Progress of Experimental Research on Nuclear Safety in NPIC
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H. Gong
Abstract
Two kinds of Generation III commercial nuclear power plants have been developed in CNNC (China National Nuclear Corporation), one is a small modular reactor ACP100 having an equivalent electric power 100 MW, and the other is HPR1000 (once named ACP1000) having an equivalent electric power 1 000 MW. Both NPPs widely adopted the design philosophy of advanced passive safety systems and considered the lessons from Fukushima Daichi nuclear accident. As the backbone of the R&D of ACP100 and HPR1000, NPIC (Nuclear power Institute of China) has finished the engineering verification test of main safety systems, including passive residual heat removal experiments, reactor cavity injection experiments, hydrogen combustion experiments, and passive autocatalytic recombiner experiments. Above experimental work conducted in NPIC and further research plan of nuclear safety are introduced in this paper.
Kurzfassung
China National Nuclear corporation hat zwei verschiedene Arten von kommerziellen Kernkraftwerken der Generation III entwickelt: zum einen den kleinen modularen Reaktor ACP100 (mit einer elektrischen Leistung von 100 MW) und zum anderen den Reaktor HPR1000 (früher ACP1000 genannt) mit einer elektrischen Leistung von 1 000 MW. Beide Konzepte berücksichtigen im Design fortschrittliche passive Sicherheitssysteme und lessons learned aus dem Fukushima Daichi Unfall. Das für die Forschung und Entwicklung zuständige Nuclear Power Institute of China hat die Verifikationsversuche der wesentlichen Sicherheitssysteme, die Versuche zur passiven Wärmeabfuhr, zur Einspeisung in die Reaktorgrube, zur Wasserstoffverbrennung und zur passiven autokatalytischen Rekombination beendet. Die wesentlichen Arbeiten werden in diesem Beitrag kurz skizziert.
References
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© 2016, Carl Hanser Verlag, München
Artikel in diesem Heft
- 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
Artikel in diesem Heft
- 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
