Critical flow prediction by system codes – Recent analyses made within the FONESYS network

Lanfredini, M.; Bestion, D.; D'Auria, Francesco Saverio; Aksan, N.; Fillion, Philippe; Gaillard, Pierre; Heo, Jaeseok; Karppinen, Ismo; Kim, K. D.; Kurki, Joona; Liu, L.; Shen, Anwei; Vacher, J.; Wang, D.

doi:10.1016/j.nucengdes.2020.110731

Cited by 8 publications

(4 citation statements)

References 17 publications

(18 reference statements)

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“…Here we limit ourselves to report, Figure 8, the spread of data resulting from TPCF predictions to show the slow progress made in the area during 45 years. The upper and low diagrams report data from 1976, [27], already elaborated in 1980, [28], and 2020, [29], respectively. Both vertical axes…”

Section: Tpcf T-hp 111 In Tablementioning

confidence: 99%

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Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

D'Auria¹,

Aksan²,

Glaeser³

2021

TI-IJES

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116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

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Section: Tpcf T-hp 111 In Tablementioning

confidence: 99%

“…Top: spread of models results in 1976, adapted from[27]. Bottom: spread of code results in 2020, adapted from[29]…”

mentioning

confidence: 99%

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

D'Auria¹,

Aksan²,

Glaeser³

2021

TI-IJES

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“…The inadequate knowledge of different phenomena including two phase critical flow, example as Lanfredini et al (2020); reflood, e.g. Yadigaroglu et al (1992); and counter current flow limitation, e.g.…”

Section: Introductionmentioning

confidence: 99%

Challenges and concerns for development of nuclear thermal-hydraulics

D'Auria

Hassan

2021

Nuclear Engineering and Design

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“…One of the crucial phenomena for Deterministic Safety Analysis (DSA) is the critical flow (choked flow) phenomena [6][7][8][9][10]. Basically, it is a limitation of the flow velocity (flow rate) due to the compressibility of the flowing media for a given pressure difference between the upstream and downstream reservoirs [8].…”

Section: Introductionmentioning

confidence: 99%

A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment

2021

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The paper presents a study of the critical flow phenomena modeling with MELCOR 2.2.15254 severe accident computer code. The Marviken Critical Flow Test number 21 (CFT-21) experiment was selected as a representative critical flow-focused Separate Effect Test (SET). Various modeling aspects were investigated, including the nodalization, model setup, parameters, and sensitivity coefficients. A local-type sensitivity study was performed to analytically identify the significant parameters and assess their impact on the modeling. A dedicated regression-based approach, using standard deviation, was developed to find the best-fit MELCOR modeling parameters. The primary purpose of this work was to determine the appropriate approach to model critical flow with MELCOR 2.2, investigate the model performance, assess the influence of nodalization choices, identify significant sensitivity parameters, and prepare recommendations with an emphasis on best-estimate modeling. An additional outcome was the benchmark of the recent code version with the Marviken test.

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Critical flow prediction by system codes – Recent analyses made within the FONESYS network

Cited by 8 publications

References 17 publications

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

Challenges and concerns for development of nuclear thermal-hydraulics

A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment

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