Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Zhao, Pengcheng; Liu, Zijing; Yu, T.; Xie, Jinsen; Chen, Zhenping; Shen, Chong

doi:10.1016/j.net.2020.05.023

Cited by 11 publications

(4 citation statements)

References 20 publications

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“…In this module, the physical parameters of sample points (such as the effective multiplication factor (k eff ), neutron flux, core power, and burnup) are calculated using the RMC [21], which is a three-dimensional Monte Carlo neutron transport code developed by the Reactor Engineering Analysis Laboratory (REAL), Department of Engineering Physics, Tsinghua University. Some thermal-hydraulic characteristic parameters of lead-bismuth reactors (such as the temperature of the fuel cladding and pellet and coolant velocity) are calculated using STAC [22,23], which is a code that analyzes the steady-state thermal-hydraulic characteristics of a lead-bismuth reactor…”

Section: Intelligent Single-/multiobjective Optimization Methodsmentioning

confidence: 99%

“…In this research, a single-/multiobjective optimization method (termed as RBF-OLHS-NGA/NSGA II method) was constructed for optimizing the design of a leadbismuth reactor core, which involved the RBF surrogate model, orthogonal Latin hypercube sampling (OLHS) [18], and niche/nondominated sorting genetic (NGA/NSGA II) algorithm [19,20]. A design optimization procedure of the lead-bismuth reactor (DOPPLER-R) based on this method was developed, which coupled the reactor Monte Carlo (RMC) [21] code and the steady-state thermal-hydraulic analysis code (STAC) [22,23]. Furthermore, by taking a lead-bismuth reactor as an example (SPALLER-4), the single-/multiobjective optimization method was verified by considering the core fuel loading and active zone volume as the optimization objectives; this reactor was independently designed by one of the authors of this paper.…”

Section: Introductionmentioning

confidence: 99%

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Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

Zijing

Zhao

et al. 2023

International Journal of Energy Research

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Owing to their high miniaturization potential and strong natural circulation capability, lead-bismuth reactors have been widely employed for nuclear energy utilization as well as military and civilian applications. However, there is a need to optimize the design of these reactors. To solve the complex high-dimensional nonlinear single-/multiobjective optimization problem of lead-bismuth reactors, we proposed a single-/multiobjective optimization method (termed as RBF-OLHS-NGA/NSGA II method) for optimizing the reactor core, which is based on the radial basis function (RBF) surrogate model (for prediction), orthogonal Latin hypercube sampling, niche genetic algorithm (for single-objective optimization), and nondominated sorting genetic algorithm (for multiobjective optimization). In view of this approach, the design optimization procedure of the lead-bismuth reactor based on RBF (DOPPLER-R) was developed, which combined the reactor Monte Carlo (RMC) code and the steady-state thermo-hydraulic analysis calculation (STAC) code to sample, predict, and optimize reactor core parameters. Furthermore, the single-/multiobjective optimization approach was verified by considering the core fuel loading and active zone volume of SPALLER-4 as the optimization objectives. The results show that the RBF surrogate model can accurately and rapidly predict the core characteristic parameters of lead-bismuth reactors. When compared to the value calculated using the RMC code, the relative error associated with the predicted effective multiplication factor ( k e f f ) is within ±0.1%. Compared to the unoptimized values, the single-objective optimized fuel loading reduced by 400 kg, multiobjective optimized fuel loading decreased by 455–493 kg (optimization ratio of 78%–84%), and active zone volume reduced by 166362–182888 cm3 (optimization ratio of 72%–79%). These results prove that the proposed optimization method is feasible, exhibits high efficacy, and can provide new technical ideas for single-/multiobjective optimization of multiphysics, multivariable, and multiconstraint lead-bismuth reactors.

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Section: Intelligent Single-/multiobjective Optimization Methodsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

Zijing

Zhao

et al. 2023

International Journal of Energy Research

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“…where ݉ and ݉ሶ show the fuel mass in the core and the fuel mass flow rate respectively, ܿ , represents the average heat capacity of the fuel in the core, ܶ , ܶ , , and ܶ , show the core lumped fuel temperature of the moderator surface, the core inlet temperature, and the core outlet temperature [11,14,16].…”

Section: A Neutronic Calculation Model Developmentmentioning

confidence: 99%

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

Ali

Liu

Raza³

et al. 2021

Eng. Technol. Appl. Sci. Res.

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The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance.

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“…These codes are lack of convincing validation, and no further improvement and extension work has been published. Recently, Zhao et al developed a code for steady‐state thermal‐hydraulic analysis of natural circulation LFRs and preliminarily verified by comparing with MARS calculation results 5 . However, the steady‐state code‐to‐code verification is still not persuasive.…”

Section: Introductionmentioning

confidence: 99%

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Wei

Wang

et al. 2021

Int J Energy Res

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Summary In this paper, development and preliminary validation of an in‐house code LETHAC (lead‐based fast reactor thermal‐hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal‐hydraulic transient performance of a lead‐bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most‐used LBE (lead‐bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN‐II 61‐rod wire‐wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire‐wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead‐bismuth cooled systems.

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Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Cited by 11 publications

References 20 publications

Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

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