Development of a subchannel analysis code for SFR wire-wrapped fuel assemblies

Sun, Rulei; Zhang, Dalin; Yu, Liang; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.pnucene.2017.12.005

Cited by 19 publications

(3 citation statements)

References 15 publications

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“…Specific model for each regime part and dividing criterion between them were also proposed. A sub‐channel code Subchannel Analysis Code (SUBAC) 12 applicable to wire‐spaced assembly was referred in this study.…”

Section: Methods For Improving Simulation Of Inter‐assembly Heat Transfermentioning

confidence: 99%

Development a methodology for evaluating inter‐assembly heat transfer effect through reactor core in system safety analysis of sodium‐cooled fast reactor

Chen

Zhang

et al. 2021

Int J Energy Res

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Summary Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly.

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Section: Methods For Improving Simulation Of Inter‐assembly Heat Transfermentioning

confidence: 99%

Development a methodology for evaluating inter‐assembly heat transfer effect through reactor core in system safety analysis of sodium‐cooled fast reactor

Chen

Zhang

et al. 2021

Int J Energy Res

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“…The results obtained for the high-, medium-, and low-flow-rate validation cases are presented in Figure 11. As in ORNL's 19-pin case, we compared the results obtained with the present code with the ones obtained in the experiments and SUBAC code [35]. We have selected SUBAC for the code-to-code comparison since it is, to our knowledge, the only subchannel code for wirewrapped SFRs, with openly available results, that presented the best agreement between the code predictions and experiment measurements for the current benchmark.…”

Section: Toshiba's 37-pin Benchmarkmentioning

confidence: 96%

“…Subchannel efforts are new in Pronghorn when compared to other subchannel codes that have been developed, improved, and validated for many years. Hence, we thought it appropriate to include the results obtained by the SUBAC [35] and MATRA-LMR [36] codes in our comparison against experiments. We have selected these two codes because they have previously presented openly available results for the ORNL 19-pin benchmark and are, to our knowledge, the most developed subchannel codes for LMFRs.…”

Section: The Configuration and External Heat Fluxes Of This Cartridge...mentioning

confidence: 99%

Development of a Subchannel Capability for Liquid-Metal Fast Reactors in Pronghorn

Tano Retamales,

Schunert,

Kyriakopoulos

et al. 2022

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Numerical study on the thermal stratification characteristics in the upper plenum of sodium-cooled fast reactor (SFR)

Wang

Chen

Zhang

et al. 2020

Annals of Nuclear Energy

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Development of a subchannel analysis code for SFR wire-wrapped fuel assemblies

Cited by 19 publications

References 15 publications

Development a methodology for evaluating inter‐assembly heat transfer effect through reactor core in system safety analysis of sodium‐cooled fast reactor

Development a methodology for evaluating inter‐assembly heat transfer effect through reactor core in system safety analysis of sodium‐cooled fast reactor

Development of a Subchannel Capability for Liquid-Metal Fast Reactors in Pronghorn

Numerical study on the thermal stratification characteristics in the upper plenum of sodium-cooled fast reactor (SFR)

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