COBRA-IV: the model and the method

Stewart, Charles W.; Wheeler, C.L.; Cena, R.J.; McMonagle, C.A.; Cuta, Judith M.; Trent, D.S.

doi:10.2172/5358588

Cited by 61 publications

(25 citation statements)

References 9 publications

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“…The ASSERT code is originated from the COBRA-IV computer program [8,9]. It has been developed to meet the specific requirements for the thermalhydraulic analysis of two-phase flow in horizontally oriented CANDU fuel bundles.…”

Section: Subchannel Analysismentioning

confidence: 99%

Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

Park

Song

2016

EPJ Nuclear Sci. Technol.

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Abstract. A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF) of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

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Section: Subchannel Analysismentioning

confidence: 99%

Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

Park

Song

2016

EPJ Nuclear Sci. Technol.

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“…This work is mainly focused on development of the safety analysis tool for the molten salt system and the governing equations that are from integral balance laws, a kind of simplification of Navier-stokes equations, are widely used in the safety analysis codes of nuclear system, such as the single channel code, the sub-channel code (Stewart et al, 1977). Since there is no graphite moderator inside the core of MOSART, a one-dimensional single channel model based on integral balance laws on Eulerian control volume is applied and the node scheme is shown in Fig.…”

Section: Thermal Hydraulics Modelmentioning

confidence: 99%

Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

Zhang

Xiao

et al. 2013

Annals of Nuclear Energy

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“…Some traditional subchannel codes, such as SAS4A/SASSYS-1 [9] and SUPERENERGY-2 [10], treat the wire effects as solely enhancement of turbulent mixing but ignore the directional cross flow between subchannels. Others, such as COBRA-4 [11], ASFRE [12], and MATRA-LMR-FB [13], apply "forcing function" type models to account for the diversion flow but are limited in their applications to certain validated conditions (flow regime, channel geometry, or operating conditions) and rely on complex coefficients which were derived from fitting certain sets of experimental data.…”

Section: Development Of the 3d Momentum Source Modelmentioning

confidence: 99%

Progress report on development of intermediate fidelity full assembly analysis methods.

Hu¹,

Fanning²

2011

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While high fidelity modeling capabilities for various physics phenomena are being pursued under advanced modeling and simulation initiatives under the DOE Office of Nuclear Energy, they generally rely on high-performance computation facilities and are too expensive to be used for parameter-space exploration or design analysis. Onedimensional system codes have been used for a long time and have reached a degree of maturity, but limit their validity to specific applications. Thus, an intermediate fidelity (IF) modeling method is being pursued in this work for a fast-running, modest-fidelity, whole-core transient analyses capability. The new approach is essential for design scoping and engineering analyses and could lead to improvements in the design of the new generations of reactors and to the reduction of uncertainties in safety analysis.This report summarizes the initial effort on the development of the intermediatefidelity full assembly modeling method. The requirements and the desired merits of the IF approach have been defined. A three-dimensional momentum source model has been developed to model the anisotropic flow in the wire-wrapped rod bundle without the need to resolve the geometric details. It has been confirmed that the momentum source model works well if its affecting region is accurately imposed. The validity of the model is further verified by mesh and parameter sensitivity studies. The developed momentum source model, in principle, can be applied to any wire-wrapped bundle geometries and any flow regimes; while the modeling strategy can be applied to other conditions with complex or distorted geometry, such as flow in blocked channels.

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COBRA-IV: the model and the method

Cited by 61 publications

References 9 publications

Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

Progress report on development of intermediate fidelity full assembly analysis methods.

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