Results 61 to 70 of about 2,651,698 (166)

Preliminary thermal-hydraulic analysis of the EU-DEMO Helium-Cooled Pebble Bed fusion reactor by using the RELAP5-3D system code

Fusion engineering and design, 2021
In the frame of the activities promoted and encouraged by the EUROfusion Consortium aimed at developing the EU-DEMO fusion reactor, great emphasis has been placed at a very early stage of the design to incorporate the provisions needed to improve the ...
S. D’Amico, P. A. di Maio, X. Jin, F. A. Hernández González, I. Moscato, Guangming Zhou   +5 more
semanticscholar   +1 more source

Bayesian Uncertainty Quantification of Reflooding Model With PSO–Kriging and PCA Approach

Science and Technology of Nuclear Installations, Volume 2025, Issue 1, 2025.
To improve the process of best estimate plus uncertainty (BEPU) for nuclear safety assessment and calibration of thermal–hydraulic models for error reduction, inverse uncertainty quantification (IUQ) is proposed in recent years to quantify the uncertainty of model parameters in reactor program.
Ziyue Zhang, Dong Li, Nianfeng Wang, Meng Lei, Hesham M.H. Zakaly   +4 more
wiley   +1 more source

Estimating Allowable Operator Action Time for Mitigating Multiple‐Failure Accidents in Korean Advanced Power Reactor

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
In the case of multiple‐failure accidents in light water reactors, it is imperative to demonstrate that a significant fuel damage can be prevented via prearranged mitigation measures performed by operators. However, the time elapsed from the initiation of the event to the implementation of mitigation actions significantly influences the progress and ...
Jia Yu, Han Sol Park, Yeon-Gun Lee, Geng Chen   +3 more
wiley   +1 more source

Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes

Science and Technology of Nuclear Installations, 2015
In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and
Viet-Anh Phung, Pavel Kudinov
doaj   +1 more source

Numerical Analysis of the CIRCE-HERO PLOFA Scenarios

Applied Sciences, 2020
The present work deals with simulations carried out at the University of Pisa by using the System Thermal Hydraulics code RELAP5/Mod3.3 to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone ...
Moscardini Marigrazia, Galleni Francesco, Pucciarelli Andrea, Martelli Daniele, Forgione Nicola   +4 more
doaj   +1 more source

Experimental and Numerical Results of LIFUS5/Mod3 Series E Test on In-Box LOCA Transient for WCLL-BB

Energies, 2021
The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket).
Marica Eboli, Francesco Galleni, Nicola Forgione, Nicolò Badodi, Antonio Cammi, Alessandro Del Nevo   +5 more
doaj   +1 more source

Opposing Mixed Convection Heat Transfer for Turbulent Single‐Phase Flows

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. The objectives of this study are to gain a comprehensive understanding of opposing flow mixed convection heat transfer and to establish the prediction methodology by ...
Kosuke Motegi, Yasuteru Sibamoto, Takashi Hibiki, Naofumi Tsukamoto, Junichi Kaneko, Guojun Yu   +5 more
wiley   +1 more source

Improvement and Validation of System Code for Thermal-hydraulic Analysis of In-pool Passive Residual Heat Removal System

Yuanzineng kexue jishu
After Fukushima nuclear accident, the passive concept is widely adopted in the thermal safety design of advanced nuclear reactors. The in-pool passive residual heat removal system (PRHRS) can also provide security guarantee for the domestic reactor ...
LIAN Qiang1, 2, 3, ZHU Longxiang1, 2, 3, TANG Simiao1, 2, 3, HUANG Tao4, ZHANG Yong4, PAN Liangming1, 2
doaj   +1 more source

Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

Science and Technology of Nuclear Installations, 2014
This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady ...
Po Hu, Paul P. H. Wilson
doaj   +1 more source

Time‐Series Forecasting of a Typical PWR Undergoing Large Break LOCA

Science and Technology of Nuclear Installations, Volume 2024, Issue 1, 2024.
In this work, a machine learning (ML) metamodel is developed for the time‐series forecasting of a typical nuclear power plant response undergoing a loss of coolant accident (LOCA). The plant model of choice is based on the APR1400 nuclear reactor. The key systems and components of APR1400 relevant to the investigated scenario are modelled using the ...
Michal Kaminski, Aya Diab, Doddy Kastanya   +2 more
wiley   +1 more source