HZDR.
of the code coupling solution. The PHENIX reactor scale experiment (Fig. 3), with existing ultimate test data on a full scale safety issue scenario, and the HLM TALL-3D (Fig. 5) integral scale experiment on a separate effect scale safety issue scenario will produce a unique test data base, which is at the same time complementary to the existing data base for V&V of code coupling solutions.
2.6. Education and training More than 50 young researchers (Master students, PhD students and Post-Doc researchers) are directly involved in the THINS project. Most of the theses are mainly based on the THINS activities. In addition to project meetings, cluster workshops and training courses were organized for young participants. At the cluster workshops students’ presentations and invited topical lectures were integrated in the workshop program. This gave the participating students the opportunity to follow all scientific presentations focusing on the THINS project progress. The topical lectures went far beyond the tasks pursued within the THINS project and made the audience familiar with particular new developments in the field of nuclear reactor technology. Special training courses were provided to young participants related to the application of CFD codes and to the coupling methodology of CFD codes with system codes.
The methodology used in this project is indicated in Fig. 12. Based on phenomenological analysis of the identified crosscutting phenomena, experimental programs were defined with the purpose to provide experimental evidence to a better understanding of physical phenomena and test data for the development of models and validation of codes. Totally 12 experiments have been defined. Due to limited possibility of experiments with low Prandtl number fluids, direct numerical simulation (DNS) will be carried out, to provide additional data for the data base. In parallel, numerical investigations are carried out with three different scales of codes, i.e. system analysis codes, sub-channel analysis codes and CFD codes. Numerical support will be provided to experimental work in various phases, such as selection of measurement techniques, design of test facilities, definition of test matrix and test data analysis. The data base consisting of both experimental and numerical results can be then applied for the development of physical models and validation of simulation codes. The models considered in this project are for both system analysis codes and CFD codes, e.g. heat transfer and friction pressure drop in liquid metal for system analysis codes, and turbulence models for CFD simulations. To enhance the interaction between the work packages, three Clusters are introduced in the project management structure, to enhance the interaction between the work packages in which similar or related experimental or numerical approaches may be employed. Therefore, the clusters strongly support the realization of the project objectives. Three clusters have been established, i.e. (i) CFD model and validation; (ii) system code and validation; (iii) experimental techniques and data bases. The main tasks of clusters are to distribute important information between WPs which might be of general interest for the THINS partners, dealing with experiments, CFD or system code simulations, to organize cluster workshops and to organize required benchmarks. 3.2. Experimental activities Experimental studies are important tools and also necessary boundary conditions to achieve the main target of the project, i.e. development of models and more reliable simulation codes. Twelve experiments have been included in the THINS project and are summarized in Table 1. Real reactor data for sodium as a coolant is provided from the PHENIX end of life test. Six experiments use LBE
Fig. 12. Technical approach applied.
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Table 1 Summary of experimental studies. No.
Table 2 Summary of numerical codes used.
Name of test
Fluids
Main phenomena
1
PHENIX
Na
2
Rod bundle
LBE
3 4
E-SCAPE CIRCE
LBE LBE
TALL-3D Free surface LIFUS L-Star GPLoop DeLight SCMix HOMER
LBE LBE LBE and water Air Air and particles SC water SC water He and air
System dynamics, pool mixing Flow and heat transfer behavior Pool mixing Pool mixing, heat transfer in bundle Natural circulation Free surface shape LBE/H2 O interaction Local heat transfer Particle transportation Heat transfer Thermal mixing Mixing of gases
5 6 7 8 9 10 11 12
Figures 3
Codes CFD ANSYS-CFX
1 – 4 5 9 10 2 11 6 7 8
Star-CD/Star-CCM+ Fluent OpenFOAM TransAT SIMMER-III Armando THEMAT FLUIDITY System analysis RELAP5 ATHLET
as fluid. They are dealing with several crosscutting phenomena of LMR, e.g. mixed convection in large pool, mixed heat transfer in rod bundle, HLM/water interaction and free surface flow. It aims at validating system codes, CFD-codes and coupled system/CFD solutions. One experiment is devoted to heat transfer in gas cooled reactor cores with prismatic type fuel structure, whereas another experiment related to gas cooled reactor is dealing with graphite dust behavior, which is of crucial importance for the safety of HTR/VHTR. Two experiments are designed for supercritical water cooled reactors. In the experiment DeLight, heat transfer in single tubes under SCWR conditions will be carried out with the emphasis on heat transfer deterioration and its influencing parameters. The experiment SCMix deals again with the mixing behavior under strong buoyancy effect. At the facility HOMER, buoyancy driven mixing behavior will be investigated using gases of different density. Related to crosscutting thermal-hydraulic issues it is easily recognized that flow and heat transfer under strong buoyancy effect (mixed convection) is the key topic in this project. It is also important phenomenon in various kinds of innovative nuclear systems, as LMR, GCR and SCWR.
3.3. Numerical activities The final goal of the project is the development of advanced physical models, numerical simulation tools and their validation on experimental data. There is strong interaction between numerical activities and experimental studies. As mentioned in Fig. 12, numerical activities should in one hand provide support to experimental work and in the other hand utilize the experimental results to improve the numerical tools. In this project a large number of numerical codes are used, which can be divided into three categories, according to their spatial resolution. Table 2 summarizes the numerical codes used. In the CFD class, 11 codes are selected by the project partners. The focus is clearly on the widely used commercial codes like ANSYS-CFX, Star-CD/Star-CCM+, Fluent, TransAT, and the open accessible code OpenFOAM. However, for certain dedicated purposes also in-house codes like THEMAT, FLUIDITY and SIMMER-III are used. Each experiment is supported by several codes. Different codes might be used by the same partner for different experiments or even for the same experiment, to realize code-to-code comparison purpose. Five system analysis codes are selected, which are widely applied in the nuclear community. Key focus is clearly on the European codes, i.e. CATHARE from CEA, DYN2B from IRSN and ATHLET from GRS. For analyzing pool type LMR, coupling of system analysis code with CFD code will be realized, i.e. CATHARE with TRIO-U,
CATHARE TRACE DYN2B Sub-channel analysis Coarse mesh CFD
Related experiments TALL-3D, rod bundle, E-SCAPE, L-Star, free surface, LIFUS Rod bundle, TALL-3D, free surface, DeLight E-SCAPE, CIRCE, L-Star, DeLight, GPLoop Rod bundle, free surface, PHENIX, DeLight, HOMER, GPLoop Free surface LIFUS, CIRCE Free surface, with ALE-Moving Mesh Algorithm HOMER PGLoop, models for particle transportation CIRCE, E-SCAPE, TALL-3D, also coupling with Star-CCM+ TALL-3D, PHENIX, also coupling with OpenFOAM and ANSYS-CFX TALL-3D, PHENIX, also coupling with Trio-U TALL-3D, Rod bundle PHENIX Rod bundle
ATHLET with OpenFOAM, ATHLET with ANSYS-CFX and RELAP with Star-CCM+ . 3.4. Technical consortium The technical consortium of the project consists of 24 partners, of those 23 institutions from European Union and one partner from US, as illustrated in Fig. 13. The consortium includes nearly all relevant institutions in European Union involved in the nuclear thermal-hydraulics, especially dealing with crosscutting topics of thermal-hydraulics of innovative nuclear systems. 4. Summary of results achieved In this paper only a short summary of main results is presented. More details can be found in the other technical papers in this special issue. 4.1. Advanced reactor core thermal hydraulics Thermal hydraulic behavior, mainly pressure drop and heat transfer, in 19 rod bundles with grid spacers cooled by LBE was experimentally investigated at KIT. Based on test data, correlations for pressure loss coefficient and Nusselt number were derived and compared with other correlations available for liquid metal and numerical simulations performed by the project partners. Related to turbulence modeling, a look-up table approach was developed to dynamically determine the local turbulent Prandtl number in liquid metal flows. Fig. 14 shows the calculated distribution of turbulent Prandtl number in a 19-rod bundle. As indicated turbulent Prandtl number has high values near the wall and decreases approaching the center of subchannels. A summary of the rod bundle related work, both experimental and numerical, can be found in Pacio et al. (2014). Thermal-hydraulic behavior of gas cooled fuel assemblies has been studied at the gas test loop L-STAR/SL at KIT. Test data, including heat transfer, pressure drop and flow structures on the sub-texture scale measured with LDA have been collected in the range of 3 × 103 < Re < 1 × 105 and different rod surfaces, i.e. smooth surface and ribbed surface (Gómez et al., 2014). Data obtained were made available for the project partners for validating their computational models (Gómez et al., 2014). For the case with
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Fig. 13. Technical consortium of the THINS project with 24 partners.
smooth rod surface a good agreement between numerical results and experimental data is obtained. The importance of the selection of turbulence models was clearly pointed out. Thermal-hydraulics in pebble bed was numerically investigated. The available pebble bed data by TEES are made available for the project partners. Furthermore, LES approach was applied to study two pebble bed configurations, a single cubic pebble bed and a limited sized periodic random pebble. Results indicate that non-isotropic turbulence model is needed for reliable prediction of macroscopic properties. Simulation of heat transfer in wallbounded flow is difficult due to unknown heat flux from bed center to wall. More detailed results are summarized in Shams et al. (2013a,b). 4.2. Single phase mixed convection Experiments were carried out at the CIRCE facility simulating decay heat removal (DHR) of pool-type large scale system cooled by LBE. Particular emphasis was devoted to investigate (1) heat transfer in the 37-pin bundle of the heated section, (2) mixed and natural
convection and stratification phenomena in the LBE pool, (3) transition from forced to natural circulation in the primary circuit, (4) start-up and stabilization of natural circulation of LBE through the DHR heat exchanger, and (5) capability of the DHR system. For the simulation of system dynamics in buoyancy influenced flow conditions, post-test analysis of the most DHR tests has been performed with the RELAP5 system code and the 2D SIMMER-III code for validation purposes (Bandini et al., 2014). Furthermore, results from the previous tests performed in the TALL loop have been used for RELAP5, TRACE and CATHARE validation. All simulations confirm the rather good capability of system codes to deal with HLM fluid-dynamics under both forced and natural convection regimes. Limitations are evidenced in the modeling of mixing and stratification phenomena in a large pool, as observed in the analysis of CIRCE tests. Fig. 15 compares the measured temperature distribution in the pool along elevation height with the numerical results computed with both RELAP5 and SIMMER-III codes. It seems that the thermal stratification is well predicted by the SIMMER-III code, whereas the system analysis program RELAP5 cannot accurately reproduce the thermal stratification mode.
Fig. 14. Turbulent Prandtl number in rod bundles (Boettcher, 2013).
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Fig. 15. Temperature distribution in LBE pool: comparison of test data with numerical results.
Different system codes such as CATHARE, ATHLET and DYN2B have been tested using the results of the ultimate natural circulation experiment conducted in PHENIX reactor (Pialla et al., 2014). They give rather good prediction of the different phases of the experimental transients. Some limitations in system codes for SFR reactor calculations have been identified: (1) core meshing is not able to predict strong local temperature stratification, (2) stratification predicted by system codes is linked to topologic modeling choices, (3) it is difficult for system codes to predict properly 3D phenomena in complex geometry. Therefore, the interest of coupling system and CFD codes to predict complex 3D phenomena is reinforced.
Fig. 16. Profile of temperature (T+ = T/T ) at Re = 395 and Pr = 0.025 (Shama, 2014). AHFM-2005 and AHFM-cc: previous AHFM in the open literature.
dealing with the mixing of different density gases in a rectangular channel has been performed at the HOMER facility and an experiment in a more complex geometry of a small mixing plenum using supercritical fluids (SCMix) is underway to support the development and validation of approaches for innovative nuclear systems. Besides that, direct numerical simulations of conjugate heat transfer on temperature fluctuation in liquid metal was performed and used to evaluate the application of different wall treatment techniques in an LES framework. Because of some delay in the experimental studies, test data analysis and model development will be continued also after the termination of the THINS project.
4.3. Single phase turbulence 4.4. Multiphase flow For modeling turbulent heat transfer, the current engineering tools apply statistical turbulence closures and adopt the concept of the turbulent Prandtl number based on the Reynolds analogy. However, in cases of liquid metal or supercritical fluid flows, the turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. In the frame of the project, DNS approach has been applied to produce new reference data to support the development and validation of turbulence models for computational approaches in commonly used engineering CFD tools. DNS reference data for liquid metals was generated for channel flows and a wavy wall channel at different Reynolds and Prandtl numbers. The reference data were applied to develop and validate new Reynolds Averaged Navier Stokes (RANS) models, which then were implemented in engineering codes and their evaluations have shown promising results. With respect to RANS turbulence models, the most promising engineering approach developed is the algebraic heat flux model as developed at NRG (Shama, 2014). This model covers all flow regimes and the required range of Prandtl number to allow modeling liquid metals, and was successfully tested for mixed convection conditions. Fig. 16 compares the calculated temperature distribution using different AHFM models. It show that, for the case considered the present AHFM of NRG gives the best agreement with the DNS data. Further studies are required to extend its applicability to other flow regimes. The mixed law-of-the-wall method and the method using look-up tables have been considered as successful (work-around) for the application to liquid metal flows in the forced convection regime. With respect to temperature fluctuations in innovative nuclear systems which possibly lead to thermal fatigue damage, approaches developed for current LWRs are adapted and applied to innovative reactors. To this purpose, a fundamental experiment
Experiments on the highly turbulent free-surface flow in the funnel-shaped windowless spallation target, as was designed for the MYRRHA accelerator driven system, have been performed at different flow rates at KIT. High-speed imaging was used to provide information on surface shape and height of central recirculation zone. Numerical activities focussed on the development of various free-surface models able to simulate the sharp fluid-vacuum interface. In addition, suitable near-interface turbulence models were proposed. With the advanced methodology, the feasibility of numerical simulation for free surface flow in windowless spallation target is proven (Batta et al., 2014). The LBE/water interaction that occurs in case of steam generator leak in a LBE cooled reactor, has been investigated experimentally at different injection pressures in the LIFUS facility to provide reference data for numerical validation. The numerical code SIMMER-III is used to predict the energy released in the interaction volume during the HLM/water interaction. Furthermore, to analyze the system dynamic response in terms of displacement of and stress field in the structures, fluid–structure interaction calculations were performed by coupling CFD with structural analysis using the ANSYS code and the LS-DYNA code. Good agreement between the numerical results and experimental data have been achieved (Pesetti et al., 2014). Single and multilayer particle deposition and resuspension experiments in turbulent channel flows were performed at HZDR. Monodisperse and polydisperse particles were injected into the turbulent flow field and the particle deposition on the channel floor was investigated. Different channel surfaces (smooth and ribroughened) have been used. The experimental results provide a data base for CFD code development of single and multilayer particle deposition/resuspension in turbulent flows. Fig. 17 shows the measured particle deposition velocities at four different fluid
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Fig. 17. Measure non-dimensional particle deposition velocity, comparison of the present test data with CFD results and Wood’s model.
friction velocities, compared to two different CFD approaches (one by Lecrivain & Hampel and the other by Dehli) and the model of Wood. Generally, the test results can be well predicted by Wood’s correlation, which matches accurately the CFD results (Barth et al., 2014). 4.5. Code coupling and qualification The thermal-hydraulics modeling of light-water reactors for safety analysis has traditionally relied on system codes such as ATHLET, CATHARE and RELAP. However, the 0D/1D models featured in these codes make it hard to represent the 3D effects that can occur in the large plena of LMR during accidental transient. Therefore, coupling of system codes with CFD codes will enable a reliable description of important phenomena involved. Different coupling methodologies have been developed by the project partners, which differ from each other on two fundamental aspects: • The choice of the computing domains of the system codes and CFD codes. In the domain overlapping approach chosen by CEA and KTH, the system code computes the entire computational domain of the simulation. Based on CFD simulation results, which include region boundaries, the calculations of system codes are corrected. By contrast, in the domain decomposition approach used by GRS, TUM, and KIT, the system code and CFD subdomains only
intersect at inlet/outlet sections, so that the two codes exchange data via inlet/outlet boundary conditions. • The choice of the time frequency with which data is exchanged. In the coupling methodology developed at KIT, an initial system code standalone calculation is performed. After then, each code computes the entire transient, using the results of the previous iteration of the other code as boundary conditions. In contrast, in the methods developed at CEA, KTH, and GRS/TUM, both codes run once, but concurrently, with data exchanges taking place at each time step. In order to validate these approaches of code coupling, two sources of experimental data obtained in PHENIX reactor and at the TALL-3D facility were used. The results obtained in the PHENIX case have confirmed the capability of coupled calculations to improve on the predictions of system codes (Pialla et al., 2014), both by taking into account the consequences of 3D effects on the global dynamics and by allowing for the computation of the local temperature at the exact locations of measurement thermocouples. The TALL-3D facility has completed its commissioning tests (Grishcheko, 2014) and is currently gathering experimental data. Fig. 18 shows the mass flow rate and the temperature distribution inside the 3-D pool. Before the beginning of transient a steady state was established. At the reference time point, t = 0, the pump
Fig. 18. Transient results of TALL-3D test.
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was then shut down and the transition from forced circulation to natural circulation started. As seen, during the transition, mass flow rate decreases strongly and the local fluid temperature reaches maximum values after a certain time delay compared to the minimum value of mass flow rate. A new steady state is established after about 30 min. Thermal stratification in the 3D pool is clearly recognized. These results are well suitable for the validation of CFD codes as well as the coupling methodology of CFD with system analysis codes. 5. Summary Several crosscutting thermal-hydraulic issues of various innovative nuclear systems have been identified and are the subject of the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project in which 23 partners from Europe and one partner from US collaborate. The top target of the project is the establishment of experimental data base and the development of new and more accurate physical models and numerical simulation tools. To achieve the targets twelve experimental studies are included in this project and provide test data. A large number of numerical codes of different spatial classes are selected and numerical activities are coordinated. Numerical simulation provides support to experiments, and receives test data for the improvement of numerical simulation tools. Four years after the start of the THINS project, extensive results have been achieved. Except the E-SCAPE facility, all other 11 test facilities produced test data related to various crosscutting phenomena. A large test data base is thus established for model development and code validation. New correlations or models were developed such as heat transfer coefficient, friction factor, turbulence Prandtl number and algebraic heat flux model. They were (partially) validated based on the established test data base and also implemented in numerical codes. In general, the main objectives of the THINS project are well fulfilled. More detailed results of the THINS project are described in the other papers of this special issue. Acknowledgements The authors gratefully acknowledge the contributions of all colleagues involved in the THINS project. This work is supported by the 7th Framework Program European Commission Project THINS No. FP7-249337.
References Bandini, G., et al., 2014. RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-011. Barth, T., Lecrivain, G., Jayaraju, S.T., Hampel, U., 2014. Particle deposition and resuspension in gas-cooled reactors – activity overview of the two European research projects THINS and ARCHER. Nucl. Eng. Des., in this special issue. Batta, A., Class, A., Litfin, K., Wetzel, T., Moreau, V., Massidda, L., Thomas, S., Lakehal, D., Angeli, D., Losi, G., Mooney, G., Van Tichelen, K., 2014. Experimental and numerical investigation of liquid metal free-surface flows in spallation targets. Nucl. Eng. Des., in this special issue. Boettcher, M., 2013. CFD investigation of LBE rod bundle flow. Web Journal The Connector (September), http://www.pointwise.com/theconnector/ Gauthé, P., et al., 2012. The PHENIX ultimate natural convection test. In: Proceedings of ICAPP’12, June 2012, Paper 12323. Gómez, R., et al., 2014. Experimental results for heated rod in the L-STAR facility. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-030. Grishcheko, D., 2014. Design and commissioning tests of the TALL-3D experimental facility for validation of coupled STH and CFD codes. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-041. Hassan, Y., Dominguez-Ontiveros, E.E., 2008. Flow visualization in a pebble bed reactor experiment using PIV and refractive index matching techniques. Nucl. Eng. Des. 238 (2008), 3080–3085. IAEA, August 2003. Emerging Nuclear Energy and Transmutation Systems: Core Physics and Engineering Aspects, IAEA-TECDOC-1356. International Atomic Energy Agency. Maciocco, L., December 2002. CFD Simulation of a Free-Surface Isothermal Water Flow in the MYRRHA Target Geometry (MYRRHA Benchmark), CRS4 Tech. Rep. Pacio, J., et al., 2014. Heavy-liquid metal heat transfer experiment in a 19-rod bundle with grid spacers, NED paper. Nucl. Eng. Des. 273, 33–46. Pesetti, A., Del Nevo, A., Forione, N., 2014. Experimental investigation and SIMMER III code modelling of LBE–water interaction in LIFUS5/Mod2 facility. Nucl. Eng. Des., in this special issue. Pialla, D., et al., 2014. Overview of the system alone and system/CFD coupled calculations of the PHENIX natural circulation test within the THINS project. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-010. Roelofs, F., Shams, A., Otic, I., Böttcher, M., Duponcheel, M., Bartosiewicz, Y., Lakehal, D., Baglietto, E., Lardeau, S., Cheng, X., 2014. Status and perspective of turbulent heat transfer modelling for the industrial application of liquid metal flows. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-009. Shams, A., Roelofs, F., Komen, E.M.J., Baglietto, E., 2013a. Quasi-direct numerical simulation of a pebble bed configuration. Part I: Flow (velocity) field analysis. Nucl. Eng. Des. 263, 473–489. Shams, A., Roelofs, F., Komen, E.M.J., Baglietto, E., 2013b. Quasi-direct numerical simulation of a pebble bed configuration. Part I: Thermal (temperature) field analysis. Nucl. Eng. Des. 263, 490–499. Shama, A., 2014. An algebraic heat flux model for low Prandtl fluids. In: Proc. of THINS 2014 International Workshop, Modena, Italy, January 20–22, paper-002. THIRS, 2008. Proceedings of the International Workshop on Thermal-Hydraulics of Innovative Reactor and Transmutation Systems (THIRS), April 14–16, 2008. Karlsruhe, Germany, compiled by X. Cheng. USDOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002. A Technology Roadmap for Generation IV Nuclear Energy Systems, GIF-002-00.
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems X. Cheng a,∗ , A. Batta a , G. Bandini b , F. Roelofs c , K. Van Tichelen d , A. Gerschenfeld e , M. Prasser f , A. Papukchiev g , U. Hampel h , W.M. Ma i a
Karlsruhe Institute of Technology (KIT), Germany Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy c Nuclear Research and Consultancy Group (NRG), The Netherlands d Studiecentrum voor Kernenergie – Centre d’étude de l’Energie Nucléaire (SCK·CEN), Belgium e Commissariat à l’Energie Atomique (CEA), France f Paul Scherrer Institute (PSI), Switzerland g Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Germany h Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR), Germany i Kungliga Tekniska Högskolan (KTH), Sweden b
h i g h l i g h t s • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized.
a r t i c l e
i n f o
Article history: Received 27 October 2014 Accepted 6 November 2014
a b s t r a c t Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are:
• Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena.
• Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena.
• Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue. © 2014 Elsevier B.V. All rights reserved.
Abbreviations: ADS, accelerator-driven subcritical nuclear systems; CFD, computational fluid dynamics; DHR, decay heat removal; DNS, direct numerical simulation; GFR, gas-cooled fast reactor; GIF, Gen-IV International Forum; HLM, heavy liquid metal; HTR, high temperature reactor; LBE, lead bismuth eutectic; LDA, laser Doppler anemometry; LES, large eddy simulation; LFR, lead-cooled fast reactor; LMR, liquid metal reactors; MRI, matching refractive index; MSR, molten salt reactor; PTV, particle tracking velocimetry; RANS, Reynolds-averaged Navier–Stokes equations; SC, supercritical; SCWR, supercritical water cooled reactor; SFR, sodium-cooled fast reactor; THINS, thermal-hydraulics of innovative nuclear systems; VHTR, very high temperature reactor. ∗ Corresponding author. E-mail address: [email protected] (X. Cheng). http://dx.doi.org/10.1016/j.nucengdes.2014.11.007 0029-5493/© 2014 Elsevier B.V. All rights reserved.
X. Cheng et al. / Nuclear Engineering and Design 290 (2015) 2–12
1. Introduction For the long-term development of nuclear power, reactors of generation IV (Gen-IV) with enhanced safety, economics, sustainability and non-proliferation features need to be developed. The Gen-IV International Forum (GIF) recommended six innovative nuclear energy system for meeting future energy challenges and proposed a technology roadmap for Gen-IV nuclear energy systems (USDOE, 2002). These six innovative nuclear energy systems are very high temperature reactor (VHTR), gas-cooled fast reactor (GFR), sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), supercritical water cooled reactor (SCWR) and molten salt reactor (MSR). One aspect of sustainability is the reduction of nuclear waste. Nowadays, management of nuclear waste becomes a key issue in the public acceptance of nuclear energy. Extensive studies in the last years showed that partitioning and transmutation is a promising approach of the waste management. Transmutation using accelerator-driven subcritical nuclear systems (ADS) has attracted strong attention worldwide due to its claimed favorable safety features and its high incineration rate of nuclear waste (IAEA, 2003). Thermal-hydraulics is recognized as a key scientific subject in the development of the different innovative reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by coolants, flow channel structures formed by different fuel lattice arrangements and primary circuit lay-out (pool vs. loop). They result in different micro- and macroscopic behavior of flow and heat transfer and require specific models and advanced analysis tools. On the other hand, many common thermal-hydraulic issues are identified among various innovative nuclear systems (THIRS, 2008). In 2010, the EU project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) was launched (www.ifrt.kit.edu/thins/), which focuses on several crosscutting issues and synergizes infrastructure for the thermal-hydraulic research of the innovative nuclear systems in Europe. The overall objectives of the THINS project are the development and validation of new physical models and qualification of numerical analysis tools. Specific objectives are:
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based on advanced physical models and numerical methodology. Coupling of the code solutions at various scales and the qualification of the coupled calculations extend the applicability and ensure the reliability of the numerical platform. The THINS project has a duration of five years and will terminate in January 2015. In January 2014 the International Workshop on Thermal-Hydraulics of Innovative Nuclear systems, THINS-2014, took place at the University of Modena, Italy (http://www.thins2014.unimore.it) with the aim to review the project progress, to disseminate the main achievements of the project among the international nuclear community and to foster the exchange of the newest scientific and technical results between researchers and experts. In this paper, the main objectives, technical tasks, scientific methodology and some main results are presented, in order to give an introductory overview of the THINS project and provide guidance to this special journal issue. 2. Technical tasks The THINS project is structured with the following six work packages (WPs): • • • • • •
WP1: Advanced reactor core thermal-hydraulics. WP2: Single phase mixed convection. WP3: Single phase turbulence. WP4: Multi-phase flow. WP5: Code coupling and qualification. WP6: Education and training.
The first five WPs are devoted to the individual crosscutting issues. Experiments are foreseen to provide experimental evidence and fundamental test data base. New physical models are developed and further employed to improve codes. The last WP is devoted to the education and training of young nuclear engineers and researchers. 2.1. Advanced reactor core thermal-hydraulics
• Establishment of a data base for the development and validation of new physical models and numerical codes for a more accurate description of the selected crosscutting thermal-hydraulic phenomena. Generic experiments are performed in the THINS project to produce a comprehensive data base for the validation purpose. In addition, direct numerical simulation (DNS) will be performed to provide numerical data base, which is of crucial importance for the development of turbulence models. With this project, a data base will be established for the fundamental thermal-hydraulic issues occurring in the innovative nuclear systems. • Establishment of an experimental platform for the thermalhydraulic research of the innovative nuclear systems. In the past, also in the frame of the previous European projects, experimental facilities, specifically for the thermal-hydraulics of innovative nuclear systems, have been built and operated at various institutions (http://vella.brasimone.enea.it/na.htm). The THINS project will make the optimum utilization of the available European experimental facilities and expertise, combine the resources available and establish a European experimental platform. • Establishment of a numerical platform for the design analysis of the innovative nuclear systems. Numerical codes for nuclear thermal-hydraulics cover various classes of spatial scales, i.e. system analysis based mainly on lumped parameter approach, sub-channel analysis specifically for fuel assembly and reactor core and CFD codes for detailed local flow behavior. With the THINS project, more reliable and validated codes will be proposed
Design of innovative reactor cores requires detailed analysis of the thermal-hydraulics within the fuel assemblies with high resolution inside individual sub-channels. The goal of this work package is to provide validated and verified simulation tools of the coolant flow within the reactor core components for typical states encountered in liquid metal cooled fuel assembly. The THINS project considers different numerical approaches applicable to the design analysis of innovative reactor cores. The first approach is based on sub-channel analysis codes. Efforts are made to introduce advanced numerical methods, e.g. coarse grid simulation, and to improve the physical models suitable for specific conditions of innovative reactor cores. The purpose of the second approach is to develop CFD simulation tools with more advanced models than the current state-of-the-art RANS modeling and improved accuracy. The complex nature of heat transfer in pebble bed requires the development of new algorithm techniques to describe the core, as well as robust methods capable to treat convective–conductive–radiative heat transfer with fluid flow across complex structure of pebble bed. In the THINS project a method is proposed based on macroscopic modeling of flow phenomena in pebble bed reactors. The macroscopic properties are obtained by LES simulation technique applied to a geometry consisting of spheres in an array with irregular structure. Two experiments are considered in this WP to provide test data for the validation purpose, as shown in Figs. 1 and 2. Test data in fuel assemblies with liquid metal for mixed and natural convection
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Fig. 1. Rod bundle tests with LBE in KIT.
conditions will be produced with the rod bundle test section, see Fig. 1, performed in the KALLA laboratory of KIT. Heat transfer tests on single rod with textured surface, see Fig. 2, are performed with local temperature measurement in the gas and on the heated rod surface along with LDA to provide experimental data on integral and on the sub-texture scale. The third test data set was obtained in a pebble bed configuration simulating a “fuel compact” in the HTR/VHTR (Hassan and Dominguez-Ontiveros, 2008). Velocity fields inside the packed bed are measured using Particle Tracking Velocimetry (PTV) combined with the Matching Refractive Index (MRI) methods. 2.2. Single phase mixed convection The use of the natural convection to remove residual heat from the core is widely adopted in the Gen-IV reactor concepts that implement passive safety features to enhance reliability. Correct prediction of the onset and stability of the natural convection is essential for the design of the decay heat removal system and for the transient analysis. In addition, the pool-type configuration proposed for most of the innovative Liquid Metal Reactors (LMR) to compact the primary system and to limit the risk of coolant leakage requires numerical tools able to simulate local 3-D phenomena, such as thermal stratification, for the characterization of the
Fig. 3. The PHENIX reactor in France.
thermal loads on the primary vessel and components as well as for the design of in-pool heat exchangers. In this WP, simulation of dynamic behavior of the innovative nuclear systems is mainly based on system analysis codes originally developed for LWR. The applicability of these codes to support the design process and safety analysis of LMR will be extended and validated. Main extension is focused on the implementation of closure relationships suitable for innovative cooling fluids and for a large spectrum of flow conditions, from forced convection to natural convection. For the simulation of local 3-D phenomena, CFD approach is applied. The qualification of the CFD models on the separate effect tests and the validation of the system analysis codes on integral tests represent one of the main tasks. To achieve this target, four experiments are planned, one with sodium and the other three with LBE. One of the four test facilities ESCAPE is being constructed in the frame of the THINS project. However, its operation will start after the termination of the THINS project. The other three test facilities will provide test data and are presented in Figs. 3–5. The PHENIX end-of-life tests (Gauthé et al., 2012) represent a real reactor configuration and contain complex effects of prototypical systems, see Fig. 3. Due to the limitation in measurement instrumentation, however, data available for validation are restricted, e.g. to system dynamic behavior. In the CIRCE (Fig. 4) and TALL-3D (Fig. 5) test facilities, more sophisticated measurements are applicable. Test data for the qualification for both system codes and 3-D CFD codes are expected. Main focus of all three experiments is on flows at strong buoyancy effect, i.e. mixed convection conditions.
2.3. Single phase turbulence
Fig. 2. L-STAR tests with air in KIT.
For the coolants envisaged in the innovative nuclear reactors, usually experiments are very expensive and accurate measurements are challenging or even impossible. Therefore, application of CFD for prediction of various flow characteristics becomes an
X. Cheng et al. / Nuclear Engineering and Design 290 (2015) 2–12
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Fig. 6. DeLight facility with SC Freon in DUT.
Fig. 4. CIRCE facility with LBE in ENEA.
attractive and complementary practice used in the design and evaluation process of innovative nuclear reactors. It is well agreed (Roelofs et al., 2014) that one of the key issues ensuring a reliable CFD simulation is the modeling of turbulence. For innovative nuclear systems, two features are important and need to be considered in the turbulence modeling.
(i) The coolants envisaged for advanced reactor systems cover a wide range of fluids with various physical properties, e.g. the molecular Prandtl number varies from the order of 10−3 –103 . This implies the specific behavior and prediction of turbulence and represents a challenging task. (ii) At normal operating conditions, the fluids in all innovative nuclear systems considered are at single-phase flow conditions. Therefore, this WP is devoted to single-phase turbulence with the main objective to improve and to develop turbulence models for non-unity Prandtl number flows, their implementation in engineering tools and application to liquid metal and supercritical flows. Related to innovative nuclear reactors with operation at elevated temperature, this WP focuses on two main issues, i.e. the strong buoyancy effect on turbulence and the occurrence of temperature fluctuation which may lead to thermal fatigue. A large number of CFD codes are applied (see Section 3). For validation of turbulence models and qualification of CFD codes, three experiments are conducted, as illustrated in Figs. 6–8. Both Delight (Fig. 6) and SCMix (Fig. 7) test facilities uses supercritical (SC) Freon as working fluids. In DeLight facility heat transfer at SC Freon will be investigated with the main emphasis on the investigation of mechanism of heat transfer deterioration. The main purpose of SCMix experiments is the mixing behavior of supercritical fluids under strong density variation and buoyancy effect. The HOMER facility (Fig. 8) is designed for study mixing process of two gas components with different densities. 2.4. Multi-phase flow
Fig. 5. TALL facility with LBE in KTH.
Multi-phase flows are encountered in several innovative reactor systems. A sample of these flows forms the basis of the work in this WP. Free surface flows are present in the pool-type liquid metal cooled reactors (LMR). They are also key phenomena in the accelerator driven systems (ADS) with windowless spallation targets. Bubbly flows occur from the possible incidental interaction between water and heavy liquid metal (HLM) in lead alloy cooled reactor systems. In HTR/VHTR, graphite dust is generated by abrasion and transported in the coolant loop.
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Fig. 7. SCMix facility with SC Freon in DUT.
In this WP advanced free surface modeling methods, such as the ALE-Moving Mesh Algorithm (Maciocco, 2002), will be further developed and validated which are capable of capturing sharp fluid-vacuum interfaces and simulating near-interface subscale turbulence. Concerning the HLM/water interaction, mainly the SIMMER-III code is used to investigate the related phenomena. Complete simulation of the entire HLM/water interaction procedure requires the implementation of additional modules predicting energy release and fluid–structure interaction. For the safety analysis of HTR/VHTR, transportation behavior of contaminated graphite dust generated by abrasion of fuel plays an important role. Development of appropriate CFD models for dust transport simulations is then one of the main tasks of this WP. To provide test data for model development and numerical codes qualification, three experiments are implemented in this WP, as summarized in Figs. 9–11. The free surface test at KIT using LBE investigates the characteristics of free surface and the effect of various parameters. The LIFUS facility at ENEA is devoted to study the interaction of LBE with water. It focuses on the measurement of energy release during the LBE/water interaction. At the GPLoop facility at HZDR deposition and resuspension of particles at various flow and surface conditions will be investigated.
Fig. 9. Free surface tests with LBE in KIT.
qualification towards two scenarios entirely dedicated to innovative nuclear systems, i.e. LMR (sodium cooled, lead alloy cooled) with buoyancy effects and transition from forced convection to natural convection. Two complementary experiments are taken into consideration in the THINS project to provide test data for the qualification
2.5. Code coupling and qualification The overall objective of the work in this WP is the development and qualification of new multi-scale computational solutions applicable to innovative nuclear systems. A coupled multi-scale solution will combine the merits of various classes of codes and provide design tools with feasible efficiency and high accuracy. This task is devoted to the coupling of system codes with CFD codes and its
Fig. 8. HOMER facility with gas mixture in ETH.
Fig. 10. LIFUS facility in ENEA.
X. Cheng et al. / Nuclear Engineering and Design 290 (2015) 2–12
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3. Methodology applied 3.1. Technical approach
Fig. 11. GPLoop facility for particle deposition and resuspension in HZDR.
of the code coupling solution. The PHENIX reactor scale experiment (Fig. 3), with existing ultimate test data on a full scale safety issue scenario, and the HLM TALL-3D (Fig. 5) integral scale experiment on a separate effect scale safety issue scenario will produce a unique test data base, which is at the same time complementary to the existing data base for V&V of code coupling solutions.
2.6. Education and training More than 50 young researchers (Master students, PhD students and Post-Doc researchers) are directly involved in the THINS project. Most of the theses are mainly based on the THINS activities. In addition to project meetings, cluster workshops and training courses were organized for young participants. At the cluster workshops students’ presentations and invited topical lectures were integrated in the workshop program. This gave the participating students the opportunity to follow all scientific presentations focusing on the THINS project progress. The topical lectures went far beyond the tasks pursued within the THINS project and made the audience familiar with particular new developments in the field of nuclear reactor technology. Special training courses were provided to young participants related to the application of CFD codes and to the coupling methodology of CFD codes with system codes.
The methodology used in this project is indicated in Fig. 12. Based on phenomenological analysis of the identified crosscutting phenomena, experimental programs were defined with the purpose to provide experimental evidence to a better understanding of physical phenomena and test data for the development of models and validation of codes. Totally 12 experiments have been defined. Due to limited possibility of experiments with low Prandtl number fluids, direct numerical simulation (DNS) will be carried out, to provide additional data for the data base. In parallel, numerical investigations are carried out with three different scales of codes, i.e. system analysis codes, sub-channel analysis codes and CFD codes. Numerical support will be provided to experimental work in various phases, such as selection of measurement techniques, design of test facilities, definition of test matrix and test data analysis. The data base consisting of both experimental and numerical results can be then applied for the development of physical models and validation of simulation codes. The models considered in this project are for both system analysis codes and CFD codes, e.g. heat transfer and friction pressure drop in liquid metal for system analysis codes, and turbulence models for CFD simulations. To enhance the interaction between the work packages, three Clusters are introduced in the project management structure, to enhance the interaction between the work packages in which similar or related experimental or numerical approaches may be employed. Therefore, the clusters strongly support the realization of the project objectives. Three clusters have been established, i.e. (i) CFD model and validation; (ii) system code and validation; (iii) experimental techniques and data bases. The main tasks of clusters are to distribute important information between WPs which might be of general interest for the THINS partners, dealing with experiments, CFD or system code simulations, to organize cluster workshops and to organize required benchmarks. 3.2. Experimental activities Experimental studies are important tools and also necessary boundary conditions to achieve the main target of the project, i.e. development of models and more reliable simulation codes. Twelve experiments have been included in the THINS project and are summarized in Table 1. Real reactor data for sodium as a coolant is provided from the PHENIX end of life test. Six experiments use LBE
Fig. 12. Technical approach applied.
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Table 1 Summary of experimental studies. No.
Table 2 Summary of numerical codes used.
Name of test
Fluids
Main phenomena
1
PHENIX
Na
2
Rod bundle
LBE
3 4
E-SCAPE CIRCE
LBE LBE
TALL-3D Free surface LIFUS L-Star GPLoop DeLight SCMix HOMER
LBE LBE LBE and water Air Air and particles SC water SC water He and air
System dynamics, pool mixing Flow and heat transfer behavior Pool mixing Pool mixing, heat transfer in bundle Natural circulation Free surface shape LBE/H2 O interaction Local heat transfer Particle transportation Heat transfer Thermal mixing Mixing of gases
5 6 7 8 9 10 11 12
Figures 3
Codes CFD ANSYS-CFX
1 – 4 5 9 10 2 11 6 7 8
Star-CD/Star-CCM+ Fluent OpenFOAM TransAT SIMMER-III Armando THEMAT FLUIDITY System analysis RELAP5 ATHLET
as fluid. They are dealing with several crosscutting phenomena of LMR, e.g. mixed convection in large pool, mixed heat transfer in rod bundle, HLM/water interaction and free surface flow. It aims at validating system codes, CFD-codes and coupled system/CFD solutions. One experiment is devoted to heat transfer in gas cooled reactor cores with prismatic type fuel structure, whereas another experiment related to gas cooled reactor is dealing with graphite dust behavior, which is of crucial importance for the safety of HTR/VHTR. Two experiments are designed for supercritical water cooled reactors. In the experiment DeLight, heat transfer in single tubes under SCWR conditions will be carried out with the emphasis on heat transfer deterioration and its influencing parameters. The experiment SCMix deals again with the mixing behavior under strong buoyancy effect. At the facility HOMER, buoyancy driven mixing behavior will be investigated using gases of different density. Related to crosscutting thermal-hydraulic issues it is easily recognized that flow and heat transfer under strong buoyancy effect (mixed convection) is the key topic in this project. It is also important phenomenon in various kinds of innovative nuclear systems, as LMR, GCR and SCWR.
3.3. Numerical activities The final goal of the project is the development of advanced physical models, numerical simulation tools and their validation on experimental data. There is strong interaction between numerical activities and experimental studies. As mentioned in Fig. 12, numerical activities should in one hand provide support to experimental work and in the other hand utilize the experimental results to improve the numerical tools. In this project a large number of numerical codes are used, which can be divided into three categories, according to their spatial resolution. Table 2 summarizes the numerical codes used. In the CFD class, 11 codes are selected by the project partners. The focus is clearly on the widely used commercial codes like ANSYS-CFX, Star-CD/Star-CCM+, Fluent, TransAT, and the open accessible code OpenFOAM. However, for certain dedicated purposes also in-house codes like THEMAT, FLUIDITY and SIMMER-III are used. Each experiment is supported by several codes. Different codes might be used by the same partner for different experiments or even for the same experiment, to realize code-to-code comparison purpose. Five system analysis codes are selected, which are widely applied in the nuclear community. Key focus is clearly on the European codes, i.e. CATHARE from CEA, DYN2B from IRSN and ATHLET from GRS. For analyzing pool type LMR, coupling of system analysis code with CFD code will be realized, i.e. CATHARE with TRIO-U,
CATHARE TRACE DYN2B Sub-channel analysis Coarse mesh CFD
Related experiments TALL-3D, rod bundle, E-SCAPE, L-Star, free surface, LIFUS Rod bundle, TALL-3D, free surface, DeLight E-SCAPE, CIRCE, L-Star, DeLight, GPLoop Rod bundle, free surface, PHENIX, DeLight, HOMER, GPLoop Free surface LIFUS, CIRCE Free surface, with ALE-Moving Mesh Algorithm HOMER PGLoop, models for particle transportation CIRCE, E-SCAPE, TALL-3D, also coupling with Star-CCM+ TALL-3D, PHENIX, also coupling with OpenFOAM and ANSYS-CFX TALL-3D, PHENIX, also coupling with Trio-U TALL-3D, Rod bundle PHENIX Rod bundle
ATHLET with OpenFOAM, ATHLET with ANSYS-CFX and RELAP with Star-CCM+ . 3.4. Technical consortium The technical consortium of the project consists of 24 partners, of those 23 institutions from European Union and one partner from US, as illustrated in Fig. 13. The consortium includes nearly all relevant institutions in European Union involved in the nuclear thermal-hydraulics, especially dealing with crosscutting topics of thermal-hydraulics of innovative nuclear systems. 4. Summary of results achieved In this paper only a short summary of main results is presented. More details can be found in the other technical papers in this special issue. 4.1. Advanced reactor core thermal hydraulics Thermal hydraulic behavior, mainly pressure drop and heat transfer, in 19 rod bundles with grid spacers cooled by LBE was experimentally investigated at KIT. Based on test data, correlations for pressure loss coefficient and Nusselt number were derived and compared with other correlations available for liquid metal and numerical simulations performed by the project partners. Related to turbulence modeling, a look-up table approach was developed to dynamically determine the local turbulent Prandtl number in liquid metal flows. Fig. 14 shows the calculated distribution of turbulent Prandtl number in a 19-rod bundle. As indicated turbulent Prandtl number has high values near the wall and decreases approaching the center of subchannels. A summary of the rod bundle related work, both experimental and numerical, can be found in Pacio et al. (2014). Thermal-hydraulic behavior of gas cooled fuel assemblies has been studied at the gas test loop L-STAR/SL at KIT. Test data, including heat transfer, pressure drop and flow structures on the sub-texture scale measured with LDA have been collected in the range of 3 × 103 < Re < 1 × 105 and different rod surfaces, i.e. smooth surface and ribbed surface (Gómez et al., 2014). Data obtained were made available for the project partners for validating their computational models (Gómez et al., 2014). For the case with
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Fig. 13. Technical consortium of the THINS project with 24 partners.
smooth rod surface a good agreement between numerical results and experimental data is obtained. The importance of the selection of turbulence models was clearly pointed out. Thermal-hydraulics in pebble bed was numerically investigated. The available pebble bed data by TEES are made available for the project partners. Furthermore, LES approach was applied to study two pebble bed configurations, a single cubic pebble bed and a limited sized periodic random pebble. Results indicate that non-isotropic turbulence model is needed for reliable prediction of macroscopic properties. Simulation of heat transfer in wallbounded flow is difficult due to unknown heat flux from bed center to wall. More detailed results are summarized in Shams et al. (2013a,b). 4.2. Single phase mixed convection Experiments were carried out at the CIRCE facility simulating decay heat removal (DHR) of pool-type large scale system cooled by LBE. Particular emphasis was devoted to investigate (1) heat transfer in the 37-pin bundle of the heated section, (2) mixed and natural
convection and stratification phenomena in the LBE pool, (3) transition from forced to natural circulation in the primary circuit, (4) start-up and stabilization of natural circulation of LBE through the DHR heat exchanger, and (5) capability of the DHR system. For the simulation of system dynamics in buoyancy influenced flow conditions, post-test analysis of the most DHR tests has been performed with the RELAP5 system code and the 2D SIMMER-III code for validation purposes (Bandini et al., 2014). Furthermore, results from the previous tests performed in the TALL loop have been used for RELAP5, TRACE and CATHARE validation. All simulations confirm the rather good capability of system codes to deal with HLM fluid-dynamics under both forced and natural convection regimes. Limitations are evidenced in the modeling of mixing and stratification phenomena in a large pool, as observed in the analysis of CIRCE tests. Fig. 15 compares the measured temperature distribution in the pool along elevation height with the numerical results computed with both RELAP5 and SIMMER-III codes. It seems that the thermal stratification is well predicted by the SIMMER-III code, whereas the system analysis program RELAP5 cannot accurately reproduce the thermal stratification mode.
Fig. 14. Turbulent Prandtl number in rod bundles (Boettcher, 2013).
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Fig. 15. Temperature distribution in LBE pool: comparison of test data with numerical results.
Different system codes such as CATHARE, ATHLET and DYN2B have been tested using the results of the ultimate natural circulation experiment conducted in PHENIX reactor (Pialla et al., 2014). They give rather good prediction of the different phases of the experimental transients. Some limitations in system codes for SFR reactor calculations have been identified: (1) core meshing is not able to predict strong local temperature stratification, (2) stratification predicted by system codes is linked to topologic modeling choices, (3) it is difficult for system codes to predict properly 3D phenomena in complex geometry. Therefore, the interest of coupling system and CFD codes to predict complex 3D phenomena is reinforced.
Fig. 16. Profile of temperature (T+ = T/T ) at Re = 395 and Pr = 0.025 (Shama, 2014). AHFM-2005 and AHFM-cc: previous AHFM in the open literature.
dealing with the mixing of different density gases in a rectangular channel has been performed at the HOMER facility and an experiment in a more complex geometry of a small mixing plenum using supercritical fluids (SCMix) is underway to support the development and validation of approaches for innovative nuclear systems. Besides that, direct numerical simulations of conjugate heat transfer on temperature fluctuation in liquid metal was performed and used to evaluate the application of different wall treatment techniques in an LES framework. Because of some delay in the experimental studies, test data analysis and model development will be continued also after the termination of the THINS project.
4.3. Single phase turbulence 4.4. Multiphase flow For modeling turbulent heat transfer, the current engineering tools apply statistical turbulence closures and adopt the concept of the turbulent Prandtl number based on the Reynolds analogy. However, in cases of liquid metal or supercritical fluid flows, the turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. In the frame of the project, DNS approach has been applied to produce new reference data to support the development and validation of turbulence models for computational approaches in commonly used engineering CFD tools. DNS reference data for liquid metals was generated for channel flows and a wavy wall channel at different Reynolds and Prandtl numbers. The reference data were applied to develop and validate new Reynolds Averaged Navier Stokes (RANS) models, which then were implemented in engineering codes and their evaluations have shown promising results. With respect to RANS turbulence models, the most promising engineering approach developed is the algebraic heat flux model as developed at NRG (Shama, 2014). This model covers all flow regimes and the required range of Prandtl number to allow modeling liquid metals, and was successfully tested for mixed convection conditions. Fig. 16 compares the calculated temperature distribution using different AHFM models. It show that, for the case considered the present AHFM of NRG gives the best agreement with the DNS data. Further studies are required to extend its applicability to other flow regimes. The mixed law-of-the-wall method and the method using look-up tables have been considered as successful (work-around) for the application to liquid metal flows in the forced convection regime. With respect to temperature fluctuations in innovative nuclear systems which possibly lead to thermal fatigue damage, approaches developed for current LWRs are adapted and applied to innovative reactors. To this purpose, a fundamental experiment
Experiments on the highly turbulent free-surface flow in the funnel-shaped windowless spallation target, as was designed for the MYRRHA accelerator driven system, have been performed at different flow rates at KIT. High-speed imaging was used to provide information on surface shape and height of central recirculation zone. Numerical activities focussed on the development of various free-surface models able to simulate the sharp fluid-vacuum interface. In addition, suitable near-interface turbulence models were proposed. With the advanced methodology, the feasibility of numerical simulation for free surface flow in windowless spallation target is proven (Batta et al., 2014). The LBE/water interaction that occurs in case of steam generator leak in a LBE cooled reactor, has been investigated experimentally at different injection pressures in the LIFUS facility to provide reference data for numerical validation. The numerical code SIMMER-III is used to predict the energy released in the interaction volume during the HLM/water interaction. Furthermore, to analyze the system dynamic response in terms of displacement of and stress field in the structures, fluid–structure interaction calculations were performed by coupling CFD with structural analysis using the ANSYS code and the LS-DYNA code. Good agreement between the numerical results and experimental data have been achieved (Pesetti et al., 2014). Single and multilayer particle deposition and resuspension experiments in turbulent channel flows were performed at HZDR. Monodisperse and polydisperse particles were injected into the turbulent flow field and the particle deposition on the channel floor was investigated. Different channel surfaces (smooth and ribroughened) have been used. The experimental results provide a data base for CFD code development of single and multilayer particle deposition/resuspension in turbulent flows. Fig. 17 shows the measured particle deposition velocities at four different fluid
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Fig. 17. Measure non-dimensional particle deposition velocity, comparison of the present test data with CFD results and Wood’s model.
friction velocities, compared to two different CFD approaches (one by Lecrivain & Hampel and the other by Dehli) and the model of Wood. Generally, the test results can be well predicted by Wood’s correlation, which matches accurately the CFD results (Barth et al., 2014). 4.5. Code coupling and qualification The thermal-hydraulics modeling of light-water reactors for safety analysis has traditionally relied on system codes such as ATHLET, CATHARE and RELAP. However, the 0D/1D models featured in these codes make it hard to represent the 3D effects that can occur in the large plena of LMR during accidental transient. Therefore, coupling of system codes with CFD codes will enable a reliable description of important phenomena involved. Different coupling methodologies have been developed by the project partners, which differ from each other on two fundamental aspects: • The choice of the computing domains of the system codes and CFD codes. In the domain overlapping approach chosen by CEA and KTH, the system code computes the entire computational domain of the simulation. Based on CFD simulation results, which include region boundaries, the calculations of system codes are corrected. By contrast, in the domain decomposition approach used by GRS, TUM, and KIT, the system code and CFD subdomains only
intersect at inlet/outlet sections, so that the two codes exchange data via inlet/outlet boundary conditions. • The choice of the time frequency with which data is exchanged. In the coupling methodology developed at KIT, an initial system code standalone calculation is performed. After then, each code computes the entire transient, using the results of the previous iteration of the other code as boundary conditions. In contrast, in the methods developed at CEA, KTH, and GRS/TUM, both codes run once, but concurrently, with data exchanges taking place at each time step. In order to validate these approaches of code coupling, two sources of experimental data obtained in PHENIX reactor and at the TALL-3D facility were used. The results obtained in the PHENIX case have confirmed the capability of coupled calculations to improve on the predictions of system codes (Pialla et al., 2014), both by taking into account the consequences of 3D effects on the global dynamics and by allowing for the computation of the local temperature at the exact locations of measurement thermocouples. The TALL-3D facility has completed its commissioning tests (Grishcheko, 2014) and is currently gathering experimental data. Fig. 18 shows the mass flow rate and the temperature distribution inside the 3-D pool. Before the beginning of transient a steady state was established. At the reference time point, t = 0, the pump
Fig. 18. Transient results of TALL-3D test.
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was then shut down and the transition from forced circulation to natural circulation started. As seen, during the transition, mass flow rate decreases strongly and the local fluid temperature reaches maximum values after a certain time delay compared to the minimum value of mass flow rate. A new steady state is established after about 30 min. Thermal stratification in the 3D pool is clearly recognized. These results are well suitable for the validation of CFD codes as well as the coupling methodology of CFD with system analysis codes. 5. Summary Several crosscutting thermal-hydraulic issues of various innovative nuclear systems have been identified and are the subject of the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project in which 23 partners from Europe and one partner from US collaborate. The top target of the project is the establishment of experimental data base and the development of new and more accurate physical models and numerical simulation tools. To achieve the targets twelve experimental studies are included in this project and provide test data. A large number of numerical codes of different spatial classes are selected and numerical activities are coordinated. Numerical simulation provides support to experiments, and receives test data for the improvement of numerical simulation tools. Four years after the start of the THINS project, extensive results have been achieved. Except the E-SCAPE facility, all other 11 test facilities produced test data related to various crosscutting phenomena. A large test data base is thus established for model development and code validation. New correlations or models were developed such as heat transfer coefficient, friction factor, turbulence Prandtl number and algebraic heat flux model. They were (partially) validated based on the established test data base and also implemented in numerical codes. In general, the main objectives of the THINS project are well fulfilled. More detailed results of the THINS project are described in the other papers of this special issue. Acknowledgements The authors gratefully acknowledge the contributions of all colleagues involved in the THINS project. This work is supported by the 7th Framework Program European Commission Project THINS No. FP7-249337.
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