Analytical study on degraded core quenching
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O. S. Gokhale
Abstract
Severe accident analysis of a reactor helps in emergency planning and evolution of Severe Accident Management Guidelines (SAMG). Actions recommended in the SAMG aim at arresting accident progression and limiting significant radioactive release. However, success of these SAMG actions needs to be assessed with respect to the evolution of accident. Analysis of consequences of injection of water into the reactor pressure vessel from bottom only as a SAMG action has been carried out for VVER-1000 (V320) reactor. The analysis shows that the success of this SAMG action depends not only on the state of core degradation at the time of injection, but also on the highest temperature reached in the reactor core at the time of injection as well as the availability of steam in the RPV.
Kurzfassung
Die Analyse schwerer Störfälle von Reaktoren hilft bei der Notfallplanung und der Entwicklung von Handbüchern für mitigative Notfallmaßnahmen (SAMG). Maßnahmen, die innerhalb der SAMG empfohlen werden, zielen auf die Vermeidung einer Störfallausbreitung und die Begrenzung einer signifikanten Freisetzung von Radioaktivität. Der Erfolg von SAMG-Maßnahmen muss jedoch hinsichtlich seiner Störfallentwicklung abgeschätzt werden. Eine Analyse der Konsequenzen der Wassereinspeisung von unten in den Reaktordruckbehälter (RDB) wurde für einen WWER-1000 (V320) durchgeführt. Die Analyse zeigt, dass der Erfolg dieser SAMG-Maßnahme nicht nur vom Zustand der Kernzerstörung zum Zeitpunkt der Einspeisung abhängt, sondern auch von der höchsten Temperatur, die im Kern zum Zeitpunkt der Einspeisung erreicht wird, und von der Verfügbarkeit von Dampf im RDB.
References
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© 2013, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor
- Analytical study on degraded core quenching
- Experimental investigations on control of flow instability in single-phase natural circulation loop
- Burnup calculations using serpent code in accelerator driven thorium reactors
- Investigation of neutronic effects in structural material of a hybrid reactor by using the MCNPX Monte Carlo transport code
- Nuclear aspects and cyclotron production of the positron emitter 55Co
- Calculation of age-dependent effective doses for external exposure using the MCNP code
- Effect of Cu2+/Al3+ mole ratio on structure of Cu – Al bimetallic nanoparticles prepared by radiation induced method
- A numerical method for resonance integral calculations
- Computational modeling of monoenergetic neutral particle inverse transport problems in slab geometry
- Effects on criticality of selected scattering phase functions in neutron transport equation using the Chebyshev approximation
- U1 and P1 approximations to neutron transport equation for diffusion length calculation
- Technical Notes/Technische Mitteilungen
- TN approximation for the critical size of one-speed neutrons in a slab with anisotropic scattering
- Albedo and constant source problems for extremely anisotropic scattering
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor
- Analytical study on degraded core quenching
- Experimental investigations on control of flow instability in single-phase natural circulation loop
- Burnup calculations using serpent code in accelerator driven thorium reactors
- Investigation of neutronic effects in structural material of a hybrid reactor by using the MCNPX Monte Carlo transport code
- Nuclear aspects and cyclotron production of the positron emitter 55Co
- Calculation of age-dependent effective doses for external exposure using the MCNP code
- Effect of Cu2+/Al3+ mole ratio on structure of Cu – Al bimetallic nanoparticles prepared by radiation induced method
- A numerical method for resonance integral calculations
- Computational modeling of monoenergetic neutral particle inverse transport problems in slab geometry
- Effects on criticality of selected scattering phase functions in neutron transport equation using the Chebyshev approximation
- U1 and P1 approximations to neutron transport equation for diffusion length calculation
- Technical Notes/Technische Mitteilungen
- TN approximation for the critical size of one-speed neutrons in a slab with anisotropic scattering
- Albedo and constant source problems for extremely anisotropic scattering
