Thermal analysis of VWSB-IP1 at Tarapur
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V. Verma
Abstract
High Level Liquid Radioactive Waste (HLLRW) produced during reprocessing of spent fuel from nuclear reactors is encased in the canisters after vitrification. The vitrified waste has high heat generation rate due to decay heat and needs interim storage under surveillance. The waste needs to be cooled continuously until major portion of the decay heat is dissipated. Natural circulation air cooling has been considered to cool the canisters. Canisters are placed in a storage vault and cooled by induced axial flow of air with the help of stack. The capacity of storage vault for Vitrified Waste Storage Block (VWSB) Facility proposed at Integrated Plant-1, Tarapur is designed for interim storage of waste generated of 30 yrs of IP1 plant operation. Canister and concrete temperature should be within the prescribed limit. Parametric studies have been carried out for the relevant parameters such as stack and duct dimensions, plenum height etc. Details canister temperature have been obtained using CFD code CFD-ACE+. Axial and radial temperature variation in the canisters, thimble and ventilation pipe have been evaluated in a location. Effect of natural convection (in air) within the canister and between thimble and canister is also studied. It was found that canister centerline temperature reduces by 20°C.
Kurzfassung
Hochradioaktive flüssige Abfälle fallen während der Wiederaufarbeitung von Brennstäben an. Diese Abfälle müssen bei der Zwischenlagerung kontinuierlich überwacht und gekühlt werden. Dazu soll in Tarapur, Indien, ein Zwischenlager in der Nähe der Anlage Integrated Plant-1 (IP1) errichtet werden. Bei der Auslegung wird von einer 30jährigen Betriebszeit der Anlage IP1 ausgegangen. Der dabei entstehende hochradioaktive flüssige Abfall soll in der sog. Vitrified Waste Storage Block Facility (VWSB) zwischengelagert und gekühlt werden. Es soll gewährleistet werden, dass die Behälter- und Strukturtemperaturen innerhalb eines vorgegebenen Limits bleiben. Dazu wird eine Umluftkühlung mit Luft vorgesehen. In diesem Beitrag werden Untersuchungen zur Berechnung der notwendigen relevanten Geometriedaten des Lagers mit Hilfe des CFD-Programms ACE+ vorgestellt.
References
1 Verma, V.; GhoshA.K.; Venkat Raj, V.; Kakodkar, A.: Thermal Analysis of Solidwaste Storage Surveillance Facility. Paper No. 087, Proceedings of Int. Conf. on Evaluation of Emerging Nuclear Fuel Cycle Systems, GLOBAL-95, France, September 1995Search in Google Scholar
2 Verma, V.; Ghosh, A. K.; Kushwaha, H. K.: Simulated Model Studies for Solid Waste Storage Surveillance Facility. Nuclear Engineering and Design211 (2002) 121–138Search in Google Scholar
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© 2013, Carl Hanser Verlag, München
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- Numerical simulation of turbulent flow mixing inside a square chimney structure of a research reactor
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part I: key areas
- Numerical simulation of turbulent flow mixing inside a square chimney structure of a research reactor
- Thermal analysis of VWSB-IP1 at Tarapur
- Tracer transport modeling with the Alliances platform in the presence of evapotranspiration
- Seasonally gross alpha and beta activity concentration in surface water and sediments in Sır Dam Pond
- Evaluation of radioactive emissions of lignite-fired power plants in Turkey using the Analytic Hierarchy Process
- Technical Notes/Technische Mitteilungen
- Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code
- Application of the Henyey-Greenstein and Anlı-Güngör phase functions for the solution of the neutron transport equation with Legendre polynomials: Reflected critical slab problem
- Effect of gamma-radiation on sorption and precipitation of radionuclides
