Analysis of operating characteristics of IPWR under natural circulation
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H. Zhu
Abstract
Many integrated pressurized water reactor (IPWR) designs using natural circulation operation mainly to enhance their inherent safety. The operating characteristics of primary coolant are completely different without mechanical pumps. The designs and safety analysis of forced circulation reactors are widely researched, but the natural circulation characteristics of IPWR have not been well studied by literatures. The present work discussed the thermal-hydraulic characteristics of IPWR under natural circulation conditions by using the best estimate codes RELAP5. And the effect of system parameters on natural circulation characteristics of IPWR is also studied. The results show that, the primary coolant average temperature and steam pressure are two key parameters that affect the natural circulation stable operating load. The set value of primary coolant average temperature effects the core outlet temperature and the steam temperature, but the primary coolant flow is basically the same under different primary coolant average temperature but same load conditions. The smaller steam pressure is more conducive to produce superheated steam, but there is risk of two phase flow instability in OTSG secondary side. The rapid load change process under natural circulation indicating that the reactor has a good load tracking characteristics under natural circulation, but the rapid change of primary coolant temperature will cause oscillations in system parameters.
Kurzfassung
Viele integrale Druckwasserreaktoren (IPWR) nutzen Naturumlaufbetrieb, um ihre Sicherheit zu erhöhen. Das Betriebsverhalten des Primärkühlmittels ohne Pumpen ist ein völlig anderes als das mit Pumpen. Während der Einfluss von Zwangsumlauf in Reaktoren weitgehend in der Literatur beschrieben sind, findet sich zum Naturumlauf in IPWR erst wenig Literatur. In der vorliegenden Arbeit wurden die thermohydraulischen Eigenschaften von IPWR unter Naturumlaufbedingungen unter Verwendung der Best Estimate Codes RELAP5 diskutiert. Dabei wird auch der Einfluss von Systemparametern auf den Naturumlauf in IPWR untersucht. Die Ergebnisse zeigen, dass die durchschnittliche Temperatur des Primärkühlmittels und der Dampfdruck zwei Schlüsselparameter sind, die den stabilen Betrieb im Naturumlauf beeinflussen. Der eingestellte Wert der Primärkühlmitteltemperatur beeinflusst die Kernausgangstemperatur und die Dampftemperatur, aber der Primärkühlmittelstrom ist bei unterschiedlicher Primärkühlmitteltemperatur und gleichen Lastbedingungen im Wesentlichen gleich. Der kleinere Dampfdruck ist günstiger für die Erzeugung von überhitztem Dampf, aber es besteht die Gefahr einer zweiphasigen Strömungsinstabilität auf der Sekundärseite der Dampferzeuger. Der schnelle Lastwechselprozess im Naturumlauf zeigt, dass der Reaktor im Naturumlauf ein gutes Lastverfolgungsverhalten aufweist, aber die schnelle Änderung der Primärkühlmitteltemperatur zu Schwankungen der Systemparameter führt.
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© 2018, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- Reactor safety research within the Helmholtz Association
- Analysis of the impact of different scenarios on the simulation results of unauthorized dilution of boric acid in the coolant of the primary circuit of the NPP-2006
- Assessment of void fraction predictability of system codes in subchannels
- Review on using nanofluids for heat transfer enhancement in nuclear power plants
- Analysis of operating characteristics of IPWR under natural circulation
- Impact of spacer on inter sub-channel mixing of coolant in nuclear fuel bundle: a survey and future patterns of research and advances
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- Reactor safety research within the Helmholtz Association
- Analysis of the impact of different scenarios on the simulation results of unauthorized dilution of boric acid in the coolant of the primary circuit of the NPP-2006
- Assessment of void fraction predictability of system codes in subchannels
- Review on using nanofluids for heat transfer enhancement in nuclear power plants
- Analysis of operating characteristics of IPWR under natural circulation
- Impact of spacer on inter sub-channel mixing of coolant in nuclear fuel bundle: a survey and future patterns of research and advances
