Onset of Nucleate Boiling for subcooled flow through a one-side heated narrow rectangular channel

Al-Yahia, Omar S.; Jo, Daeseong

doi:10.1016/j.anucene.2017.05.014

Cited by 33 publications

(13 citation statements)

References 19 publications

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“…The limiting conditions depend on peak power density in the fuel meat and local coolant conditions such as pressure, quality, and temperature and are related to the nucleate boiling heat transfer condition. The parameters used to verify design limits are the onset of nucleate boiling (ONB) and departure from nucleate boiling (DNB) [2,9,10].…”

Section: Onset Of Nuclear Boiling Andmentioning

confidence: 99%

“…To predict the ONB the code furnishes three different correlations: Bergles and Rohsenow (1964), Jens and Lottes (1951), and one developed for narrow rectangular channels proposed by Al-Yahia and Jo (2017) [10].…”

Section: Onset Of Nuclear Boiling Andmentioning

confidence: 99%

“…Subchannel thermal-hydraulic codes that handle rectangular channels can be used to verify thermal-hydraulic design limits of plate type fuel reactors. Some of the design parameters include fuel centerline temperatures, flow rate conditions, and local heat transfer conditions based on the estimation of onset of nucleate boiling and departure from nucleate boiling limits [10]. Most subchannel analysis codes were developed for rod type fuels and fewer are dedicated to plate type geometry.…”

Section: Introductionmentioning

confidence: 99%

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Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Castellanos-Gonzalez

Moreira

Carajilescov

2018

Science and Technology of Nuclear Installations

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This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

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Section: Onset Of Nuclear Boiling Andmentioning

confidence: 99%

Section: Onset Of Nuclear Boiling Andmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Castellanos-Gonzalez

Moreira

Carajilescov

2018

Science and Technology of Nuclear Installations

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“…In recent years, many experiments had been conduct to investigate the thermal hydraulic characteristics of plate-type fuel elements, including heat transfer characteristics (Lee and Lee, 2001;Wang et al, 2014a;Chen et al, 2015), subcooled boiling characteristics (Li et al, 2013;Wang et al, 2014b;Al-Yahia and Jo, 2017;Song et al, 2017) and critical heat flux (Debortoli et al, 1958;Sudo et al, 1985). Li et al (Li et al, 2013) studied the boiling characteristic of water in narrow rectangular channel at atmospheric pressure.…”

Section: Introductionmentioning

confidence: 99%

Study on Flow Boiling Characteristics in Rectangle Channel After Formation of Blisters

et al. 2021

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Plate-type fuel elements is one of the first fuel structure choice for the novel integrated PWR, however, blisters will appear on the cladding induced by irradiation and fission. In this work CFD method was used to investigate the subcooled boiling characteristic of the water in rectangle channel with round and pillow blisters, the modified RPI model was also proposed, we can draw conclusions as follows: In the channel with round blister, as the blisters will increase the local flow resistance and more fluid will flow through center of the channel. Boiling occurred only in the area near the edges, nearly no vapor appeared at the center of the channel. The boiling region in channel with pillow shape blisters is wider and concentrated between two pillow blisters and downstream of the non-blisters side. The dry out area are both in the downstream region of blisters for the two types of channels.

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“…This is due to that NB would be initiated by less restrictive operating conditions. Furthermore, preventing NB will protect the reactor against the other critical phenomena [5]. Long-time NB shall be avoided as it would produce bad effects on reactor operation such as worst reactivity characteristics, control rod oscillations and reactor instability due to partial loss of water coolant/moderator [5].…”

Section: Introductionmentioning

confidence: 99%

Thermal-hydraulic hot-spot analysis of IRT-5000 nuclear research reactor: Comparative safety evaluation

Hussein¹,

Ezzat²

2019

FME Transactions

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A detailed methodology is presented in this paper for the calculation of nucleate boiling safety margin (NBSM) in nuclear research reactors using a temperature function with three different thermal-hydraulic hot-spot analyses: nominal, cumulative and statistical for normal operating condition and coolant flow variation. A computer simulation program is developed for applying the methodology to the IRT-5000 reactor based on experimental core data. According to cumulative analysis as the overconservative approach, the NBSM at normal operating condition of thermal power 5 MW and coolant velocity 1.672 m/s was 2.3% with reactor power limit 5.13 MW. However, during power or coolant flow trip condition, transient nucleate boiling would occur for short period at reactor shutdown. Of great importance, variation of NBSM with reactor operating condition for the different hot-spot analyses is found to be of strong linear correlation such that one analysis is sufficient to specify the results of the other analyses.

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Onset of Nucleate Boiling for subcooled flow through a one-side heated narrow rectangular channel

Cited by 33 publications

References 19 publications

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Study on Flow Boiling Characteristics in Rectangle Channel After Formation of Blisters

Thermal-hydraulic hot-spot analysis of IRT-5000 nuclear research reactor: Comparative safety evaluation

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