Thermal-hydraulic analysis of FAIDUS design for severe accident mitigation strategy in a medium size sodium-cooled fast reactor

2021 ◽  
Vol 379 ◽  
pp. 111222
Author(s):  
Prasant Kumar Panigrahi ◽  
K. Velusamy
Keyword(s):  
Fast Reactor ◽  
Mitigation Strategy ◽  
Severe Accident ◽  
Medium Size ◽  
Hydraulic Analysis ◽  
Thermal Hydraulic Analysis

scholarly journals Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

2017 ◽  
Vol 2017 ◽  
pp. 1-16 ◽  
Author(s):  
Siniša Šadek ◽  
Davor Grgić ◽  
Zdenko Šimić
Keyword(s):  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
European Code ◽  
Station Blackout ◽  
Computer Codes ◽  
Safety Systems ◽  
Thermal Hydraulic Analysis

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

Modeling of Thermal Hydraulic Characteristics for a LBE-Cooled Fast Reactor Helical Coiled Type Steam Generator

2020 ◽  
Author(s):  
Xueyou Ding ◽  
Qinglong Wen ◽  
Zhiqiang Chen ◽  
Shenhui Ruan ◽  
Cheng Cheng
Keyword(s):  
Steam Generator ◽  
Fast Reactor ◽  
Fluid Model ◽  
Load Condition ◽  
Transfer Model ◽  
Hydraulic Parameters ◽  
Mechanism Analysis ◽  
Hydraulic Analysis ◽  
Analysis Model ◽  
Thermal Hydraulic Analysis

Abstract Lead and Bismuth Eutectic (LBE) cooled fast reactor is attracting attention due to its advantages in safety. A thermal hydraulic analysis model is developed for an LBE cooled fast reactor helical coiled type steam generator, which could be used for predicting thermal hydraulic parameters distribution along the length of the tube. Based on two fluid model, the mathematical heat transfer model includes six regions: single water and vapor, subcooled boiling, saturated boiling, transition boiling and film boiling. In order to describe the heat and mass transfer between phases, the interphase model is presented. The steam generator is simplified with single tube concept and all the flow variables are evaluated at each point of the grid in which the domain is discretized. Full load condition and a postulated scenario, such a flow rate step changed in secondary side, is calculated in this study. The results showed that the changing process of the thermal hydraulic parameters of helically coiled steam generator conforms to the qualitative mechanism analysis results of thermal hydraulic analysis.

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