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International Journal of Advanced Nuclear Reactor Design and Technology is a peer-reviewed journal published by Elsevier, covering Nuclear reactor physics and engineering, Nuclear Engineering Thermal-Hydraulics, Nuclear Materials and Properties.
| Publisher | Elsevier |
|---|---|
| Reference style | Numbered (Elsevier) Numbered — [1], [2] in the text [1] A. Smith, B. Jones, C. Lee, A representative article title, International Journal of Advanced Nuclear Reactor Design and Technology 12 (2023) 45–58.
Formats any DOI in International Journal of Advanced Nuclear Reactor Design and Technology style. No sign-up. |
| Publishes research in | Nuclear reactor physics and engineering Nuclear Engineering Thermal-Hydraulics Nuclear Materials and Properties Heat Transfer and Boiling Studies Heat transfer and supercritical fluids |
| ISSN | 2468-6050 |
| Citation impact (2-yr) | 0.81 |
| h-index | 14 |
| i10-index | 27 |
| Total citations | 753 |
| Open access | Yes |
| Top institutions publishing here | China General Nuclear Power Corporation (China) |
| Journal website | www.sciencedirect.com |
| You get | A submission-ready PDF and the editable LaTeX source — ready to submit. |
Citations each year’s papers have accumulated so far — the most recent years are still building up.
Gamma radiation shielding properties of baked and unbaked kaolin bricks produced with 0%,10%, 20%, 30%, 40%, 50% of granite have been experimentally and theoretically investigated. A 3 × 3 inches NaI(Ti) detector was employed in the measurement of linear attenuation coefficients. The study concluded that all unbaked samples examined were valid for radiation shielding in…
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 roughness (Ra) of alloy 690 was studied, and a BPNN model for predicting electrochemical corrosion parameters of alloy 690 was established (CP and CCD). The results…
China advanced large-scale Pressurized Water Reactor (PWR) implements series of passive safety systems to provide core cooling and decay heat removal in case of some unexpected accidents. The development of reactor best estimate system analysis codes has realized the transition from conservative safety analysis to Best Estimate plus Uncertainty (BEPU) analysis for the purpose of…
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 reached, while the applied heat flux was gradually increased. Subsequently, the boiling curves derived from the experimental results were divided into partial nucleate boiling (PDB) and…
In order to meet realistic requirements for flexible and reliable power supply in special scenarios such as remote areas and deep ocean exploration, a novel megawatt portable nuclear power system called SUPERHERO (SUPERcritical carbon dioxide cycle-HEat pipe ReactOr power system) is proposed in this paper. The system consists of a heat pipe-cooled reactor and a…
