Development of a sub-channel thermal hydraulic analysis code and its application to lead cooled fast reactor

Li, Shuzhou; Cao, Lina; Khan, Muhammad Saleem; Chen, Hongli

doi:10.1016/j.applthermaleng.2017.02.044

Cited by 24 publications

(5 citation statements)

References 16 publications

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“…After obtaining the power data of each assembly from neutron calculations, the capability of the core to automatically allocate mass flow rate was evaluated by using the subchannel code KMC‐SUB developed by USTC, as shown in Figure .…”

Section: Core Designmentioning

confidence: 99%

“…In this paper, applying the optimization methods presented in Section 2, the core was designed considering the safety design limits (the maximum coolant velocity, the maximum cladding temperature, the minimum coolant temperature, and the maximum burn‐up limited by the cladding radiation damage permitted). Then, the neutronic characteristics were evaluated via Monte Carlo code (MCNP).The subchannel code KMC‐SUBwas used to perform the steady‐state thermal hydraulics (T/H) analysis of the core. With the power distribution and the reactivity coefficients as input, the transient safety characteristics analysis of M 2 LFR‐1000 under typical accidental conditions was conducted via Relap5.…”

Section: Introductionmentioning

confidence: 99%

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Preliminary design of a medium-power modular lead-cooled fast reactor with the application of optimization methods

et al. 2018

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Summary Based on research and development experience from Gen III, Gen III+, and Gen IV reactor concepts, a 1000‐MWt medium‐power modular lead‐cooled fast reactor M2LFR‐1000 was developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing application of optimization methods in preliminary design phase. By using the optimization methods presented, primarily considering the safety design limits (the maximum coolant velocity, the maximum cladding temperature, and the maximum burn‐up limited by the cladding radiation damage permitted), the preliminary design of 1000‐MWth medium‐power modular lead‐cooled fast reactor M2LFR‐1000 was carried out, including the design of fuel rods, fuel assemblies, reactivity control system, primary system, secondary system, decay heat removal system, and so on. The analysis of neutron characteristics (including reactivity feedback coefficients) and thermal hydraulics characteristics (the maximum fuel temperature and the maximum cladding outer surface temperature) of the core under normal steady‐state condition was carried out to evaluate the core design. Also, the analysis of 2 typical protected transients (protected transient over power accident and protected loss of flow accident) was conducted. Other analysis work of the reactor is to be done, such as the transient analysis via computational fluid dynamic codes and the seismic response analysis of the reactor. But the preliminary analysis results obtained so far under normal steady state and transient conditions confirm the inherent safety characteristics of the reactor design.

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Section: Core Designmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Preliminary design of a medium-power modular lead-cooled fast reactor with the application of optimization methods

et al. 2018

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“…The ADS designed in the fourth-generation nuclear power plants for fast nuclear reactors has proposed molten salt ADS, molten alloy ADS, sodium-cooled ADS, and lead-bismuth alloy (LBE)-cooled ADS [11][12][13]. Among them, liquid metal is widely utilized as a coolant material in engineering due to its good thermal conductivity, neutron performance, irradiation resistance, and high heat transfer [14,15].…”

Section: Introductionmentioning

confidence: 99%

Analysis of Factors Influencing Measurement Accuracy for High-Temperature Liquid Metal Flow Sensors in Nuclear Power

et al. 2023

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Permanent magnet metal flowmeters are implemented to monitor the flow of liquid metal coolant in a nuclear reactor, whose measurement accuracy plays a significant role to ensure the safety and normal operation of the nuclear reactor. According to a theoretical analysis of the permanent magnet metal flowmeter, several factors such as temperature, nonlinear degree, and zero potential will affect the accuracy of measuring the flow. However, for a heavy-caliber permanent magnet metal flowmeter, the influence of the magnetic Reynolds number provides obvious nonlinearity, which affects the measurement accuracy of the flowmeter. Consequently, we use the finite element method to calculate the magnetic field of the flowmeter and analyze the cause of the nonlinearity. Additionally, the influence of nonlinear error is significantly reduced by designing the structure of the flowmeter and the appropriate arrangement of the electrodes.

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“…5) ANTEO+ (Lodi et al, 2016a), which can be used to calculate liquid metal cooled fast reactors, including sodium, lead, LBE, and is suitable for a variety of geometric structures such as bare-rod, grid-type spacers, and wire-type spacers. 6) KMC-sub (Li et al, 2017), which adds the relation applicable to liquid LBE and is used to calculate the steady-state condition of SNCLFR-100. ( 7)…”

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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Development of a sub-channel thermal hydraulic analysis code and its application to lead cooled fast reactor

Cited by 24 publications

References 16 publications

Preliminary design of a medium-power modular lead-cooled fast reactor with the application of optimization methods

Preliminary design of a medium-power modular lead-cooled fast reactor with the application of optimization methods

Analysis of Factors Influencing Measurement Accuracy for High-Temperature Liquid Metal Flow Sensors in Nuclear Power

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

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