Validation of the Assert Subchannel Code: Prediction

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May 13, 1995 - bundle. The ASSERT subchannel code has been vali- ... ments; its combination of three-dimensional prediction .... toward a preferred phase distribution. ... under strict version control. ... This makes the experiment less than.
Nuclear Technology

ISSN: 0029-5450 (Print) 1943-7471 (Online) Journal homepage: http://www.tandfonline.com/loi/unct20

Validation of the Assert Subchannel Code: Prediction of Critical Heat Flux in Standard and Nonstandard Candu Bundle Geometries M. B. Carver, J. C. Kiteley, R. Q.-N. Zhou, S. V. Junop & D. S. Rowe To cite this article: M. B. Carver, J. C. Kiteley, R. Q.-N. Zhou, S. V. Junop & D. S. Rowe (1995) Validation of the Assert Subchannel Code: Prediction of Critical Heat Flux in Standard and Nonstandard Candu Bundle Geometries, Nuclear Technology, 112:3, 299-314, DOI: 10.13182/ NT95-A35156 To link to this article: http://dx.doi.org/10.13182/NT95-A35156

Published online: 13 May 1995.

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VALIDATION OF THE ASSERT SUBCHANNEL CODE: PREDICTION OF CRITICAL HEAT FLUX IN STANDARD AND NONSTANDARD CANDU BUNDLE GEOMETRIES

HEAT TRANSFER AND FLUID FLOW

KEYWORDS: subchannel, critical heat flux, rod bundles

M. B. CARVER, J. C. KITELEY, R. Q.-N. Z H O U , and S. V. J U N O P AECL Research, Chalk River Laboratories, Advanced Reactor Development Division Thermalhydraulics Development Branch, Chalk River, Ontario KOJI JO, Canada

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D. S. ROWE Rowe & Associates, Redmond, Washington 98503

Received December 20, 1993 Accepted for Publication November 30, 1994

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed, and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized.

INTRODUCTION The advanced solution of subchannel equations in reactor thermal-hydraulics (ASSERT) code 1,2 has been developed to address the computation of flow and phase distribution within the horizontal subchannels

of Canada uranium deuterium (CANDU) pressurized heavy water reactor (PHWR) fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the fuel bundle. The ASSERT development program contains a number of coordinated experimental and analytical projects, each providing information essential for the central project. This parallel phenomenological investigation is required to ensure that the computer code incorporates the mechanisms necessary to simulate experimentally observed trends. The development strategy permits constructive validation of the computer code accompanied by progressive improvement, particularly in the area of constitutive relationships. The ASSERT code uses the subchannel approach used in the development of the COBRA-IV computer code. 3 Subchannels are defined as the coolant flow areas between fuel elements (rods) bounded by the rods and imaginary lines linking adjacent rod centers. Subchannels are divided axially into control volumes, which communicate axially in the same subchannel, and laterally across fictitious boundaries with control volumes in neighboring subchannels, as shown in Fig. 1. Early subchannel codes, such as COBRA, were designed primarily to model flow in vertical fuel channels and use a homogeneous mixture model of two-phase flow. ASSERT uses an advanced drift flux model that permits the phases to have unequal velocities and unequal temperatures and includes gravity terms that make it possible to analyze phase separation tendencies that may occur in horizontal flow. THERMAL-HYDRAULIC MODEL AND SOLUTION SCHEME The thermal-hydraulic conservation equations used in ASSERT are derived from the two-fluid equations. 4

Geometric Basis for Subchannel Analysis REACTOR CORE SHOWING FUEL CHANNELS

CALANDRIA TUBE

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END PLATE

SUBCHANNEL i (k)

GAP (k)

DISTANCE, 1 X CENTROID TO CENTROID SPECIFICATIONS

FUEL BUNDLE SHOWING SUBCHANNELS

Fig. 1. Definition of subchannel geometry in a CANDU reactor.

ASSERT can solve the resulting equations using either the drift flux (relative velocity) or the homogeneous mixture model with various options for thermal disequilibrium. The closure relationships required are the equations of state and constitutive relationships expressing relative velocity, fluid friction, wall heat transfer, interfacial heat transfer, and thermal mixing in terms of the primary variables, i.e., the pressure and the flow velocity, density, and enthalpy of each phase. The governing equations are given in Table I. Under single-phase conditions, the distribution of flow and temperature is affected by the exchange of

momentum and energy between subchannels. This is caused by two mechanisms: pressure-driven cross flow and turbulent exchange. The pressure-driven cross flow is governed primarily by the geometry and the appropriate assignment of pressure drop form loss factors associated with flow past fuel bundle appendages and across gaps. The need to model turbulent exchange arises from the fact that the discretization of the equations in a subchannel grid requires a simplification of the diffusion terms in the usual three-dimensional momentum equations. In single phase, the assumption is made that turbulent exchange in the gap connecting two

TABLE I Governing Equations in ASSERT* Mixture mass (conservative form) dp

dt

+ V-(pV) = 0 ,

where p = (a p)g + (a p)f = otgPg + afpj (pV) = (apV)g+

{apV)f = pV .

Mixture momentum (conservative form) d(PV) dt

+ V- pVV+

(ap)„(ap) rJ g

+ VP

VrV*

= -F*w + pg .

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Mixture energy (transportive form) „ _ dh p — + pV-Vh + Vdt

(ap)g(ap)f

Uh-hf)v:

= qT -V-{(aq")g+{aq")fr



Phasic energy (transportive form) Liquid: dhf (ap)/-r- + ot

(apV)fVhf

= q':f* -V-(aq")}

+ qf

.

Vapor: dhg (ocp)g-f + (apV)g-Vhg at = Q"g~ V-{