High temperature reactors for cogeneration applications
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K. Verfondern
Abstract
There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 °C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 °C.
Kurzfassung
Nuklearenergie besitzt ein großes Potential auch im Wärmemarkt. Viele Industrien haben einen hohen Bedarf an Prozesswärme und Prozessdampf bei unterschiedlichen Temperaturen und Drücken, einsetzbar z. B. bei der Meerwasserentsalzung, in Fernwärmenetzen oder in chemischen Prozessen. Die künftige Generation von Nuklearanlagen wird in der Lage sein, in das weite Feld der Kraft-Wärme-Kopplung einzudringen, um so den Verlust durch Abwärme zu verringern und gleichzeitig den Wirkungsgrad der Anlage zu erhöhen. Dies erfordert eine Anpassung an die Bedürfnisse der Industrien im Hinblick auf Größe und Anwendung. Alle Nuklearkonzepte der 4. Generation sind für Kählmittelaustrittstemperaturen > 500 °C ausgelegt, die Anwendungen im niedrigen und mittleren Temperaturbereich erlauben. Ein VHTR wäre sogar in der Lage, den gesamten Temperaturbereich bis zu rund 1 000 °C abzudecken.
References
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© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
- A spray cooling technique for spent fuel assembly stored in pool
- KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
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- Validation of the ATHLET-SC code by trans-critical transient data
- Qualification of CFD-models for multiphase flows
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- Critical flow phenomena and modeling in advanced nuclear safety technology
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- Thermohydraulic safety issues for liquid metal cooled systems
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
- A spray cooling technique for spent fuel assembly stored in pool
- KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
- 10.3139/124.110680
- Validation of the ATHLET-SC code by trans-critical transient data
- Qualification of CFD-models for multiphase flows
- The reactor dynamics code DYN3D
- Critical flow phenomena and modeling in advanced nuclear safety technology
- 10.3139/124.110682
- Safety and security aspects in design of digital safety I&C in nuclear power plants
- Thermohydraulic safety issues for liquid metal cooled systems
- Design and safety analysis of the helium cooled solid breeder blanket for CFETR
- Qualification of pebble fuel for HTGRs
- High temperature reactors for cogeneration applications
