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ISSN: 2468-6050

A general two-group constants estimator for 17x17 PWR assembly configurations using Artificial Neural Networks

In this study, a preliminary general two-group constants predictor using artificial neural networks (ANNs) for pressurized water reactor (PWR) based assembly designs is established. Users can input...

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Drop Analysis and Numerical Simulation of Spent Fuel Transfer Basket

The spent fuel transfer basket is responsible for the transfer and storage of the assemblies, and the fall condition of the basket needs to be considered to ensure the critical safety of the spent fuel...

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A study on data analysis utilizing the K-edge absorption spectroscopy in mixed thorium and uranium solutions

X-ray absorption spectroscopy stands as a non-destructive assay technique commonly employed in nuclear fuel reprocessing plants to analyze the concentration of elements such as uranium,plutonium,thorium.The...

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Research on Applicability of Containment Test Data for HPR1000 Based on Phenomena Scaling

The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests.To...

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Debris filtering efficiency assessment of the fuel assembly bottom nozzle

The debris-induced fretting wear is a key factor to cause the fuel failure in pressurized water reactor (PWR). Assessing the debris filtering efficiency of bottom nozzle is an important way to analyze...

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Preliminary study on the postulated siting accident source term of integrated small reactor

In accordance with the characteristics of integrated small reactor, a study was conducted on the postulated siting accident source term scheme. The principles for selecting accident types were explored...

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Analysis of natural circulation characteristics for pool-type fast reactor system

In the inherent safety design of the pool-type fast reactor, when the normal heat transfer routine of the primary system is lost during accident conditions, the residual heat of the reactor core is...

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Numerical study on PWR fuel rod cladding ballooning and burst behavior with the thermo-mechanical coupling finite element method

The ballooning behavior of fuel rod cladding is an important issue of nuclear reactor, which may initiate a severe nuclear reactor accident. Based on COMSOL multi-physical module, this research established...

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The impact of natural removal on the release of nuclides into the containment after accident

The natural removal coefficient directly affects the amounts of radioactive iodine and aerosols released into the containment after the accident. The radioactivity calculation models in LOCA were established,...

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Thermomechanical and radiological investigations of storage casks for radioactive spent fuel assemblies

In this research, the Monte Carlo particle transport method and COMSOL Multiphysics were utilized to simulate and analyze the TK-13 cask, which contains spent fuel assemblies of the WWER-1000 reactor....

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Review of spatial scale dispersion models (ATDMs) to simulate environmental dispersion and deposition of radionuclides and the overview of GIS coupling with dispersion models

Estimation of the radiological consequences including the calculation of radiation doses to the exposed population requires numerical evaluation of atmospheric radioactive releases from nuclear facilities...

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Study of a multi-factor coupling removal model for gas-borne source terms under jet flow conditions

After a pipe rupture in the reactor steam generator, high-temperature and high-pressure radioactive materials from the primary circuit enter the secondary side of the steam generator via a jet flow....

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24-Month long cycle fuel management research based on accident tolerant fuel

As the installed nuclear power plants steadily increase in China, the demand for 24-month long cycle fuel management is gradually increasing as well. 18-month refueling strategy is mainly applied in...

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An open-source Thermo-hydraulic Uniphase Advection and Convection Solver for Salt Flows (TUAS)

Due to a lack of open-source GUI simulators and corresponding system level simulation libraries for Gen IV Reactors such as the Fluoride Salt Cooled High Temperature Reactor (FHR), a systems level library...

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Experimental study of Portland cement concrete performance at moderately-elevated temperature

Concrete, as the most commonly-used construction materials, is used for thousands of structures, including buildings in nuclear power plants. Some concrete structures in nuclear power plants are subjected...

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Characterization of oxygen transport in a lead-bismuth natural circulation loop

To investigate the oxygen mass transfer characteristics of a lead-bismuth natural circulation loop, numerical simulations of flow, heat transfer, and oxygen mass transfer processes were conducted using...

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Condensation film flowing dynamics on inclined downward facing plate: Experimental and modeling study

In-containment refueling water storage tanks have been equipped as an important coolant source for passive cooling systems in the third-generation reactors. Condensate film collection from the containment...

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Thermal analysis of irradiated secondary neutron source assembly with wet transportation

The irradiated secondary neutron source assembly can be used to start-up reactors. After irradiated, secondary neutron source assembly will produce neutron and γ. So, when used to start-up reactors,...

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Study on the characteristics of aerosol diffusiophoresis deposition on the outer surface of a vertical circular tube

The deposition behavior of aerosols due to diffusiophoresis directly influences the migration and deposition of radioactive materials following an accident. This is critical in analyzing source term...

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Optimization study of lead-based reactor heat exchangers based on Dakota coupling CFD

The lead-bismuth fast reactor is one of the fourth-generation reactor types.The microchannel heat exchangers are used as intermediate heat exchangers, which significantly influence the efficiency of...

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Research on reactor power prediction of nuclear power plant based on multivariate optimization GRU model

In the operation of nuclear power plants, the accurate prediction of power change trends is crucial for ensuring safety and stability. In this work, a ML-GRU-RS method, based on model-agnostic meta-learning...

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Method of out-of-pile high-temperature tests of low-melting materials in conditions of modeling a severe nuclear reactor accident

This article is devoted to the technique for conducting experiments to study the nature of the interaction of low-melting metals with corium in the conditions of simulating a severe nuclear reactor...

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Investigate of electrochemical properties of Nd(II) and Nd(III) in eutectic LiCl–KCl molten salt

Pyroprocessing is one of the promising methods to reprocess the spent nuclear fuel. Neodymium (Nd) constitutes about one third of rare earth in fission products, however, the electrochemical properties...

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Analysis of explosion incidents in nuclear fuel reprocessing facilities and recommendations for their prevention

Human society endeavors to harness nuclear energy for peaceful utilization, intertwining it with global strategies to address pressing issues such as energy demand, environmental preservation, and energy...

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Machine learning for forecasting factory concentrations of nitrogen oxides from univariate data exploiting trend attributes

The development of post-processing technology for spent nuclear fuel is essential to ensuring the sustainable growth of nuclear energy. However, post-processing facilities release copious amounts of...

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