COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Wheeler, C.L.; Stewart, Charles W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

doi:10.2172/7359476

Cited by 95 publications

(27 citation statements)

References 1 publication

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“…Thus film flow rate around the rod need not be uniform but these segments of the films around the rod will have mass exchange also as they are not physically separated. However, the process of exchange of liquid film around the rod is a complex phenomenon and there is a strong tendency to achieve uniformity in the film flow rate in the circumferential direction as mentioned by Whalley [17] and Butterworth [18]. Hence, in the present analysis after the individual film flow calculations, the film flow rate is made uniform before proceeding to the next node.…”

Section: Assumptions For the Lfd Analysismentioning

confidence: 90%

“…As discussed earlier, COBRA-IV-I [17] has been used as the driver module to determine the subchannel conditions and the film dryout code (FIDOM-Rod) is used for the film dryout analysis. COBRA-IV-I is a well established code which is used for the thermal hydraulic analysis of the nuclear rod bundle.…”

Section: Description Of the Models Of The Lfd Analysis (Fidom-rod)mentioning

confidence: 99%

“…The constitutive models for the droplet entrainment and deposition rates play an important role for the good accuracy of the dryout prediction and hence the models tested for tube under the BWR operating conditions [7,8] has also been applied to the rod bundle in the present work using the subchannel conditions derived from the subchannel code COBRA-IV-I [17]. The lateral flow rate of the film on the rod is also an important aspect of the dryout modeling and the previous approaches mentioned above generally assume that the cross flow consists of vapor only and liquid contribution is negligible [15].…”

Section: Introductionmentioning

confidence: 99%

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A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

Chandraker

Vijayan

Sinha

et al. 2012

JPES

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The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were also found to perform well for the rod bundle under BWR conditions.

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Section: Assumptions For the Lfd Analysismentioning

confidence: 90%

Section: Description Of the Models Of The Lfd Analysis (Fidom-rod)mentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

Chandraker

Vijayan

Sinha

et al. 2012

JPES

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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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“…CONVERT is a simplified version of SPECSET (Wheeler et al 1976) which provides a similar function for the COBRA-IV (Wheeler et al 1976;Stewart et al 1977) code. CONVERT allows SUPERENERGY to be variably dimensionable without suffering from the constraints of other in-code redimensioning schemes, such as the use of machine dependent programming, argument lists, and singly dimensioned variables.…”

Section: *!(!") « Plen(k)^(n5^cla0oo(k)+oap(k))o25rfrace^pl*(cladmentioning

confidence: 99%

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Basehore¹,

Todreas²

1980

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COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Cited by 95 publications

References 1 publication

A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

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

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

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