Results 51 to 60 of about 1,865 (177)

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Science and Technology of Nuclear Installations, 2017
Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions.
Eltayeb Yousif, Zhijian Zhang, Zhaofei Tian, Hao-ran Ju   +3 more
doaj   +1 more source

Development and Validation of Multiscale Coupled Thermal-Hydraulic Code Combining RELAP5 and Fluent Code

Frontiers in Energy Research, 2021
In nuclear reactor system research, the multiscale coupled thermal-hydraulic (T-H) system code and CFD code is one of the most prevalent research areas, and it could help improve simulation fidelity and optimize nuclear reactor design.
Lin Sun, Minjun Peng, Genglei Xia, Xuesong Wang, Mingyu Wu   +4 more
doaj   +1 more source

Development and verification of code for simulation of vapor-liquid two phase 2D pressure wave propagation

He jishu, 2023
BackgroundThe propagation of pressure waves in nuclear energy systems will cause hydraulic load effects, and it is particularly important for the analysis of structural loads to accurately simulate the propagation process of pressure waves.
GONG Zhengyu, GU Zhixing, PAN Qiwen, ZHANG Muhao, DAI Jianing   +4 more
doaj   +1 more source

SCDAP/RELAP5 independent peer review [PDF]

, 1993
The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3.
Jenks, R.P. (Los Alamos National Lab., NM (United States)), Corradini, M.L. (Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering), Kelly, J.E. (Sandia National Labs., Albuquerque, NM, Haste, T.J. (AEA Technology, Winfrith (United Kingdom)), Dhir, V.K. (Dhir, (V.K.) Santa Monica, CA (United States)), Heames, T.J. (Science Applications, Inc., Albuquerque, NM (United States))   +5 more
core   +3 more sources

Preliminary experimental validation of multi-loop natural circulation model based on RELAP5/SCDAPSIM/MOD 4.0

International Journal of Advanced Nuclear Reactor Design and Technology, 2020
Passive safety systems mostly working on the principles of natural circulation are given much priority in increasing the reliability of reactors’ inherent safety features.
Abbati Zahraddeen, Jiarui Chen, Kun Cheng, Fulong Zhao, Sichao Tan   +4 more
doaj   +1 more source

Analysis With Different Containment Models for Containment Response of NHR200II Under Loss‐of‐Coolant Accident

Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.
In the design of nuclear power plant systems, the thermal–hydraulic response in the containment during a loss‐of‐coolant accident (LOCA) plays a crucial role. In this study, the advanced small modular reactor NHR200‐II is selected as the research object.
Yan Wang, Yifan Meng, Zixuan Wang, Mohammad Rezwan Habib   +3 more
wiley   +1 more source

Input Calibration and Validation of RELAP5 Against CIRCUS-IV Single Channel Tests on Natural Circulation Two-Phase Flow Instability

Science and Technology of Nuclear Installations, 2015
RELAP5 is a system thermal-hydraulic code that is used to perform safety analysis on nuclear reactors. Since the code is based on steady state, two-phase flow regime maps, there is a concern that RELAP5 may provide significant errors for rapid transient ...
Viet-Anh Phung, Pavel Kudinov, Dmitry Grishchenko, Martin Rohde   +3 more
doaj   +1 more source

Natural Convection Heat Transfer Analysis With Different DHX Operation Conditions of Sodium‐Cooled Fast Reactor Using ATHLET Code

Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.
Sodium‐cooled fast reactor (SFR) is one of the fourth‐generation nuclear reactor types, which has inherent safety features such as low corrosion and no severe interaction between molten fuel and sodium. However, the severe accident of core melting in SFR needs to be considered.
Yingying Huo, Qing Yang, Xiaojing Liu, Arnab Biswas   +3 more
wiley   +1 more source

ICONE-22450 RELAP5-3D CODE APPLICATION FOR RBMK-1500 REACTOR CORE ANALYSIS [PDF]

, 2020
The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important.
Evaldas Bubelis, Algirdas Kaliatka, Eugenijus Uspuras   +2 more
core  

New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

Science and Technology of Nuclear Installations, 2016
The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration.
P. Balestra, F. Giannetti, G. Caruso, A. Alfonsi   +3 more
doaj   +1 more source