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ISSN: 2468-6050

Assessment of radiological safety and emergency response of VVER-1200 type reactor

We studied the radiological safety and emergency response of the VVER-1200 type nuclear power plant using the RASCAL 4.3 and HOTSPOT 3.1.2 codes for INES levels 5, 6, and 7 reactor accidents due to...

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Preliminary study on FROBA-SCO2 with transient models and its application for Reactivity Initiated Accident and loss of flow accident

As a coolant enabling the reactor with a high efficiency and low volume, the supercritical CO2 has becoming a hot topic nowadays. However, the fuel rod behaviors under accidents are not studied thoroughly,...

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Calculation of neutron flux and neutron visualization in a spherical reactor core applying a conceptual approach and drawing a comparison with the results obtained by the CITATION code

In this research, a conceptual approach is presented to obtain the neutron flux at a reactor core. A simple spherical reactor core was analyzed by a mathematical model defined by Delphi programming...

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Experimental Study of Turbulent Bubbly Flow in 180-Degree Elbow by PIV-PS Technique

The curvature effect and bubbles are the key factors that affect the gas-liquid two-phase flow and turbulent characteristics in the 180-degree elbow. Limited by the difficulty of experimental measurement...

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Improvement of a design of the molten salt fast reactor

This study provides an improved design of molten salt fast reactors to reduce the maximum magnitude velocity in the core. The results indicated that the improved design reduces the magnitude velocity...

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Steady-state thermal fluids analysis for the HTR-PM equilibrium core

The high temperature gas-cooled reactor-pebble bed module (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is a demonstration project of nuclear...

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Determination of the neutron and gamma ray dose rates of the irradiating beam tubes of a pool-type research reactor by using Monte Carlo simulation and experimental detectors regarding radiation protection issues

In this paper, the reactor core, irradiating beam tubes, and radiological shields of a 5 MW open pool type Material Testing Reactor (MTR) are simulated in detail by using the MCNPX 2.6 code and its...

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Experimental study on flow characteristics of rectangular narrow channel

Plate fuel elements is gradually used in advanced reactor structure design because of its advantages of high fuel consumption and high safety. The flow channel of the plate fuel element is a typical...

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Thermal-hydraulic performance evaluation of annular fuel based on modified FROBA-ANNULAR

In this study, based on the modification of the pellet cladding mechanical interaction, gap closure process, and the cladding water corrosion model, Fuel Rod Behavior Analysis-ANNULAR (FROBA-ANNULAR)...

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Reliability assessment methods to address fast transient of reactor core

In order to enhance the safety of new advanced reactors, reliability based approach to the design of thermal hydraulic system becomes necessary. In this work, “load exceeds capacity” based approach...

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Optimization of the TRISO fuel particle distribution based on octahedral and icosahedral-based segmentation methods in the pebble-bed nuclear core

The random distribution of TRISO fuel particles in the fuel elements may form local hotspots, thereby increasing the thermal stress on TRISO particle layers that may cause particle failures. Optimization...

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Study on the thermal and geometrical parameters of helical coil once-through steam generator system

Due to the special structure, helical coil once-through steam generators (HCOTSGs) can withstand greater thermal expansion stress, have a larger heat exchange area, and cause the secondary flow phenomenon...

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Statistical research on 2/3 of the first core nuclear fuel cost in nuclear power plants

In this present article, a systematical research and analysis is newly introduced on traditional investment estimation of nuclear power plants, where 2/3 of the first core nuclear fuel cost is incorporated...

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Improvement and evaluation of ATWS Protective Signal and mitigation system for ACPR1000 nuclear power plant

At the event of Loss of Normal Feed Water-An Anticipated Transient without Scram (LOFW-ATWS) in ACPR1000 unit, there exists the primary overpressure risk probability when the regulation function of...

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Study on startup characteristics of prototype once-through steam generator for China fast reactor

Once-through steam generator (OTSG) is a vital heat transfer equipment of China Fast Reactor-600 (CFR-600), and its startup characteristics are of great importance to the safe operation of nuclear power...

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IVR design and review of advanced NPPs in China: Issues and perspectives

The In-Vessel Retention (IVR) is an attractive strategy which could avoid the failure of Reactor Pressure Vessel (RPV) and consequent phenomenon in severe accident of nuclear power plants (NPPs). Most...

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Dimensional analysis on incipient flow instability in one-side high heat loaded rectangular flat heat sink under sub-cooled flow boiling conditions

The onset of flow instability (OFI) is a major threat to the operation of a fusion power plant. In particular, because the divertor is loaded with a high heat flux of up to 20 MW/m2, it may be vulnerable...

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Working properties of GEM-TPC at low pressure

Despite the fact that nuclear technology has already been widely used in the energy sector, the research in this regard have been kept on worldwide. The precise detection of fission fragment tracks...

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Heat-transfer characteristics of screw tube in one-side high heat load condition for fusion reactor divertor application

In this study, the heat transfer characteristics of one-side heated screw tubes were analyzed using subcooled flow boiling experiments. The experiments were performed until the critical heat flux was...

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Simulation study on thermal characteristics of primary coolant system of lead-bismuth cooled reactor

According to the structural characteristics of the lead-bismuth (LBE) cooled reactor and the flow and heat transfer characteristics of the coolant, a real-time simulation model of the primary coolant...

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Development and evaluation of fuel performance analysis code FuSPAC

The nuclear fuel rod should maintain self-sustainable nuclear fission, as the first safety barrier to prevent the leakage of radioactive products. The fuel rod performance analyses are essential to...

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Influence of zinc injection on deposition of corrosion products on inner wall of heat transfer tube

The zinc injection process has been widely used in the primary coolant of PWR nuclear power plants. This study explored its effect in the water coolant of fusion reactors. The research uses the International...

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Virtual reactor laboratory integrated with cyber security from Bangladesh perspective

This paper intends to present the current status of the education and training program of BAEC TRIGA Research Reactor (BTRR) as well as the prospect of Virtual Reactor Laboratory (VRL) as a tool to...

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Pressure drop characteristics in single-side heated circular smooth channel under sub-cooled flow boiling conditions

In this study, the two-phase pressure drop for a one-side heated smooth channel was explored through subcooled flow boiling experiments for divertor applications. When the channel heat load was gradually...

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Heat transfer characteristics of single-layer and two-layer corium pool in elliptical lower head

In this study, we investigated the characteristics of natural convection heat transfer of single-layer and two-layer corium pool in the elliptical lower head by the numerical method. Numerical simulations...

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