Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

Y. W. Wu; Xin Li; Xiaolei Yu; S. Z. Qiu; G. H. Su; W. X. Tian

doi:10.1016/j.pnucene.2013.05.001

Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

Y. W. Wu, Xin Li, Xiaolei Yu, S. Z. Qiu, G. H. Su^*, W. X. Tian

^*Corresponding author for this work

Research output: Contribution to journal › Article › peer-review

17 Citations (Scopus)

Abstract

Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO₂-PuO₂ was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions.

Original language	English
Pages (from-to)	65-78
Number of pages	14
Journal	Progress in Nuclear Energy
Volume	68
DOIs	https://doi.org/10.1016/j.pnucene.2013.05.001
Publication status	Published - 2013
Externally published	Yes

Keywords

Sodium-cooled fast reactor
Subchannel code
Thermal hydraulics

ASJC Scopus subject areas

Nuclear Energy and Engineering
Safety, Risk, Reliability and Quality
Energy Engineering and Power Technology
Waste Management and Disposal

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10.1016/j.pnucene.2013.05.001

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title = "Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor",

abstract = "Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2-PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions.",

keywords = "Sodium-cooled fast reactor, Subchannel code, Thermal hydraulics",

author = "Wu, {Y. W.} and Xin Li and Xiaolei Yu and Qiu, {S. Z.} and Su, {G. H.} and Tian, {W. X.}",

note = "Funding Information: This work is supported by the National Magnetic Confinement Fusion Science Program ( 2011GB113001 ), the Special Research Funds for the Doctoral Program of Universities ( 20110201120047 ) and the Fundamental Research Funds for the Central Universities .",

year = "2013",

doi = "10.1016/j.pnucene.2013.05.001",

language = "English",

volume = "68",

pages = "65--78",

journal = "Progress in Nuclear Energy",

issn = "0149-1970",

publisher = "Elsevier Limited",

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T1 - Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

AU - Wu, Y. W.

AU - Li, Xin

AU - Yu, Xiaolei

AU - Qiu, S. Z.

AU - Su, G. H.

AU - Tian, W. X.

N1 - Funding Information: This work is supported by the National Magnetic Confinement Fusion Science Program ( 2011GB113001 ), the Special Research Funds for the Doctoral Program of Universities ( 20110201120047 ) and the Fundamental Research Funds for the Central Universities .

PY - 2013

Y1 - 2013

N2 - Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2-PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions.

AB - Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2-PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions.

KW - Sodium-cooled fast reactor

KW - Subchannel code

KW - Thermal hydraulics

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Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

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Keywords

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