A sub-channel analysis code for advanced lead bismuth fast reactor

Wang, Jun; Tian, Wenxi; Tian, Yan; Su, Guanghui; Qiu, Suizheng

doi:10.1016/j.pnucene.2012.09.010

Cited by 38 publications

(5 citation statements)

References 30 publications

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“…Furthermore, it is easier to couple with system program for overall analysis especially under emergency conditions. 10 In this study, a 2-dimensional thermal-hydraulic analysis program with SIMPLE algorithm was developed based on boundary-fitted coordinates. To expand its application to overall circumstances, 2 kinds of experiments were introduced for validating the accuracy of this program in predicting thermal stratification phenomena, followed by the turbulence models used and the computational results.…”

Section: Discussionmentioning

confidence: 99%

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Validation of a methodology for thermal stratification analysis in sodium-cooled fast reactors

et al. 2018

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Summary Thermal stratification phenomena usually occur in the upper plenum after a scram of sodium‐cooled fast reactors, which should be closed concerned in the fields of structural integrity assessment and residual heat removal capacity. A 2‐dimensional analysis program under cylindrical coordinate was developed for predicting the in‐vessel thermal hydraulics. Non‐orthogonal block‐structured grids were generated to resolve the problems with complex structures. A second‐order discrete scheme based on midpoint rule was applied to the spatial discretization of convection and diffusion terms. Two sets of experiments characterized by distinct shapes of their apparatuses were used for the validation, mainly from the viewpoints of vertical temperature distribution and rising behaviors of the stratification interface. Results show that RANS‐type turbulent models make the significant impacts on different flow regimes. In the calculation of a scaled model with plenty of stagnant sodium in the upper region, the realizable k − ε model (RKE) considering the mean deformation rate gives better outcomes than the standard k − ε model (SKE). For the plant‐type upper plenum with considerable flow rate along the entire height, buoyancy modeling is the crucial issue to follow the upward movement of the interface and the relaxation process of the temperature gradient. In this case, employment of the turbulent Prandtl number reflecting the damping effect by incorporating the local Richardson number well reproduced the experimental results.

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Section: Discussionmentioning

confidence: 99%

“…Comparing with software, self‐developed program possesses efficient and flexible characteristics. Furthermore, it is easier to couple with system program for overall analysis especially under emergency conditions …”

Section: Introductionmentioning

confidence: 99%

Validation of a methodology for thermal stratification analysis in sodium-cooled fast reactors

et al. 2018

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“…The code is developed based on COBRA‐EN. The detailed thermal hydraulic models for sub‐channel analysis code have been introduced by Wang et al In this paper, the modifications of SUBAS code based on COBRA‐EN are introduced in the following aspects. Heat transfer models …”

Section: Descriptions Of Sub‐channel Analysis Code Subasmentioning

confidence: 99%

Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor

et al. 2018

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Summary Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.

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“…The results show that the calculated results are in good agreement with the experimental results. 4) SACOS-PB (Wang et al, 2013a), which is concluded that the hexagonal fuel assembly is the best choice for the LBE fast reactor by analyzing the temperature distribution of the fuel assembly and diverse component types. Wu et al (2021) improved SACOS-PB, and Wang et al (2021) utilized it for flow blockage incident analysis.…”

Section: Introductionmentioning

confidence: 99%

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

Chen,

Qiu,

Zhang

et al. 2023

Front. Energy Res.

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The present study focuses on the thermal hydraulic characteristics of fuel assemblies in the China Initiative Accelerator Driven subcritical System (CiADS), which utilizes grid-type spacers to support lead bismuth eutectic (LBE) cooled fast reactors. While the structural design and mechanical analysis of CiADS triangular fuel assemblies with grid-type spacers have been preliminarily investigated, there is a pressing need to study the thermal hydraulic behavior of liquid LBE within these fuel assemblies. This analysis is crucial for assessing the safety and economy of the CiADS subcritical reactor. The sub-channel analysis code, based on the lumped parameter method, plays a vital role in rapidly evaluating the safety features of LBE-cooled fast reactor fuel assemblies. To comprehensively evaluate the thermal hydraulic performance of fuel assemblies with grid-type spacers in the CiADS LBE cooling system, the flow pressure drop model, turbulent mixing model, and convective heat transfer model for the fuel assembly structure with grid-type spacers were incorporated into the existing CiADS sub-channel analysis code. This enhanced code has been successfully employed in the thermal hydraulic analysis of fuel assemblies utilizing wire-type spacers in the CiADS system. To verify the validity and accuracy of the modified CiADS sub-channel analysis code in calculating triangular fuel assemblies with grid-type spacers, the code was first utilized to replicate a 19-rods flow heat transfer experiment with grid-type spacers cooling liquid LBE. A comparison between the simulation and experimental results confirms that the CiADS sub-channel analysis code can predict the coolant temperature and fuel rod surface temperature in fuel assemblies with grid-type spacers. Subsequently, the structure and thermal hydraulic design of the latest CiADS fuel assembly with grid-type spacers were reviewed. Finally, taking into account the aforementioned enhancements to the CiADS sub-channel analysis code, a comprehensive thermo-hydraulic calculation of the CiADS fuel assembly with grid-type spacers was conducted. The results demonstrate that the coolant outlet temperature, maximum cladding temperature, and maximum fuel pellet temperature all fall within the design parameters of CiADS. Furthermore, the impact of the number of rods in different rod bundles (91, 61, 37, and 7 rods) on thermal hydraulic performance was analyzed. The findings indicate that, while ensuring calculation accuracy and efficiency, the results obtained from the 37-rod bundle can effectively reflect the thermal hydraulic behavior of the reactor core.

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A sub-channel analysis code for advanced lead bismuth fast reactor

Cited by 38 publications

References 30 publications

Validation of a methodology for thermal stratification analysis in sodium-cooled fast reactors

Validation of a methodology for thermal stratification analysis in sodium-cooled fast reactors

Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

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