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Conference Article: VEC - A transient whole circuit model for AGRs

Abstract

In the development of the safety arguments for AGRs the need for natural circulation of the primary coolant to be demonstrated as an effective diverse means of post-trip heat removal from the core has been identified. Initial analysis was based upon the use of specific fuel channel and boiler models and an iterative solution. In order to improve both the representational accuracy and the speed and flexibility of modelling, a complete representation of the primary gas circuit has been developed under the computer code name VEC. The aim of the model has been to construct an overall description of the primary circuit including the heat source (core), heat sinks (boilers, pressure vessel cooling system) major component thermal inertias and major gas flow paths so that the macroscopic transient behaviour of all of the circuit components can be determined. Although initially established for natural circulation analysis the model has developed into a design tool to investigate a wide range of normal operational and fault conditions. The paper will describe the detailed model which has been developed for Heysham II AGR. The primary coolant flow network, the linking equations and buoyancy flow representation, the heat transfer and heat transport equations and the solution methods will be discussed. The flexibility of problem specification and solution and the ease of use of the programme will be considered. A comparison of the results of the whole circuit model, VEC, with the detailed individual component models will be presented to indicate the acceptability of the whole circuit model in providing results for safety analysis. The role of this whole circuit analysis in the overall presentation of the fault transient safety substantiation will be briefly discussed. Indicative results for fault transient analysis for Heysham II will be presented and the large margins to the safety constraints demonstrated.

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Reference:

Specialists' meeting on safety and accident analysis for gas-cooled reactors. Oak Ridge, TN (USA). 13-15 May 1985

International Atomic Energy Agency, Vienna (Austria)

IAEA-TECDOC--358, pp:165-181