Distributed resistance modeling of wire-wrapped rod bundles

Ninokata, Hisashi; Efthimiadis, A.; Todreas, Neil E.

doi:10.1016/0029-5493(87)90306-2

Cited by 67 publications

(12 citation statements)

References 8 publications

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“…With the hottest fuel pin model one can evaluate Journal of Nuclear Science and Technology 3 the temperature flattening effect taking into account the intra-subassembly flow redistribution and energy exchanges between the hottest sub-channel and average sub-channel in the fuel subassemblies. The mixing coefficient for the energy exchanges caused by the cross flows due to the wrapping wires around the fuel pins has been obtained through a survey analysis by use of the sub-channel analysis code ASFRE [12]. Super-COPD is an improved version of the plant dynamics analysis code applied to the safety analysis for licensing of the Japanese prototype SFR, MONJU [10,13] and its onedimensional flow network model is composed of the flow elements, mixture elements and heat exchanger elements with various functions.…”

Section: One-dimensional Safety Analysis Methodsmentioning

confidence: 99%

Development of natural circulation analysis methods for a sodium cooled fast reactor

Oyama¹,

Endo²,

Doda

et al. 2015

Journal of Nuclear Science and Technology

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A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.

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Section: One-dimensional Safety Analysis Methodsmentioning

confidence: 99%

Development of natural circulation analysis methods for a sodium cooled fast reactor

Oyama¹,

Endo²,

Doda

et al. 2015

Journal of Nuclear Science and Technology

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“…Model. Hydraulic resistance model which was initially proposed by Ninokata et al [17] is developed to account for the existence of the wire wrap spacer in rod bundle. These models are of "distributed resistance" type and add flow resistance term into the axial and transverse momentum equations as a function of subchannel geometry and flow velocity.…”

Section: Wire Wrapmentioning

confidence: 99%

Subchannel Analysis of Wire Wrapped SCWR Assembly

Shan

Liu

Song

et al. 2014

Science and Technology of Nuclear Installations

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Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can obtain a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) the pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump effectively; (3) the wire wrap pitch has significant effect on the flow in the assembly. SmallerHwire/Drodwill result in stronger cross flow a more uniform coolant temperature profile, and also a higher pressure drop.

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“…Momentum exchange between fluid and the solid wall, however, can not be accounted for by the forced crossflow models since these models are based solely on the continuity consideration. Hence an attempt has been successfully made to develop new three-dimensional distributed-resistance wirewrap spacer models which give the axial and transverse momentum equation flow resistance as a function of subchannel geometry and flow velocity (8). The new models are generally enough to cover a wide range of geometrical parameters (P!DR, H!DR) and flow regimes (laminar, turbulent).…”

Section: H Thermohydraulics In Fuel Subassembliesmentioning

confidence: 99%

Transient Thermohydraulics Research Activities Related to LMFBRs at PNC

Shiba¹

1989

Journal of Nuclear Science and Technology

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This paper reviews research activities at the Power Reactor and Nuclear Fuel Development Corp. japan, with emphasis on experimental studies and analytical code development related to transient thermohydraulics in LMFBRs. Much effort in this area has been mainly devoted to studies on low-heat-flux and low-flow phenomena including natural circulation sodium boiling at decay heat levels, mixed convection flow, in-vessel thermal stratification, and plant system dynamics after reactor shutdown.

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Distributed resistance modeling of wire-wrapped rod bundles

Cited by 67 publications

References 8 publications

Development of natural circulation analysis methods for a sodium cooled fast reactor

Development of natural circulation analysis methods for a sodium cooled fast reactor

Subchannel Analysis of Wire Wrapped SCWR Assembly

Transient Thermohydraulics Research Activities Related to LMFBRs at PNC

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