assessment of subchannel code assert-pv for

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ABSTRACT. Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry.
International Congress on Advances in Nuclear Power Plants Track 7: Thermal Hydraulics Analysis and Testing Contribution ID: 14354

ASSESSMENT OF SUBCHANNEL CODE ASSERT-PV FOR PREDICTION OF CRITICAL HEAT FLUX IN CANDU BUNDLES Y.F. Rao, Z. Cheng and G.M. Waddington Thermalhydraulics Branch, Atomic Energy of Canada Limited, Chalk River, Ontario, Canada, K0J 1J0 ABSTRACT Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently-released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompass the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

Development of the recently-released ASSERT-PV 3.2 focused on improving code predictions of (i) flow distribution, (ii) dryout power and CHF location, and (iii) post-dryout (PDO) sheath temperature distribution. The significant model changes or additions in ASSERT-PV 3.2 (v3.2) compared to the previous version ASSERT-PV V3R11 are described in Reference [4]. Assessments of the improved capabilities for flow distribution, dryout power and CHF location, and PDO sheath temperature predictions have been completed using experiment data sets including the Stern Laboratories (SL) 28-, 37- and 43-element (CANFLEX) bundle experiments. Significant improvement has been confirmed for all key output parameters and over all three CANDU bundles. References [5] and [6], respectively, present the assessment results for the flow-distribution and PDO sheath temperature predictions, whereas this paper focuses on the assessment of the CHF prediction.

1 INTRODUCTION ASSERT-PV (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics, Pressure-Velocity solution procedure) [1][2][3] is a computer code developed at AECL mainly for thermalhydraulic analysis of CANDU reactor fuel bundles. The code is also capable of modeling other reactor fuel bundles, including PWR and BWR assemblies in vertical or horizontal channels. As well, the code can accommodate a range of fluids, including single-and two-phase heavy water, light water, various Freons, and an air-water mixture [4][5]. The code is based on the subchannel concept widely used for fuel bundle and fuel assembly thermalhydraulic analysis, where subchannels are the coolant flow areas bounded by the fuel elements and imaginary planes linking adjacent element centre lines. The two-phase flow model used in ASSERT-PV is based on an advanced drift-flux model, a five equation model that can consider thermal non-equilibrium and the relative velocity of the liquid and vapour phases. Equations of mass, momentum and energy are transformed into a set of finite-volume equations and solved as an initial-value problem using a fully implicit scheme for each subchannel while taking into account inter-subchannel interactions. The subchannel solutions provide detailed singleand two-phase flow distributions, such as subchannel flows (axial and cross flows), enthalpy or quality, and void fraction. The code also calculates fuel sheath temperature distribution, pressure distribution, and local critical heat flux (CHF), from which dryout power is determined.

Subsequent sections in the paper are arranged as follows: (i) improvement of ASSERT-PV CHF models; (ii) SL CHF experiments; (iii) ASSERT-PV results and accuracy; (iv) sensitivity study of each model change; and (v) conclusions. 2 IMPROVEMENT OF CHF MODEL IN ASSERT-PV 3.2 The ASSERT-PV V3R1 CHF model set can predict dryout power and CHF axial location in CANDU bundles reasonably well, but generally cannot predict CHF radial location as well. A new CHF model set has been developed for ASSERT-PV 3.2. Modification to the CHF models in ASSERT-PV focused on 1

The format of the code version identifier has changed during the course of new code development.

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