Modelling of void formation in the subcooled boiling regime in the athlet code to simulate flow instability for research reactors

Hainoun, A.; Hicken, E.; Wolters, J.

doi:10.1016/s0029-5493(96)01233-2

Cited by 82 publications

(4 citation statements)

References 11 publications

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“…As an example, the ATHLET thermal hydraulic code developed at the GRS, Society for Plant and Reactor Safety, was planed to analyse leaks and transients for power reactors. However, to extend the applicability of the code to the safety analysis of research reactors, a model was implemented permitting a description of the thermal-dynamic non-equilibrium effects in the subcooled boiling regime (Hainoun et al, 1996). In the same way, the RELAP5 code has been modified to better simulate the research reactors operation conditions (low pressure, low mass flow rate, low power).…”

Section: Thermal Hydraulic Modelingmentioning

confidence: 99%

Nuclear Power - System Simulations and Operation

Tsvetkov¹

2011

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Statements and opinions expressed in the chapters are these of the individual contributors and not necessarily those of the editors or publisher. No responsibility is accepted for the accuracy of information contained in the published articles. The publisher assumes no responsibility for any damage or injury to persons or property arising out of the use of any materials, instructions, methods or ideas contained in the book.

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Section: Thermal Hydraulic Modelingmentioning

confidence: 99%

Nuclear Power - System Simulations and Operation

Tsvetkov¹

2011

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“…However, to extend the applicability of the code to the safety analysis of research reactors, a model was implemented permitting a description of the thermal-dynamic non-equilibrium effects in the subcooled boiling regime (Hainoun et al, 1996). In the same way, the RELAP5 code has been modified to better simulate the research reactors operation conditions (low pressure, low mass flow rate, low power).…”

Section: Thermal Hydraulic Modelingmentioning

confidence: 99%

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Costa¹,

Reis²,

Silva³

et al. 2011

Nuclear Power - System Simulations and Operation

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“…Some of these codes explicitly include the basic bubble parameters invoked above (nucleation site density, bubble frequency, and departure size) in an attempt to capture the complexity of subcooled boiling, particularly under low-pressure conditions. e.g., 54) This mechanistic approach has also been adopted in certain core codes, e.g., 55) system codes, e.g., 56) and extensively in computational fluid dynamics (CFD) codes. e.g., [57][58][59] The role of nucleation site density and its suppression under convective conditions has also been recognized as important for predicting the interfacial area concentration in two-fluid T-H codes.…”

Section: Factors Controlling Boiling Suppressionmentioning

confidence: 99%

A Separate-Effect-Based New Appraisal of Convective Boiling and its Suppression

Aounallah¹

2008

Journal of Nuclear Science and Technology

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The development of convective boiling heat transfer correlations and analytical models has been based almost exclusively on the knowledge of global heat transfer coefficients, while the predictive capabilities of the correlation constituting components (typically additive convection and boiling) have remained usually elusive. This becomes important when, for example, developing a mechanistic subcooled void model based on wall heat flux partitioning, or when applying a correlation beyond its developmental range. In the latter case, the preponderance of the individual heat transfer mechanisms, through the phenomenon of boiling suppression, can become significantly different, thus leading to uncharted uncertainty extrapolations. An examination of existing experimental data, obtained under fixed hydrodynamic conditions, has allowed the isolation of the boiling heat transfer contribution over a broad range of thermodynamic qualities (0 to 0.8) and mass fluxes (1,100 to 3,900 kg/(m 2 Ás)) for water at 7.2 MPa. Boiling suppression has been quantified, thus providing valuable new insights on the basic functional relationships of boiling in convective flows. This work has allowed a new interpretation and representation of the standard flow ''boiling map'' (Collier's) to be developed. The convection enhancement and boiling suppression components (F and S) of the well-known Chen's correlation-an important constitutive relationship implemented in several best-estimate (realistic) thermal-hydraulics codes-have been individually determined, showing the pitfall of splitting the correlation for mechanistic boiling heat transfer modelling, and the important role of compensating errors in uncertainty extrapolation. An initial attempt to formulate a new correlation, based for the first time on segregated heat transfer components, is also included.

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Modelling of void formation in the subcooled boiling regime in the athlet code to simulate flow instability for research reactors

Cited by 82 publications

References 11 publications

Nuclear Power - System Simulations and Operation

Nuclear Power - System Simulations and Operation

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

A Separate-Effect-Based New Appraisal of Convective Boiling and its Suppression

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