Current Projects
Innovation in Supercritical CO2 Power generation systems
The primary objective of this four-year work programme is to undertake cutting edge multidisciplinary research and development to make a step change in understanding of Supercritical CO2 based power generation systems’ technology and its potential to enable a step change in thermal energy power cycles to be a major contributor to achieving the 2050 zero emissions targets while providing specialised training for 15 doctoral researchers to help establish the backbone of an important industry. The technical objectives of this research are:
- Develop advanced models and design tools that enable the optimal integration of sCO2 power systems components for various thermal energy sources and end use applications
- Develop accurate prediction tools for the simulation of transient operation of sCO2 power cycles and investigate innovative concepts of control and optimisation of operation
- Develop innovative methods to enhance aerodynamic and mechanical performance, reliability, and operability of key system components
- Develop advanced modelling and experimental methods that enable selection and development of materials, coatings and manufacturing techniques
To achieve the objectives of this training programme effectively, ISOP proposes four research WPs and requests funding from the EU for 15 Doctoral Candidates for a total of 540 person months who will work on an ambitious plan to advance the sCO2 power cycles technology beyond the state-of-the-art. The project aims to contribute to the EU agenda on European Research Area by training “a new generation of creative, entrepreneurial and innovative early-stage researchers”, who can face future challenges and to “convert knowledge and ideas into products and services for economic and social benefit”. In addition, support to and compliance with the United Nation’s Sustainable Development Goals will be at the heart of the training of the doctoral candidates and the scientific and economic outcomes of this research.
IKE will specially work on the experimental investigation of pseudo-condensation close to the critical point of CO2 and on the numerical investigation of the mixing process in headers of supercritical CO2 heat exchangers.
https://www.linkedin.com/company/isopproject
Funded by: This Project is supported by the European Union’s Horizon Europe Research and Innovation Program under Grant Agreement no. 101073266
Duration: January 2023– December 2026
Contact: Prof. Dr.-Ing. J. Starflinger
Analysis of the incident and accident behavior of SMR with the system code AC²
Numerous neighboring countries are very interested in developing or building small modular reactors (SMRs). The reasons for this are the predicted lower investment costs per module and a potentially high level of safety due to the increased use of passive safety systems. Light water-based SMRs (LW-SMRs) of the integral pressurized water reactor (iDWR) type, including the NuScale and NUWARD concepts, have great potential for realization. It cannot be assumed that manufacturers and developers of SMR concepts will release their simulation methods. In order to be able to carry out reliable and independent simulations on the safety of these reactors with the national calculation chain (including AC²), it is therefore necessary to further improve models in AC² that are relevant for the simulation of SMRs. Furthermore, AC² plant data sets are required for the analysis of operating states as well as incidents and accidents in order to build up expertise on these plants and their safety concepts among the partners. The aforementioned points are to be developed by the partners PSS, GRS and IKE in the joint project applied for. The overall objective of the joint project is to improve AC² models and data sets for the simulation of LW-SMRs of the iDWR type and to increase the expertise and international networking of the joint partners for safety assessment and the operational, incident and accident behavior of these designs.The project thus addresses topics A2.2 and A2.3 of the funding strategy for nuclear safety research. The further development of AC² is part of the longterm development of methods for the safety assessment of nuclear facilities. The project also addresses the Federal Government's concept for the development of young scientists in nuclear technology. Two young scientists will be employed for the planned work, who will also have the opportunity to do a doctorate.
Funded by: Bundesministerium Umwelt, Naturschutz, nukleare Sicherheit und Verbraucherschutz
Duration: July 2024 – June 2027
Contact: Prof. Dr.-Ing. J. Starflinger
The overall objective of the joint project is the development and application of machine learning (ML) methods for the simulation of phenomena in the late accident phase. In an interdisciplinary approach, competences from institutes of nuclear technology (IKE and PSS) and computer science (MLSim) are to be combined.
The cooling of particle debris is of great importance for targeted accident mitigation. Numerical simulations are indispensable for estimating the effectiveness of measures. However, the multiphase processes occurring in the late phase are highly complex and can only be simulated realistically with great effort. The project will therefore pursue the innovative approach of providing extensive data by means of simulations of a validated detailed model (COCOMO-3D), which can be used to train and validate ML models that can reproduce essential results of the detailed model in a fraction of its computing time.
Experiments are to be carried out in the DEBRIS and FLOAT facilities, which will be used for extended validation and validation of the COCOMO-3D code. ML methods for pattern recognition will also be used to evaluate the experimental data. In close cooperation between nuclear engineering and computer science, ML methods are to be developed on the basis of "Physics-informed Neural Networks" and, based on these, fast-running ML models for quenching and the long-term coolability of debris beds are to be created. By integrating them into the AC2 programme system, its model basis will be expanded. The applicability of the developed models will be demonstrated by uncertainty analyses and validation studies with AC2. This will demonstrate the potential of ML methods for further applications in reactor safety research.
The work is designed as a doctorate in which young scientists are trained and thus also serves to expand competence.
Funded by: BMBF, funding code: 02NUK078
Duration: 01.03.2023 - 28.02.2027
Contact: Prof. Jörg Starflinger, Prof. Mathias Niepert, Cooperation with Prof. M.K. Koch, PSS, Ruhr-University Bochum
So-called Micro Modular Reactors (MMRs) are currently being developed with government funding in the USA, for example, to replace diesel generators in both civil and military contexts. MMRs can be transported both on conventional trucks and in aircraft. Transporting an MMR in the vicinity of Germany or through German airspace is a plausible scenario in the medium term. Relevant design variants of MMRs are listed in Annex A. Due to their low power, simple transportability and novel concepts, MMRs differ strongly from other SMRs such as the NuScale design, which has many similarities with newer pressurised water reactors. Up to now, there has been no in-depth research or in-house expertise on MMR concepts in Germany. This project proposal now closes this research and competence gap on these highly innovative reactor concepts.
Therefore, a nuclear computational chain for these MMR designs based on the GRS computational chain is to be established, improved and validated in a 4-year project. IKE, GRS and other national and international organisations interested in this topic will thus be enabled to evaluate this reactor type from a safety point of view or to perform their own analyses independently and thus to build up their own competences on MMRs.
The planned work comprises reactor-physical and thermal-hydraulic model improvements and extensions at the GRS computational chain, their validation on the basis of experiments at the IKE Stuttgart and the exemplary application for safety analyses. The non-commercial design study Special Purpose Reactor (SPR) is used as reference MMR due to the availability of published analyses.
The work is to be carried out in a project duration of 48 months.
The overall objective of the project is to extend the GRS nuclear computational chain for MMRs, validate it against experimental and analytical results and then use it to analyse selected scenarios. This will be achieved through the following sub-objectives:
- Creation of a 3D neutronics model for analyses with the GRS code FENNECS and qualification against published reference data (GRS),
- Further development of ATHLET for MMR with potassium-filled heat pipes, validation against experiments and qualification against published reference data (GRS).
- Experiments on potassium-filled heat pipes to support the model development (IKE)
- Further development of ATHLET for CO2 and air-based Joule cycles as operational heat extraction for power generation (IKE).
- Specification of a consistent MMR model and qualification of the computational chain through integral simulations of the operational and accidental behaviour of the MMR (both).
- Recruitment and development of young scientists with knowledge in neutronics, thermohydraulics and safety analyses with the GRS computational chain (both).
Funded by: BMBF, funding code: 02NUK074
Duration: 01.03.2023 - 28.02.2027
Contact: Prof. Jörg Starflinger, Dr.-Ing. Michael Buck, cooperation with Dr.-Ing. Andreas Schaffrath and Dr. rer nat. Andreas Wielenberg, GRS
The PASTELS project will demonstrate how innovative passive safety systems can support modernisation and optimisation of the European nuclear industry by providing relevant new safety options. The overall objective of the project is to improve the ability of European nuclear actors to design and deliver innovative passive safety systems - which are particularly promising as they do not rely on power supply or human intervention - and simulate their behaviour to support the safety demonstration.
PASTELS will make significant progress in the study of two specific passive systems, the Containment Wall Condenser (CWC) and the Safety Condenser (SACO) by:
- Building on and leveraging existing available computational codes to simulate the relevant thermal-hydraulic phenomena,
- Developing a robust, validated, multi-scale simulation methodology of passive systems,
- Performing new experimental studies to obtain the relevant validation data.
The project will deliver extensive methodology guidelines as well as a roadmap to achieving the licensing and implementation of these innovative passive system technologies in future European Nuclear Power Plants (NPPs).
The PASTELS project has obtained the NUGENIA label on 23/09/2019 (Certificate number: 2019NUG0076 - NUGENIA - SNETP).
Funded by: European Union, Horizon 2020, Förderkennzeichen: 945275
Duration: 01.09.2020 – 28.02.2024
Contact: Prof. Dr.-Ing. Jörg Starflinger, Dr.-Ing. Michael Buck,
Passive storage pool cooling by heat pipes II
The primary goal of the cooperation project PALAWERO II (passive storage pool cooling by heat pipes II) is the further development and validation of the thermohydraulic system code ATHLET (analysis of the thermodynamics of leaks and transients) for the consideration of passive heat dissipation with closed two-phase thermosiphons, thermosiphons for short, for cooling of wet storage in nuclear facilities. On the part of the IKE, the experimental data for the simulation validation are to be provided by GRS Garching. The knowledge gained in the previous projects, the BMWi project FKZ 1501515 (PALAWERO) and the GRS project FKZ RS1543, should be taken up, expanded and optimized.
The main research areas of the IKE are the following:
Improvement of the inner and outer heat transfer at thermosiphons
Performance of long, double-curved thermosiphons
Thermosiphon bundle tests under forced and natural convection on the cooling side
These three focal points are examined on four independent test stands. These are the test stands "ATHOS" (Atmospheric THERmosyphon COoling System) known from the previous project, as well as the newly designed test stands "THOR" (THermOsiphon Laboratory test stand), "Mini-ATHOS" (Miniature-Atmospheric THERmosyphon COoling System) and the " boiling test stand". An overview of the four systems is given below.
The "boiling test rig" aims to improve the heat transport capacity of thermosiphons in the inner heat transfer area on a small scale of 3 m. Due to the modular construction of the test stand, different series of measurements are to be carried out with differently processed thermosiphons. The measurement focus is on improving the internal heat transfer in the area of the evaporation and condensation zone. Various measures are to be taken for this purpose, such as the defined roughening by sandblasting, the introduction of axial grooves, network structures or coating.
With the help of the "Mini ATHOS", the focus is on the external heat transfer in the second part of the project area. With the help of CFD simulations, suitable rib structures should be selected in order to measure them experimentally under forced convective air cooling. The test stand forms the link to the large test facility "ATHOS" and enables influencing variables to be determined under laboratory conditions.
With the "THOR" the effects of 10 m long thermosiphons with different inclinations of the adiabatic zone are to be observed under defined laboratory conditions. In addition, inclined evaporation and condensation zones are also examined.
The "ATHOS" test stand is used to test long thermosiphons under realistic conditions. For this purpose, the heating is provided by two electrically heatable 3 m³ water tanks. The arrangement of the tanks allows tests with individual and tube bundle configurations, both straight and curved. Cooling takes place via a 7.5 m high chimney, which has an adjustable forced convective air flow, so that precise air flows can also be selected in addition to natural draft cooling. The test stand is designed as a long-term test stand and can therefore map seasonal effects.
Funded by: BMWi, funding reference: FKZ 1501515
Duration: 06/01/2020 - 05/31/2025
Contact: M.Sc. Marc Kirsch, M.Sc. Sergio Caceres, Dr.-Ing. Rudi Kulenovic
A sub-project within the joint project CHF and Post-CHF Heat Transfer at Very High Pressure (CPC-HD)
In future nuclear systems, supercritical fluids such as water or CO2 will be used more and more frequently. Precise knowledge of the heat transfer in a wide pressure range is necessary for the safe design of such thermo-hydraulic systems, since transcritical processes are also run through and the system can be in the subcritical range during start-up, shut-down or in the event of accidents. Up to a pressure value of 70% of the critical pressure (reduced pressure pr = 0.7), the heat transfer up to the critical heat flux (CHF) and beyond (post-CHF) has already been intensively investigated. In contrast, the knowledge for reduced pressure values above 0.7 is still very limited.
The aim of the sub-project MEADOW is the experimental investigation and modeling of the heat transfer at pressures in the immediate vicinity of the critical point, with the phenomenon of the boiling crisis at high steam contents (dryout) being in the foreground. Experiments with CO2 as the working medium above and below the critical heat flux density contribute to the creation of a database on the one hand and enable an improved understanding of the physical processes on the other. Based on the database, new models are then developed to describe the heat transfer and the dryout, especially at high pressures. The development of fluid-to-fluid scaling models also enables the transfer of test data from three different fluids (water, CO2 and R134a). By implementing the newly developed models in the thermohydraulic system code ATHLET, its significance for innovative reactor systems is increased.
The project partners within the joint project CPC-HD are the Institute for Applied Thermofluidics of the Karlsruhe Institute of Technology (IAFT, KIT), the Chair for Energy Systems at the Technical University of Munich (LES, TUM) and the Society for Plant and Reactor Safety (GRS) gGmbH.
Funded by: Federal Ministry of Education and Research (BMBF), funding reference: 02NUK062B
Duration: June 2021 – May 2025
Contact: M.Sc. J. Bronik, M.Sc. S. Leopoldus, Dr.-Ing. M. Buck, Prof. Dr.-Ing. J. Starflinger
Weiterentwicklung der Simulationsmodelle für die späte Störfallphase zur Unterstützung der Verbesserung von Severe Accident-Strategien (WESISS)
In most existing reactors, even with advanced core destruction, the main option for protecting the barrier is to feed cooling water into the primary circuit. If a retention in the RPV is not possible, the modelling of the melt behavior in the containment is of utmost importance due to the relevance of the last barrier to prevent a release of radioactive material into the environment. In this case, the question of a possible cooling capabilities and thus preservation of the protective effect of the containment through accident management strategies must be investigated. Particularly in the case of boiling water reactors, flooding of the reactor pit is a possibility. Simulation tools play an important role in assessing the chances of success or the time budget gained by the measure. The simulation tool COCOMO-3D (Corium Coolability Model) allows a realistic simulation of three-dimensional processes during the late phase of core meltdown in the RPV and in the containment on a uniform model basis. Partial models of COCOMO-3D are part of the code ATHLET-CD, which is used for the simulation of transients, accident and incident sequences in safety analyses for light water reactors.
The object of the project, which is funded within the framework of the reactor safety research of the BMUV, is the further development of the models on the basis of extended knowledge about the processes during the late phase of core meltdown accidents. Furthermore, the project focuses on the investigation of the practical applicability of the simulation models, in particular for the investigations on the possibilities of cooling and thus stabilising core melt both in the RPV and in the containment within the scope of accident management measures. In order to enable extensive parameter studies with the use of three-dimensional simulations and high spatial resolution, a parallelisation of COCOMO-3D is to be carried out, which can be used on modern parallel computers.
Sponsered by: BMUV, Förderkennzeichen: 1501635
Duration: August 2021 – September 2024
Contact: M.Sc. A. P. Nedumparambil, M.Sc. M. Petroff , Dr. -Ing. Michael Buck
Finalized Projects
Passive Cooling of Fuel Pool by Heat Pipes ̶ Improvement and Validation of Numerical Models
After the damage of Fukushima Daiichi nuclear power plant blocks 1-4 the reliably safe removal of residual heat is focused even more in current reactor safety research. In this context, especially passive, self-sustaining heat removal systems are of high interest due to their operation based on laws of nature (e.g. gravity, natural convection). Compared to active systems (pump circuits with heat exchangers) they have essential advantages e.g. no need of external power supply and process control. According to their passive operation heat pipes or two-phase thermosiphons (gravity assisted heat pipes) can be applied in spent fuel pools as a redundant, inherently safe heat removal system.
Within the scope of this research project experimental investigations are performed to characterize the heat transfer behaviour of long heat pipes and thermosiphons (tube length >10 m). For this purpose, on the one hand experimental tests with a laboratory setup at well-defined thermal boundary conditions and on the other hand measurements at practical, weather dependent atmospheric conditions using a setup installed in parts on the roof of the laboratory building will be realized providing a wide range of data for different heat pipe/thermosiphon operating states. The experimental data will be used for validation and improvement of existing numerical models for heat pipes and thermosiphons as well as for the derivation of a verified heat transfer correlation and a validated mechanistic simulation model, which will be implemented in the system code ATHLET and subsequently can be applied in safety analyses for the evaluation of heat pipe and thermosiphon heat transfer systems in nuclear installations.
The research project is carried out in close cooperation with the Gesellschaft für Anlagen und Reaktorsicherheit (GRS), which will develop the mechanistic simulation model for ATHLET in the framework of a parallel research project.
Sponsored by: Federal Ministry for Economic Affairs and Energy, contract number: 1501515
Duration period: December 2015 – November 2019
Contacts: M.Eng. C. Graß, Dipl.-Ing. T. Boldt, Dr.-Ing. R. Kulenovic and Prof. Dr.-Ing. J. Starflinger
Further development and validation of AC² for the simulation of innovative LW-SMR
As part of the BMWi's reactor safety research, the system code AC² is analyzed, evaluated and further developed. The joint project "Further development and validation of AC² for the simulation of innovative LW-SMR" (VASiL) is a cooperation between three institutions: GRS, PSS and IKE. The work planned in this project includes the further development and validation of AC², here in particular the program module ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transients). Numerous small modular reactors (SMR) are currently being built or developed in countries, some of which differ significantly from the light water reactors that have been predominantly operated up to now. Many SMR manufacturers also try to passively ensure the long-term cooling of their systems by dissipating heat to an ultimate heat sink.
Figure 1. Illustration of residual heat removal systems in three different SMR concepts: NuScale, SCOR, SMART.
The sub-project of the University of Stuttgart (IKE) focuses specifically on the modelling, validation and applicability of innovative heat exchanger concepts using so-called bayonet heat exchangers and heat pipes for the passive heat dissipation from the containment to the environment.
Suitable models for simulating pressure losses and heat transfer in-/outside of bayonet heat exchangers for single and two-phase flows are to be implemented, tested and validated. Here, different configurations with horizontal and vertical arrangements of the tube bundles as well as different variants of feeding into the bayonet tube (via the central tube or annulus) are to be implemented. It should also be possible to take design measures to improve heat transfer (e.g. ribbing) into account in the modelling.
Appropriate models are to be implemented, tested and validated for the simulation of heat dissipation using heat pipes. Here, the focus should be on loop thermosiphons, in which steam and condensate are routed separately. Essentially, it must be checked and ensured that the modeling of pipe flows in ATHLET can also be used correctly for the configuration in such a heat pipe.
Figure 2. Representation of a bayonet heat exchanger (left) and a loop thermosyphon (right)
Sponsored by: BMWi, Förderkennzeichen: 1501607C
Duration: 01.06.2020 - 31.05.2023
Contact: M.Sc. Sinem Cevikalp, M.Sc. Nelson Rincón, Dr. -Ing. Michael Buck
Untersuchung der Kühlbarkeitsgrenzen von Schüttungen im Post-Dryout Siedebereich sowie beim
Fluten in Anwesenheit nicht-kondensierbarer Gase zur Validierung von COCOMO-3D
Bei einem schweren Reaktorstörfall mit Kühlmittelverlust kann es zu einem Aufschmelzen und einer Verlagerung von Kernmaterial in das untere Plenum des Reaktordruckbehälters (RDB) kommen, wo die Schmelze im Restwasser fragmentieren und ein Schüttbett (Debris) ausbilden kann. Bei unzureichender Abfuhr der Nachzerfallswärme des Schüttbetts kann der RDB-Behälter derart thermisch belastet werden, dass ein Versagen der RDB-Wand (Durchschmelzen) eintritt. Die Abschätzung der Kühlbarkeit von Debris-Schüttbetten ist daher eine wesentliche sicherheitstechnische Fragestellung in der Reaktorsicherheitsforschung, insbesondere hinsichtlich des Accident-Managements zur Beherrschung des genannten Szenarios in einer weit fortgeschrittenen späten Störfallphase.
Gegenstand des im Rahmen der Reaktorsicherheitsforschung vom BMWi geförderten Vorhabens sind grundlegende Einzeleffekt-Experimente mit volumetrisch beheizten Schüttbetten. Die Durchführung der Experimente an der IKE Versuchsanlage DEBRIS dient der Erstellung einer experimentellen Datenbasis zur Validierung und Weiterentwicklung numerischer Simulationsmodelle, welche speziell die thermofluiddynamischen Vorgänge in Debris-Schüttungen adäquat beschreiben. Die Modelle fließen in das integrale Rechenprogramm-System ATHLET/ATHLET-CD ein, das zur Simulation von Transienten, Störfall- und Unfallabläufen in Sicherheitsanalysen für Leichtwasserreaktoren eingesetzt wird.
Gefördert durch: Bundesministerium für Wirtschaft und Energie, Förderkennzeichen: 1501580
Laufzeit: November 2018 – Oktober 2021
Ansprechpartner: M. Petroff und Dr.-Ing. R. Kulenovic
Qualification of Analysis tools for the evaluation of residual heat-driven, self-sufficient systems for decay heat removal
Since the accident in the boiling water reactors in Fukushima Dai-ichi with meltdowns in three reactor blocks, the decay heat removal became a main task in reactor safety research. According to the national nuclear regulations, nuclear power plants need technical safety concepts, which guarantee that radioactive substances are kept inside the nuclear power plant by different safety barriers. The effectiveness of the safety barriers is for the adherence to the following safety objectives:
- Reactivity control
- Core heat removal
- Limitation of release
The decay heat removal still plays an important role for the safety objective core heat removal. At levels of defense 1 to 4a, ensure that the following requirements are met:
- There is enough coolant and heat sinks.
- The heat transfer from the nuclear fuel to the heat sink is ensured.
- The heat removal out of the fuel-cooling pond is ensured.
In the frame of the project, different tasks will be carried out, which are the basics for the safety-related evaluation of a new concept to transfer decay heat from the reactor pressure vessel into a diverse ultimate heat sink. This transfer shall be realized by a self-sufficient system for decay heat removal, based on a Brayton cycle with supercritical carbon dioxide (sCO2) as working fluid.
Currently, only restricted simulations with the German system code ATHLET of such a decay heat removal system and its interaction with existing nuclear power plants, are possible. Furthermore, numerical models of many components of the heat removal system are not validated for sCO2 as working fluid.
The primary aim of the project is to enable scientists and experts to simulate and evaluate a self-sufficient sCO2 system in a professional opinion to remove decay heat to a diverse ultimate heat sink.
Sponsored by: Federal Ministry for Economic Affairs and Energy (BMWi), Project Number: 1501557
Duration period: November 2017 - October 2020
Contacts: M.Sc. K. Theologou, Dr.-Ing. R. Mertz und Prof. Dr.-Ing. J. Starflinger
Supercritical CO2 cycle for FLEXible and sustainable support to the electricity system
Current fossil-fuel power plants have been designed to operate in base-load conditions, i.e to provide a constant power output. However, their role is changing, due to the growing share of renewables, both in and outside the EU. Fossil-fuel plants will increasingly be expected to provide fluctuating back-up power, to foster the integration of intermittent renewable energy sources and to provide stability to the grid. However, these plants are not fit to undergo power output fluctuations.
In this context, sCO2-Flex consortium addressees this challenge by developing and validating (at simulation level the global cycle and at relevant environment boiler, heat exchanger(HX) and turbomachinery) the scalable/modular design of a 25MWe Brayton cycle using supercritical CO2, able to increase the operational flexibility and the efficiency of existing and future coal and lignite power plants.
sCO2-Flex will develop and optimize the design of a 25MWe sCO2 Brayton cycle and of its main components (boiler, HX, turbomachinery, instrumentation and control strategies) able to meet long-term flexibility requirements, enabling entire load range optimization with fast load changes, fast start-ups and shut-downs, while reducing environmental impacts and focusing on cost-effectiveness. The project, bringing the sCO2 cycle to TRL6, will pave the way to future demonstration projects (from 2020) and to commercialization of the technology (from 2025). Ambitious exploitation and dissemination activities will be set up to ensure proper market uptake.
Consortium brings together ten partners, i.e academics (experts in thermodynamic cycle/control/simulation, heat
exchanging, thermoelectric power, materials), technology providers (HX, Turbomachinery) and power plant operator (EDFcoordinator) covering the whole value chain, constituting an interdisciplinary group of experienced partners, each of them providing its specific expertise and contributing to the achievement of the project’s objectives.
Sponsered by: Europäische Union (Horizon 2020), Förderkennzeichen: 764690
Duration period: Januar 2018 - Dezember 2020
Contacts: M.Sc. A. Wahl, Dr.-Ing. R. Mertz und Prof. Dr.-Ing. J. Starflinger
Supercritical CO2 heat removal system
The supercritical CO2 heat removal system “sCO2-HeRo” safely, reliably and efficiently transfers the decay heat from the reactor core to an ultimate heat sink without any external power sources like emergency diesel generators or batteries. Since this system is powered by the decay heat itself, in case of a station blackout and loss of ultimate heat sink accident scenario in a nuclear power plant, it can be considered as an excellent backup cooling system for the reactor core. Because of that, the sCO2-HeRo system can be seen as an innovative reactor safety concept as it improves the safety of BWRs (Boiling Water Reactors) and PWRs (Pressurized Water Reactors) through a self-propellant, self-sustaining and self-launching, highly compact Brayton-cycle cooling system using supercritical carbon dioxide as working fluid.
This system provides breakthrough options with scientific and practical maturity, which will be finally demonstrated and experimentally proven by practical reactor simulation studies in the unique glass model of the Gesellschaft für Simulatorschulung GfS, accompanied by simulation studies with the German thermal-hydraulic system code ATHLET.
The objective of the project is to show the feasibility of the sCO2-HeRo system in a scaled down demonstrator, which will be installed into the pressurized water reactor glass model (step towards technology readiness level 3 (TRL3)). Therefore, all components must be designed, manufactured, experimentally investigated and installed into the glass model. These are a compact heat exchanger (CHX), which connects the steam side of the glass model with the sCO2-HeRo cycle, a turbo-compressor system with integrated generator and a sink heat exchanger, which transfers the heat from the sCO2-HeRo cycle to the ultimate heat sink, the ambient air.
The experimental investigation on the heat transfer capacity in the compact heat exchanger will be done in the supercritical CO2 test loop SCARLETT at IKE University of Stuttgart. The sink heat exchanger, purchased by colleagues from Prague, as well as the turbo-compressor machine set, designed and manufactured by University of Duisburg/Essen, will be tested in the CO2 SUSEN loop at CV Rez.
The sCO2-HeRo project is funded by the European Union under the Horizon 2020 program.
Sponsered by: Europäische Union (Horizon 2020), Förderkennzeichen: 662116
Duration period: September 2015 - August 2018
Contacts: M.Sc. M. Strätz, Dr.-Ing. R. Mertz und Prof. Dr.-Ing. J. Starflinger
Experimental Investigations and Numerical Simulations of Turbulent Flows and Fluid-Structure-Interaction close to Weld Seams and Through-Wall Cracks
The understanding and quantitative description of fluid- and structure-mechanical interactions in piping systems in case of turbulent flow conditions combined with thermal fluctuations as a consequence of mixing processes is essential for the safety-related evaluation and the development of failure criteria in the cooling system of nuclear power plants. The current research project UNSCHRO investigates, in cooperation with the Material Testings Institute (MPA) Stuttgart, the loading of the piping material close to circumferential weld seams and through-wall crack caused by thermal fluctuations downstream of a mixing tee. The modular concept of the Fluid-Structure-Interaction (FSI) test facility enables investigations of turbulent flow conditions and the temperature distribution in the pipe wall close to a weld seam as well as a through-wall crack with realistic experimental conditions (maximum pressure of 80 bar, maximum temperature of 280 °C). Furthermore, the fluid-structure-interaction is analyzed by means of time-dependent coupled numerical simulations (CFD and FEM) as well as compared to the experimental results.
Sponsored by: Federal Ministry of Education and Research (BMBF), contract number: 02NUK040B
Duration period: October 2014 – June 2018
Contacts: M.Sc. S. Schmid, M.Sc. Mi Zhou and Dr.-Ing. R. Kulenovic
Experimental and numerical investigation of flow phenomena in the mixing region of a T- junction
Flow turbulence and thermal stratification can lead to thermal fatigue and corrosion in piping systems of nuclear power plants. Those can cause malfunction and damage in components of the cooling circuit even after a short operational life time. Therefore, representative flows with and without density stratification should be experimentally investigated near the mixing region of a horizontal T -junction configuration. Density differences, representing temperature differences of the real cooling water of a power plant piping system, is modelled by salt solution in the branch pipe of IKE’s MFI-test-rig (mixed-fluid-interaction). It should be investigated which conditions lead to specific flow structure formation. Accompanying CFD-Simulations with OpenFOAM are performed using the Large-Eddy Simulation method (LES). The results of the investigation contribute to the understanding of turbulent and stratified flow in the vicinity of the T-junction. It also provides possibilities for developing efficient methods for validation and prediction of turbulent mixing. Subsequently, the applicability of the LES should be demonstrated by experimental validation in the case of a vertical T-junction configuration.
Sponsored by: Bundesministerium für Wirtschaft und Energie, contract number: 1501508
Duration period: August 2015 - July 2018
Contacts: M.Sc. A. Isaev and Prof. Dr.-Ing. E. Laurien
Photosynthetische Bio-Wasserstoffproduktion mit Purpurbakterien ohne Licht
In einem Gemeinschaftsprojekt mit Kollegen vom Institut für Biomaterialien und biomolekulare Systeme, Abteilung Bioenergetik, wird im Wasserstofflabor des IKE ein neuartiger Ansatz zur Produktion von Bio-Wasserstoff mit Purpurbakterien erforscht. Hierzu wurde am IKE ein Versuchsstand gebaut, bestehend aus zwei Behältern, die mit Druck-, Temperatur- und pH-Sensoren, Spektrometeranschlüsse etc. ausgestattet sind. An eine Messdatenerfassung angeschlossen konnte die Wasserstoffproduktion in Vorversuchen bereits nachgewiesen werden. Ziel des vorliegenden Projektes ist es, in einer ersten, neunmonatigen Sondierungsphase die Wasserstoffproduktion zu quantifizieren. Bei Erfolg ist es möglich, eine zweite mehrjährige Machbarkeitsphase durchzuführen, um die Biowasserstofferzeugung in ein Produkt der Bioökonomie zu überführen.
Gefördert durch: Bundesministerium für Bildung und Forschung, „Neue Ideen für die Bioökonomie“, Förderkennzeichen: 031B0135
Laufzeit: März 2016 – November 2016 (Sondierungsphase)
Ansprechpartner: Dipl.-Ing. T. Boldt, Dr.-Ing. R. Mertz, Dr. rer.nat. C. Autenrieth (IBBS), Prof. Dr. R. Ghosh (IBBS), Prof. Dr.-Ing. J. Starflinger
(Code für das Europäische Management schwerer Störfälle)
Das F&E-Projekt CESAM zielt auf die Verbesserung der europäischen Referenzcodes ASTEC (Accident Source Term Evaluation Code) ab, der für Managementanalysen bei schweren Unfällen in Kernkraftwerken (KKWs) verwendet wird. Die Forschungsaktivitäten werden gemeinsam von 18 Projektpartnern aus 12 europäischen Ländern und Indien durchgeführt. Zuerst werden die aktuellen Modellierungsfähigkeiten von ASTEC für die relevanten Phänomene bei schweren Unfällen bewertet. Dadurch ergeben sich Empfehlungen für die weitere Code-Entwicklung und Modellverbesserungen im Rahmen des Projekts. Die verbesserten ASTEC-Modelle werden danach anhand von Experimenten validiert und für Reaktor-Anwendungen geprüft. ASTEC-Referenzdatensätze für die wichtigsten generischen Typen von Kernkraftwerken werden von den Partnern zusammen erstellt. Danach können verschiedene Unfallszenarien in KKWs simuliert werden, um mögliche Verbesserungen der Maßnahmen bei schweren Störfällen zu analysieren, und um geeignete Anleitungen hinsichtlich der Anwendung von ASTEC für KKW-Analysen festzulegen.
Gefördert durch: Europäische Union (7tes Rahmenprogram)
Laufzeit: April 2013 – März 2017
Ansprechpartner: Dipl.-Ing. C. D‘ Alessandro, Dr.-Ing. M. Buck und Prof. Dr.-Ing. J. Starflinger
Der Bodensee ist mit einer maximalen Tiefe von 254 m, einer Oberfläche von ca. 535 km² und einem Gesamtvolumen von annähernd 50 km³ der größte Voralpensee am Nordrand der Alpen. Für den Bodensee wurde ein Online Informations- und Notfallschutzsystem zur Vorhersage des hydrodynamischen Verhaltens und der Wasserqualität im Rahmen eines Verbundforschungsprojektes BodenseeOnline entwickelt, das vom Bundesministerium für Bildung und Forschung und von der Deutschen Forschungsgemeinschaft finanziert wurde. Ergebnisse von BodenseeOnline werden auch von der Internationalen Gewässerschutzkommission für den Bodensee für wasserwirtschaftliche Entscheidungen genutzt. Außerdem steuern die Seewasserwerke am Bodensee wichtige Daten für den Betrieb und die Kontrolle von BodenseeOnline bei.
BodenseeOnline besteht aus einer umfassenden Datenbank, in der sowohl historische Daten als auch aktuelle Messinformationen gespeichert werden. Für die Simulationsrechnungen kommt ein dreidimensionales hydrodynamisches Modell für die Seeströmung in Kombination mit einem Windmodell (IKE) und einem biogeochemisches Modell, das mit dem hydrodynamischen Modell gekoppelt ist, zur Anwendung. Die Nutzer Institut für Seenforschung (ISF), Internationale Gewässerschutzkommission (IGKB), Arbeitsgemeinschaft Wasserwerke Bodensee-Rhein (AWBR) und Gefahrenabwehr (Feuerwehr, Wasserpolizei, Katastrophenschutz) am Bodensee haben über einen geschützten Zugang Zugriff auf alle wichtigen Daten und Modellinterpretationen, die eine detaillierte Einschätzung der jeweiligen Situation ermöglichen.
Der öffentliche Teil von BodenseeOnline stellt für Anwohner und Besucher des Sees folgende aktuelle Informationen über die nächsten 3 Tage zur Verfügung:
- Wassertemperaturen
- Windverhältnisse über dem See
- Wellenhöhen und Richtung
- Seeströmung
Mittlerweile hat die LUBW das System übernommen, wo es von KUP und IKE weiterentwickelt wird.
Gefördert durch: Ministerium für Umwelt, Klima und Energiewirtschaft, AZ: 1-0272.2
Laufzeit: Februar 2014 – Dezember 2016
Ansprechpartner: N. Kaufmann M.A.
Experimentelle Untersuchungen zu Kühlbarkeit und Fluten prototypischer Schüttbett-Konfigurationen -
Phase II: Flutexperimente
Bei einem schweren Reaktorstörfall mit Kühlmittelverlust kann es zu einem Aufschmelzen und einer Verlagerung von Kernmaterial in das untere Plenum des Reaktordruckbehälters (RDB) kommen, wo die Schmelze im Restwasser fragmentieren und ein Schüttbett (Debris) ausbilden kann. Bei unzureichender Abfuhr der Nachzerfallswärme des Schüttbetts kann der RDB-Behälter derart thermisch belastet werden, dass ein Versagen der RDB-Wand (Durchschmelzen) eintritt. Die Abschätzung der Kühlbarkeit von Debris-Schüttbetten ist daher eine wesentliche sicherheitstechnische Fragestellung in der Reaktorsicherheitsforschung, insbesondere hinsichtlich des Accident-Managements zur Beherrschung des genannten Szenarios in einer weit fortgeschrittenen späten Störfallphase.
Gegenstand des im Rahmen der Reaktorsicherheitsforschung vom BMWi geförderten Vorhabens sind grundlegende Einzeleffekt-Experimente mit volumetrisch beheizten Schüttbetten. Die Durchführung der Experimente an der IKE Versuchsanlage DEBRIS dient der Erstellung einer experimentellen Datenbasis zur Validierung und Weiterentwicklung numerischer Simulationsmodelle, welche speziell die thermofluiddynamischen Vorgänge in Debris-Schüttungen adäquat beschreiben. Die Modelle fließen in das integrale Rechenprogramm-System ATHLET/ATHLET-CD ein, das zur Simulation von Transienten, Störfall- und Unfallabläufen in Sicherheitsanalysen für Leichtwasserreaktoren eingesetzt wird.
Gefördert durch: Bundesministerium für Wirtschaft und Energie, Förderkennzeichen: 1501507
Laufzeit: September 2015 – August 2016
Ansprechpartner: Dipl.-Ing. S. Leininger und Dr.-Ing. R. Kulenovic
Für die zukünftige Energieerzeugung ist Kohlendioxid (CO2) bei superkritischem Druck (sCO2) ein attraktiver Wärmeträger. Der Wärmeübergang dieser überkritischen Strömungen ist sehr komplex und wird daher in dem Projekt theoretisch untersucht. Dazu wird die Methode der Direkten Numerischen Simulation (DNS) mit der CFD-Software OpenFOAM entwickelt und auf dem Höchstleistungsrechenzentrum Stuttgart (HLRS) angewendet. Bei einer mit konstanter Wärmestromdichte beheizten Rohrströmung kann die Wandtemperatur bei Durchgang durch den pseudokritischen Punkt an derjenigen Stelle stark ansteigen, wo die thermophysikalische Eigenschaften an der Wand stark variieren. Die grundlegenden Mechanismen dieser Verschlechterung und auch diejenigen einer anschließenden Verbesserung werden mit Hilfe der Turbulenzstatistik untersucht. Die Ergebnisse erklären den Einfluss von Beschleunigung und Auftriebskräften auf die Relaminarisierung der Strömung. Diese Untersuchung unterstützt auch die Entwicklung von zukünftigen Turbulenzmodellen.
Gefördert durch: Forschungsinstitut für Kerntechnik und Energiewandlung e.V.
Laufzeit: März 2013 – März 2016
Ansprechpartner: Prof. Dr.-Ing. E. Laurien
Basierend auf den Ergebnissen des FP 7 ENEN-RU Projekts, soll im FP7 ENEN-RU II Projekt das Rahmenprogramm der Zusammenarbeit zwischen EU und der russischen Föderation im Hinblick auf Ausbildung und Training im Nuklearbereich ausgebaut werden. Dies umfasst eine weitere Analyse der Zusammenarbeit, sowohl kurz- als auch langfristig; Definition von Kooperationsmöglichkeiten und –hemmnissen und außerdem die Fortführung gemeinsamer Unterrichts- und Übungssessions, die im Rahmen des Projekts durchgeführt werden.
Die Einzelziele des Projektes sind:
- Definition eines Plans zur Implementierung, basierend auf den Bedürfnissen einer langzeitigen Kooperation, auf den man sich im Rahmen des Vorgängerprojekts ENEN-RU einigte
- Lösung der Schwierigkeiten einer Kooperation, die im ENEN-RU Projekt identifiziert wurden,
- diesen Plan zur Implementierung nachhaltig umzusetzen,
- das Rahmenprogramm des Wissensmanagement zu betreiben
- die Unterrichts- und Trainingseinrichtungen, -labors und ausrüstung aufzulisten und fördern.
Gefördert durch: Europäische Union, GA Nr.: 605149
Laufzeit: Juli 2014 – Juni 2017
Joint Advanced Severe Accidents Modelling and Integration for Sodium-Cooled Fast Neutron Reactors
(Fortschrittliches Modellieren schwerer Störfälle und Integration von Natrium-gekühlten schnellen Neutronenreaktoren)
Im Rahmen des JASMIN Projekts ist das IKE an der Entwicklung eines robusten, fortschrittlichen Simulationstools für Sicherheitsanalysen von Natriumgekühlten schnellen Reaktoren, hauptsächlich zur Untersuchung schwerer Störfälle, beteiligt. Ziel ist es, einen neuen europäischen Störfallcode ASTEC-Na zu schaffen, der verbesserte Modelle, die dem Fortschritt der neueren Forschung im Bereich der Leichtwasserreaktoren Rechnung tragen, mit moderner Softwarearchitektur und hoher Flexibilität vereint, um innovatibe Reaktordesigns berücksichtigen zu können. Ausgangspunkt ist der von IRSN und GRS für Leichtwasserreaktoren entwickelte Integralcode ASTEC. Bei den heute verfügbaren Codes für schnelle Natrium gekühlte Reaktoren handelt es sich um Entwicklungen aus den 80er Jahren, die an damalige Reaktordesigns angepasst sind. Im Gegensatz zu den älteren Codes, in denen wichtige Phänomene zur Bewertung der Sicherheit schneller Systeme separat behandelt werden, soll der neue Code ASTEC-Na die Möglichkeit bieten, alle diese Phänomene in einem einheitlichen Tool zu erfassen (Initialisierungsphase, Verhalten des Sicherheitsbehälters, Quelltermermittlung,...). Damit wird der aktuelle Stand der Entwicklungen für Leichtwasserreaktoren genutzt und für Untersuchungen von Störfällen in schnellen Reaktoren nutzbar gemacht.
Das Projekt wurde im Dezember 2011 ins Leben gerufen mit einer Dauer von vier Jahren. Beteiligte Organisationen sind:
- Institut de Radioprotection et de Sûreté Nucléaire
- Karlsruhe Institute for Technology
- Gesellschaft für Anlagen- und Reaktorsicherheit mbH
- Agenzia Nazionale per le Nuove Tecnologie, l'Energia e lo Sviluppo Economico Sostenibile
- Centro de Investigaciones Energeticas Medio Ambientales y Tecnologicas
- Universität Stuttgart
- EURATOM Joint Research Centres
- AREVA NP SAS
- Electricité de France SA
Gefördert durch: Europäische Union, FP7-295802
Laufzeit: Dezember 2011 – November 2015
Ansprechpartner: Dipl.-Phys. N. Guilliard und Dr.-Ing. M. Buck
Modellentwicklung zu Schmelzeverhalten und Kühlbarkeit in Reaktordruckbehälter und Sicherheitsbehälter
Im Projekt MSKRS werden Simulationsmodelle für Sicherheitsanalysen bereitgestellt, die eine realistische Simulation der Prozesse während des Kernschmelzens im Reaktordruckbehälter sowie im Sicherheitsbehälter ermöglichen. Insbesondere sollen die Möglichkeiten einer Kühlung und Stabilisierung von Schüttbetten, sowohl im Reaktordruckbehälter als auch im Sicherheitsbehälter untersucht werden können. Hierzu sollen neben der Weiterentwicklung bereits in ATHLET-CD (Analyse der Thermohydraulik von Lecks und Transienten- Core Degradation) integrierter Modelle auch erstmalig Modelle zum Schmelzeverhalten in einer wassergefüllten Reaktorgrube bereitgestellt und für die Ankopplung an Systemcodes für den Sicherheitsbehälter vorbereitet werden. Um auch Schüttbettkonfigurationen beurteilen zu können, die aufgrund ihrer Asymmetrie eine dreidimensionale Beschreibung erfordern, werden die Modellansätze von zwei auf drei Dimensionen erweitert.
Gefördert durch: Bundesministerium für Wirtschaft und Energie, Förderkennzeichen: 1501482
Laufzeit: August 2014 – Juli 2017
Ansprechpartner: Dipl.-Ing. W. Hilali, Dr.-Ing. M. Buck und Prof. Dr.-Ing. J. Starflinger
Experimentelle und numerische Untersuchung der Thermo-Fluid Dynamik der Strömungsvermischung in einem T-Stück-Rohrleitungssystem
Die turbulente Vermischung von heißen und kalten Kühlmittelströmungen in T-förmigen Rohrleitungsverbindungen kann in Kernkraftwerken (z. B. im Nachwärmeabfuhrsystem) sicherheitsrelevante, komplexe thermomechanische Strukturbelastungen im Rohrleitungssystem induzieren, die bei zyklischer Wiederkehr zur thermischen Materialermüdung (High Cycle Thermal Fatigue - HCTF) führen. Im Bereich der Vermischungszone kann dieses Strömungs-phänomen eine Schädigung des Rohrwandmaterials (z. B. kleine Risse, wanddurchdringende Risse mit Leckagen) bis hin zu einem vollständigen Rohrversagen (Rohrabriss) verursachen.
Im Rahmen des Projekts wird die Vermischung der Strömungen in einem T-Stück für kraftwerksrelevante Bedingungen (∆Tmax=260 K, pmax=75 bar) experimentell am FSI-Versuchskreislauf (Fluid Structure Interaction), der von der MPA –Materialprüfungsanstalt Universität Stuttgart und IKE gemeinsam betrieben wird, untersucht. Gleichzeitig werden numerische Strömungssimulationen (CFD) mit Hilfe der Grobstruktur-Simulationsmethode (LES - Large-Eddy Simulation) durchgeführt, um einen detaillierten Einblick über die Vermischungsvorgänge im T-Stück zu erhalten. Anhand der experimentellen Daten werden die Simulationsergebnisse verifiziert.
Gefördert durch: Deutscher Akademischer Austauschdienst (DAAD)
Laufzeit: Oktober 2012 – März 2016
Ansprechpartner: M.Tech. K. Selvam und Prof. Dr.-Ing. E. Laurien
Superkritisches CO2 - Machbarkeitsstudie
In diesem Projekt wird die Realisierbarkeit eines autarken Nachwärmeabfuhrsystems untersucht, welches bei einem Unfall in einem Kernreaktor mit gleichzeitigem Ausfall der Hauptwärmesenke und der Notstromversorgung die Nachwärme sicher und zuverlässig in eine diversitäre ultimative Wärmesenke abführen kann.
Hierzu wird die Machbarkeitsstudie „sCO2-MA“ durchgeführt, die zum einen Möglichkeiten und Einsatzgrenzen eines solchen Systems in einem Kraftwerk untersucht. Hierzu werden Parameterstudien mit dem deutschen Systemcode ATHLET durchgeführt. Zum anderen soll die technische Realisierbarkeit untersucht werden, d.h. es sollen Komponenten des Kreislaufes spezifiziert werden. Im Kontakt mit Herstellern wird so ein mögliches Layout des Nachwärmeabfuhrsystems entworfen.
Mit Hilfe dieser Machbarkeitsstudie soll das Potenzial eines solchen Systems zur Nachrüstung in einem Kernkraftwerk oder einer kerntechnischen Anlage abgeschätzt werden, um eine Entscheidung über den Start einer detaillierteren Untersuchung zu treffen.
Gefördert durch: Bundesministerium für Wirtschaft und Energie, Förderkennzeichen: 1501494
Laufzeit: März 2015 - Februar 2016
Ansprechpartner: Dr.-Ing. R. Mertz und Prof. Dr.-Ing. J. Starflinger
Numerical simulation of flows in a containment using CFD methods
Two-phase flows with water droplets greatly affect the thermal-hydraulic behavior in the containment of a Pressurized Water Reactor (PWR). Such flows occur, inter alia, in French and Eastern European Pressurized Water Reactors in the form of spray cooling. The spray cooling ensures, in case of a leak in the primary circuit, the reduction of the increased pressure in the containment due to the spilled steam by decreasing the temperature. Also, there is a natural convection or a circulation due to the density difference between the hot steam and the colder gases located in the containment. Moreover, the hot steam can cool down on cold walls (wall condensation); volume condensation can also occur due to the saturated moist air in the presence of low temperatures in the containment.
The aim of the project TROBEL is to simulate the named flow phenomena spray cooling, natural convection and condensation using the CFD methods.
Sponsored by: Federal Ministry for Economic Affairs and Energy, contract number: 1501493
Duration period: March 2015 - February 2018
Contact: M.Sc. C. Kaltenbach, Dipl.-Ing. A. Mansour and Prof. Dr.-Ing. E. Laurien
Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Störfallmaßnahmen
Im deutschen Verbundprojekt WASA-BOSS arbeiten Forscher verschiedener Institute und Universitäten zusammen, um die Leistung von Severe Accident Codes zu verbessern. Im Rahmen des Projektes werden verschiedene Maßnahmen untersucht, die einen potentiellen Unfall mit auftretendem Kernschmelzen aufhalten sollen. Des Weiteren soll der gesamte Unfallablauf vom auslösenden Ereignis bis hin zur Freisetzung in die Umgebung simuliert werden. Dazu wird die Modellierung von Prozessen im unteren Plenum von Siedewasserreaktoren erweitert, sowie eine Option der Außenkühlung des Reaktordruckbehälters entwickelt. Dabei kommen die Störfallcodes ATHLET-CD, COCOSYS und MELCOR zum Einsatz. Sowohl die Modellierung als auch die Validierung der Code-Erweiterungen basiert auf relevanten Experimenten.
Gefördert durch: Bundesministerium für Bildung und Forschung, Förderkennzeichen: 02NUK028C
Laufzeit: März 2013 - Februar 2016
Ansprechpartner: Dr.-Ing. M. Buck und Prof. Dr.-Ing. J. Starflinger
Contact
Michael Buck
Dr.-Ing.Head of department RSU
Rudi Kulenovic
Dr.-Ing.Head of department EW