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ISSN: 2468-6050

Electrochemical behavior and corrosion rate prediction study of alloy 690

Alloy 690 is a material commonly used in heat transfer tubes of steam generators (SGTs). The electrochemical corrosion behavior in different volumes (V) of NaCl–Na2S2O3 solution for different surface...

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Development of an edge-cloud collaboration framework for fission battery management system

As a “plug-and-play” nuclear system, the technology of fission battery enables nuclear reactor systems to function as batteries. Undoubtedly, advanced intelligent technologies have to be adopted 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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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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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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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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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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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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Numerical study of gas-injection induced pool sloshing behavior using MPS method

When a Core Disruptive Accident (CDA) occurs in a Sodium-cooled Fast Reactor (SFR), as the accident progresses, its core may form a pool of molten fuel. When the molten fuel pool expands, a certain...

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Investigation on radial fuel relocation and its influence on heat split phenomenon of dual-cooled annular fuel element

Radial fuel relocation due to fuel fragmentation can significantly change the heat split, further affecting the thermal performance and safety of dual-cooled annular fuels. Considering the rarity of...

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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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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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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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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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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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A movable boundary model for helical coiled once-through steam generator using preconditioned JFNK method

Helical coiled once-through steam generator (H-OTSG) is the largest and the most complicated heat exchanger in the high-temperature gas-cooled reactor (HTGR), whose performance plays an important role...

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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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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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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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Analysis on containment thermal hydraulic behaviour under passive containment heat removal system operation condition for HPR1000

HPR1000 NPP is designed with Passive Containment Heat Removal System (PCS) following the technical route called active and passive combined concept. The installed PCS system will have effect on the...

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Simulation of hydrogen deflagrations with the lumped parameter code COCOSYS

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated...

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Supersonic shear layer characteristics of two fluids with a splitter plate

Supersonic shear layer development affects mixing between the primary and secondary fluids in a supersonic ejector substantially. The splitter plate plays a key role in shear layer development in the...

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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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Analysis for fuel rod performance under LOCA based on the FROBA-ROD 2.0 code

Loss of Coolant Accident (LOCA) is one of the major design-basis accidents in LWRs. Fuel rod may fail due to the rupture of the cladding. By developing a LOCA simulation module, the steady-state fuel...

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Transient thermal analysis of IVR strategy under a LB-LOCA based on ASTEC code

A Large-Break Loss of Coolant Accident (LB-LOCA) in a generic, 1000 MWe and 3-loop Chinese Pressurized Water Reactor (PWR) with the In-Vessel corium Retention (IVR) strategy was simulated by using ASTEC...

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