The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant
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H.-T. Lin
Abstract
In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1 088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1 100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.
Kurzfassung
In diesem Beitrag wurde das TRACE/SNAP-Modell des in Bau befindlichen Kernkraftwerks Lungmen (2 Blöcke des Typs Advanced Boiling Water Reactor (ABWR)) in Taiwan für die Simulation und Analyse der Ultimate Response Guideline (URG) verwendet. Die wichtigsten Maßnahmen der URG sind Druckentlastung und Niederdruck-Einspeisesystem des Reaktors und die Containment Entlüftung. Schwerpunkt dieser Arbeit ist die Bewertung der URG des Lungmen-Kernkraftwerks (KKW) unter Fukushima-ähnlichen Bedingungen. Der erste Schritt ist dabei die Etablierung des Lungmen NPP TRACE/SNAP-Modells. Zur Bewertung der Systemrückmeldung wurden FSAR-Daten zum Vergleich mit den Ergebnissen des Thermohydraulikcodes TRACE verwendet. Der zweite Schritt ist die URG Simulation und Analyse unter Fukushima-ähnlichen Bedingungen mit Hilfe des Lungmen NPP TRACE/SNAP-Modells. In diesem Schritt wurde auch der „no URG“ Fall behandelt um die URG-Leistungsfähigkeit des Lungmen-KKW bewerten zu können. Zur Bestätigung der mechanischen Eigenschaften und Integrität der Brennelemente, wird im letzten Schritt eine FRAPTRAN-Analyse durchgeführt. Nach den Ergebnissen der TRACE-Analyse kann die URG die maximale Hüllrohrtemperatur (PCT) unter 1 088.7 K halten unter Fukushima-ähnlichen Bedingungen. Dies zeigt, dass das Lungmen-KKW sicher betrieben werden kann. Die Ergebnisse der FRAPTRAN-Analyse zeigen auch, dass die Integrität der Brennelemente unter diesen Bedingungen nicht gehalten werden kann.
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© 2015, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved data evaluation methodology for energy ranges with missing experimental data
- Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems
- Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant
- The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant
- Development of a hydrogen diffusion gothic model of MARK III-containment
- Analysis of safety parameters for the MINERVE reactor
- Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378
- Neutronic performance of (ReprocessedU/Th)O2 fuel in CANDU 6 reactor
- Modelling study on production cross sections of 111In radioisotopes used in nuclear medicine
- Study of 60Co as gamma source in backscatter gamma densitometers
- Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey
- Technical Notes
- Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved data evaluation methodology for energy ranges with missing experimental data
- Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems
- Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant
- The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant
- Development of a hydrogen diffusion gothic model of MARK III-containment
- Analysis of safety parameters for the MINERVE reactor
- Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378
- Neutronic performance of (ReprocessedU/Th)O2 fuel in CANDU 6 reactor
- Modelling study on production cross sections of 111In radioisotopes used in nuclear medicine
- Study of 60Co as gamma source in backscatter gamma densitometers
- Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey
- Technical Notes
- Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions
