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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2019.1
- 수록면
- 1,487 - 1,503 (17page)
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The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOTemploying the drift-flux model are presented. This code aims at providing an accurate yet fast corethermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targetingmassively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations,and numerical solutions are obtained by applying the Finite Volume Method (FVM) and theSemi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed tosimulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decompositionscheme that involves both radial and axial decomposition to enable highly parallelized execution. TheESCOT solutions are validated through the applications to various experiments which include CNEN 4 4,Weiss et al. two assemblies, PNNL 2 6, RPI 2 2 air-water, and PSBT covering single/two-phase andunheated/heated conditions. The parameters of interest for validation include various flow characteristicssuch as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift,and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data inthe extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. Theexecution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and foran OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluidthree field model and the axial domain decomposition scheme works as well as the radial one yielding asteady-state solution for the OPR1000 core within 30 s with 104 processors
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