Validation of the Assert Subchannel Code: Prediction of Critical Heat Flux in Standard and Nonstandard Candu Bundle Geometries

Carver, Michael; Kiteley, J.C.; Zhou, Rongliang; Junop, S. V.; Rowe, D.S.

doi:10.13182/nt95-a35156

Cited by 13 publications

(4 citation statements)

References 4 publications

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“…A subchannel analysis was performed for a 37S fuel bundle with/without the inner pitch length modification using the ASSERT PV code [7]. The ASSERT code is originated from the COBRA-IV computer program [8,9].…”

Section: Subchannel Analysismentioning

confidence: 99%

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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%

“…It has been developed to meet the specific requirements for the thermalhydraulic analysis of two-phase flow in horizontally oriented CANDU fuel bundles. Especially, it is distinguished from COBRA-IV in terms of following features [7]:…”

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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“…For the sensitivity studies of the effect of the inner pitch circle and small center element size on the CHF or dryout power of a fuel bundle, the subchannel analysis was performed using the ASSERT code [9], which was transferred from AECL to KAERI under a technology transfer arrangement (TCA) between KAERI/AECL. The ASSERT code is originated from the COBRA-IV computer program [13,14].…”

Section: Subchannel Analysismentioning

confidence: 99%

“…In this study, the above two approaches which were proposed to enlarge the flow areas of the center subchannels were examined in terms of the CHF or dry-out power enhancement by using the subchannel analysis code, ASSERT [9]. In addition, the thermal characteristics of two kinds of modifications were compared to those of a 37S fuel bundle.…”

Section: Introductionmentioning

confidence: 99%

CHF Enhancement of Advanced 37-Element Fuel Bundles

Park

Jung

Ryu

2015

Science and Technology of Nuclear Installations

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A standard 37-element fuel bundle (37S fuel bundle) has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

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