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ISSN: 2468-6050

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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Effect of LaB6 addition on mechanical properties and irradiation resistance of 316L stainless steels processed by selective laser melting

The 316L stainless steels (SS) have become one of the most common structural materials in nuclear system applications due to their excellent physicochemical performance. Selective laser melting (SLM)...

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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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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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Reactor dynamic simulation to analyze possible scenarios after a spurious opening of the safety flapper valve of TRR during the normal operation regarding the inherently safety features and Design Extension Conditions (DEC) by using the RELAP5 code

LOFA is one of the most important PIE in nuclear reactors. For the pool-type research reactor, the safety flapper valve is usually located below the reactor core and have a very important rule on the...

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Low temperature overpressurization analysis for CPR1000

An additional low temperature overpressure protection system which relies on the Pressurizer (PRZ) pressure relief valves has been put forward under the situation in which the Residual Heat Removal...

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Studies on post-dryout heat transfer in R-134a vertical flow

An experimental study of post-dryout heat transfer with the coolant R-134a was performed in a vertical round tube with upward flow. The experiments were conducted in an uniformly heated tube with an...

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Neutrons diffusion variable coefficient advection in nuclear reactors

The analysis of nuclear reactors in dissimilar geometries is an important topic in sciences and engineering. Two approaches are used in literature for homogeneous systems: computational and analytical...

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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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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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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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