Results 41 to 50 of about 2,939 (182)

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Alexandria Engineering Journal, 2018
Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-
Z. Tabadar, M. Jabbari, M. Khaleghi, M. Hashemi-Tilehnoee   +3 more
doaj   +1 more source

Improvement and validation of the delayed neutron precursor transport model in RELAP5 code for liquid fuel molten salt reactor

He jishu, 2021
BackgroundThe liquid-fueled molten salt reactor (MSR) is one of the Generation IV advanced reactor concepts, which has unique advantages in aspects of safety, economy and nonproliferation.
LI Rui, CHENG Maosong, DAI Zhimin
doaj   +1 more source

Recent Applications of RELAP5-3D at GRNSPG [PDF]

, 2012
CNA2 : FSAR activities Standard Consolidated Reference Experimental Database MASLWR benchmark OECD benchmarks CHF calculation in low mass flux condition Turbulence effects in Relap5 ...
Veronese, F., Cherubini, M., Petruzzi, A., Fiori, F., Kovtonyuk, A., Gonzalez, R., Kovtonyuk, M., D'AURIA, FRANCESCO SAVERIO, Lanfredini, M., Muellner, N.   +9 more
core  

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

Numerical Study on Laminar-Turbulent Transition Flow in Rectangular Channels of a Nuclear Reactor

Frontiers in Energy Research, 2020
Laminar-turbulent transition flow can be observed in thermal engineering applications, but the flow resistance and heat transfer characteristics are not fully understood.
Zhenying Wang, Rui Zhang, Ting Yu, Zhiyun Liu, Yu Huang, Jianjun Wang, Tenglong Cong   +6 more
doaj   +1 more source

Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1 [PDF]

Nuclear Technology and Radiation Protection, 2011
The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4.
Muhammad Atta, Iqbal Masood, Mahmood Tayyab   +2 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

Assessment of RELAP/SCDAPSIM/MOD3.4 Prediction Capability with Severe Fuel Damage Scoping Test

Science and Technology of Nuclear Installations, 2017
The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure during accident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for current ...
Noppawan Rattanadecho, Somboon Rassame, Kampanart Silva, Chris Allison, Judith Hohorst   +4 more
doaj   +1 more source

Reactivity feedback effect on loss of flow accident in PWR

Nuclear Engineering and Technology, 2018
In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents.
Basma Foad, Salwa H. Abdel-Latif, Toshikazu Takeda   +2 more
doaj   +1 more source

Post-Test Calculations on Steam Cool-Down Test QUENCH-04 with RELAP5, SCDAP/RELAP5, and TRACE (KIT Scientific Reports ; 7577)

, 2011
The capabilities of RELAP5 and TRACE are assessed based on the electrically heated bun-dle test QUENCH-04, performed in our premises and on SCDAP/RELAP5 calculations.
Homann, Christoph, Hering, Wolfgang
core   +1 more source