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Numerical and analytical predictions of nuclear steam generator secondary side flow field during blowdown due to a feedwater line break

Nuclear Engineering and Technology, 2021, 53(3), , pp.1029-1040

2021

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Steam condensation in horizontal and inclined tubes under stratified flow conditions

International journal of heat and mass transfer, 2019-10, Vol.141, p.71-87

2019

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Experimental investigation of droplet entrainment and deposition in horizontal stratified wavy flow

International journal of heat and mass transfer, 2019-12, Vol.144, p.118613, Article 118613

2019

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Void fraction prediction for separated flows in the nearly horizontal tubes

Nuclear Engineering and Technology, 2015, 47(6), , pp.669-677

2015

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Measuring local droplet parameters using single optical fiber probe

AIChE journal, 2019-06, Vol.65 (6), p.n/a

2019

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Improvement of the critical heat flux correlation in a thermal-hydraulic system code for a downward-flow narrow rectangular channel

Nuclear engineering and technology, 2022-10, Vol.54 (10), p.3962-3973

2022

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Non-uniform flow distribution in the steam generator U-tubes of a pressurized water reactor plant during single- and two-phase natural circulations

Nuclear engineering and design, 2004-07, Vol.231 (3), p.303-314

2004

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Partition method of wall friction and interfacial drag force model for horizontal two-phase flows

Nuclear Engineering and Technology, 2022, 54(4), , pp.1495-1507

2022

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Development of an empirical correlation for the onset of flow instability in narrow rectangular channels

Nuclear engineering and design, 2021-04, Vol.375, p.111090, Article 111090

2021

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Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Science and Technology of Nuclear Installations, 2017-01, Vol.2017 (2017), p.1-13

2017

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An explicit approximation of the central angle for the curved interface in double-circle model for horizontal two-phase stratified flow

Nuclear Engineering and Technology, 2024, 56(8), , pp.3139-3143

2024

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Assessment of the Emergency Operating Procedures for Main Steam Line Break Accident of the OPR1000 Nuclear Reactor Using the MARS Code

Taehan Kigye Hakhoe nonmunjip. B, 2023-11, Vol.47 (11), p.575-583

2023

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Flow regime development analysis in adiabatic upward two-phase flow in a vertical annulus

The International journal of heat and fluid flow, 2011-02, Vol.32 (1), p.164-175

2011

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Realistic thermal analysis of the CANDU spent fuel dry storage canister

Nuclear Engineering and Technology, 2023, 55(12), , pp.4597-4606

2023

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Improvement of the subcooled boiling model for the prediction of the onset of flow instability in an upward rectangular channel

Nuclear engineering and technology, 2022-03, Vol.54 (3), p.1126-1135

2022

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Application of the machine learning technique for the development of a condensation heat transfer model for a passive containment cooling system

Nuclear Engineering and Technology, 2022, 54(6), , pp.2297-2310

2022

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Development of Bubble Size Correlation for Adiabatic Forced Convective Bubbly Flow in Low Pressure Condition Using CFD Code

Applied sciences, 2020-08, Vol.10 (16), p.5443

2020

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Numerical prediction of transient hydraulic loads acting on PWR steam generator tubes and supports during blowdown following a feedwater line break

Nuclear Engineering and Technology, 2021, 53(1), , pp.322-336

2021

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Improvement of the MARS subcooled boiling model for a vertical upward flow

Nuclear Engineering and Technology, 2019, 51(4), , pp.977-986

2019

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Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model

Journal of pressure vessel technology, 2017-06, Vol.139 (3)

2017

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Measurement of local wall temperature and heat flux using the two-thermocouple method for a heat transfer tube

Nuclear Engineering and Technology, 2019, 51(7), , pp.1853-1859

2019

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Development of a new condensation model for the nearly-horizontal heat exchanger tube under the steam flowing conditions

International journal of heat and mass transfer, 2014-12, Vol.79, p.876-884

2014

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Development of a special thermal-hydraulic component model for the core makeup tank

Nuclear Engineering and Technology, 2022, 54(5), , pp.1890-1901

2022

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Numerical prediction of a flashing flow of saturated water at high pressure

Nuclear engineering and technology, 2018-10, Vol.50 (7), p.1173-1183

2018

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Evaluation of a Numerical Analysis Model for the Transient Response of Nuclear Steam Generator Secondary Side to a Sudden Steam Line Break

Journal of pressure vessel technology, 2017-06, Vol.139 (3)

2017

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Numerical and analytical predictions of nuclear steam generator secondary side flow field during blowdown due to a feedwater line break

Steam condensation in horizontal and inclined tubes under stratified flow conditions

Experimental investigation of droplet entrainment and deposition in horizontal stratified wavy flow

Void fraction prediction for separated flows in the nearly horizontal tubes

Measuring local droplet parameters using single optical fiber probe

Improvement of the critical heat flux correlation in a thermal-hydraulic system code for a downward-flow narrow rectangular channel

Non-uniform flow distribution in the steam generator U-tubes of a pressurized water reactor plant during single- and two-phase natural circulations

Partition method of wall friction and interfacial drag force model for horizontal two-phase flows

Development of an empirical correlation for the onset of flow instability in narrow rectangular channels

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

An explicit approximation of the central angle for the curved interface in double-circle model for horizontal two-phase stratified flow

Assessment of the Emergency Operating Procedures for Main Steam Line Break Accident of the OPR1000 Nuclear Reactor Using the MARS Code

Flow regime development analysis in adiabatic upward two-phase flow in a vertical annulus

Realistic thermal analysis of the CANDU spent fuel dry storage canister

Improvement of the subcooled boiling model for the prediction of the onset of flow instability in an upward rectangular channel

Application of the machine learning technique for the development of a condensation heat transfer model for a passive containment cooling system

Development of Bubble Size Correlation for Adiabatic Forced Convective Bubbly Flow in Low Pressure Condition Using CFD Code

Numerical prediction of transient hydraulic loads acting on PWR steam generator tubes and supports during blowdown following a feedwater line break

Improvement of the MARS subcooled boiling model for a vertical upward flow

Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model

Measurement of local wall temperature and heat flux using the two-thermocouple method for a heat transfer tube

Development of a new condensation model for the nearly-horizontal heat exchanger tube under the steam flowing conditions

Development of a special thermal-hydraulic component model for the core makeup tank

Numerical prediction of a flashing flow of saturated water at high pressure

Evaluation of a Numerical Analysis Model for the Transient Response of Nuclear Steam Generator Secondary Side to a Sudden Steam Line Break

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