Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
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A. M. Vaidya
Abstract
Three dimensional CFD simulations of the moderator flow in the calandria vessel of a heavy water reactor are performed using the PHOENICS CFD code. The model includes the entire calandria vessel consisting of three shells, calandria tubes and inlet and outlet nozzle openings. The computational model prepared in PHOENICS consists of (a) standard k-∊ turbulence model, (b) PARSOL technique for handling curved objects in cartesian grids and (c) Boussinesq formulation for handling variable density flows. PHOENICS is validated by applying it to three different flow cases. The flow pattern in the calandria vessel under normal operating conditions is obtained through simulation. The effect of the presence of calandria tubes and heat generation on moderator flow pattern is studied. The simulation is also performed for various heat loads and moderator mass flow rates. The maximum temperature achieved by the moderator flow under various heat loads and moderator mass flow rates is obtained.
Kurzfassung
Die dreidimensionale CFD Simulation der Moderatorströmung im Calandria-Behälter eines Schwerwasserreaktors wird mit Hilfe des PHOENICS CFD Codes durchgeführt. Das Modell schließt den gesamten Calandria-Behälter mit ein, bestehend aus drei Ummantelungen, den Calandria-Röhren und den Ein- und Austrittsstutzen. Das PHOENICS-Rechenmodell verwendet (a) ein Standard k-∊ Turbulenzmodell, (b) das PARSOL-Verfahren zur Bearbeitung gekrümmter Objekte in kartesischen Rasterfeldern und (c) die Boussinesq-Approximation zur Behandlung von Strömungen variabler Dichte. Der PHOENICS Code wird validiert durch Anwendung auf drei unterschiedliche Strömungsverhältnisse. Das Strömungsverhalten im Calandria-Behälter unter normalen Betriebsbedingungen wird simuliert. Der Einfluss von Calandria-Röhren und Wärmeerzeugung auf das Strömungsverhalten des Moderators wird untersucht. Die Simulation wird auch für verschiedene Wärmebelastungen und Moderator-Massendurchsätze durchgeführt. Man erhält so die maximale Temperatur der Moderatorströmung bei verschiedenen Wärmelasten und Moderator-Massendurchsätzen.
References
1 Sinha, R. K.; Kakodkar, A.: Design and Development of the AHWR – The Indian Thorium Fuelled Innovative Nuclear Reactor. Nuclear Engineering and Design236 (2006) 683Search in Google Scholar
2 Carlucci, L. N.: Numerical Simulation of Moderator and Temperature Distribution in a CANDU Reactor Vessel. Atomic Energy of Canada Ltd., AECL-7911 (1982)Search in Google Scholar
3 Dharne, S. P.; Gaitonde, U. N.: Analysis of Moderator Flow and Temperature Distribution inside the Calandria of Indian PHWRs. 3rd ISHMT – ASME Heat and Mass Transfer Conference, IIT Kanpur, 1997, p. 1083Search in Google Scholar
4 Pathak, S. M.; Saha, D.; Sinha, R. K.: Experimental Studies on the Sparger for the Introduction of Moderator into the Calandria of MAPS Reactors. 3rd International Conference on Small and Medium Sized Nuclear Reactors, New Delhi, India, August, 1991Search in Google Scholar
5 Manwoong, K.; Seon-Oh, Y.; Hhu-Jung, K.: Analyses on Fluid Flow and Heat Transfer inside Calandria Vessel of CANDU-6 using CFD. Nuclear Engineering and Design236 (2006) 115510.1016/j.nucengdes.2005.10.018Search in Google Scholar
6 Hassan, E. S. F.; Hussein, M. A.: Moderator circulation in CANDU reactors: Alternative approach for the Tube matrix solution, Nuclear Technology88 (1989) 307Search in Google Scholar
7 CHAM Ltd., 2004, PHOENICS Documentation (Version 3.6)Search in Google Scholar
8 Bird, R. B.; Stewart, W. E. and Lightfoot, E. N.: Transport Phenomenon, John Wiley & Sons (SAE) Pte Ltd., Singapore, 1994Search in Google Scholar
9 Patankar, S. V.: Numerical Heat Transfer and Fluid Flow. Hemisphere Publishing Corporation, 1980Search in Google Scholar
10 Vaidya, A. M.; Maheshwari, N. K.; Vijayan, P. K.: Applicability of PHOENICS CFD Code for Solving Flows involving Buoyancy and Inertia Forces in Complex Reactor Geometries. Accepted for publication in 15th Int. Conf. on Nuclear Engg., Nagoya, Japan, 2007Search in Google Scholar
11 Churl, Y.: KAERI /TR-1999/2001, 2001Search in Google Scholar
12 Launder, B. E.; Spalding, D. B.: The Numerical Computation of Turbulent Flows. Comp. Methods in Applied Mechanics and Eng3 (1974) 26910.1016/0045-7825(74)90029-2Search in Google Scholar
13 Biswas, G.; Eswaran, V.: Turbulent Flows: Fundamentals and Modelling, Narora Publishing House, New Delhi, 2000Search in Google Scholar
© 2008, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
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- Summaries
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- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
