Results 1 to 10 of about 1,536,451 (138)

Validation of RELAP5 MOD3.3 code for Hybrid-SIT against SET and IET experimental data

Nuclear Engineering and Technology, 2020
We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with ...
Ho Joon Yoon, Waleed Al Naqbi, Omar S. Al-Yahia, Daeseong Jo   +3 more
doaj   +4 more sources

RELAP5 application for water hammer analysis in different thermal-hydraulic systems

International Journal of Thermofluids, 2022
The water hammer phenomenon investigations employing RELAP5 code with one-dimensional two-fluid six-equation models of water and steam flow were performed and the main results shared in this paper.
Algirdas Kaliatka, Mindaugas Vaišnoras, Mindaugas Valinčius, Tadas Kaliatka   +3 more
doaj   +1 more source

Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS

Energies, 2021
In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in ...
Alessandro Venturini, Marco Utili, Nicola Forgione   +2 more
doaj   +1 more source

Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor [PDF]

Nuclear Technology and Radiation Protection, 2006
RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants.
Bousbia-Salah Anis, Jirapongmed Areeya, Hamidouche Tewfik, White John R., D'Auria Francesco, Adorni Martina   +5 more
doaj   +1 more source

Development and application of multi-scale thermal fluid coupling program for molten salt cooled fast reactor based on RELAP5 and sub-channel program

He jishu, 2022
BackgroundThe natural circulation chloride cooled fast reactor (N3CFR) has the characteristics of simple structure, high inherent safety and good economy. It is an advanced fourth-generation nuclear energy system with development potential.
SONG Shiyang, CHENG Maosong, LIN Ming, DAI Zhimin   +3 more
doaj   +1 more source

Study on the Density Wave Instability in Natural Circulation System under Rolling Conditions

International Journal of Energy Research, Volume 2023, Issue 1, 2023., 2023
In the ocean environment, the thermal fluid system installed on the ship moves with the waves. This unsteady transient motion imposes additional body forces on the liquid. Thereby, the thermohydraulic properties of the thermal fluid system are changed, especially for natural circulation systems with a low driving force.
Houjun Gong, Xiaoxi Zhang, Nan Gui, Yanping Huang, Xingtuan Yang, Shengyao Jiang, Rafael M. Santos   +6 more
wiley   +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

ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate‐Type Heat Exchanger

Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023
This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E‐SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate‐type one coupling ...
Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves, Doddy Kastanya   +8 more
wiley   +1 more source

Thermal-hydraulic analysis of loss-of-cooling accident in spent fuel pool of Bushehr NPP using the RELAP5 and MELCOR [PDF]

مجله علوم و فنون هسته‌ای, 2020
Following the Fukushima Daiichi accident, the simulation of accidents related to the Spent Fuel Pool (SFP) became more important due to the high content of long-lived radionuclides, and lack of the protection by the pressure vessel despite its low decay ...
S. Gol Narges, S.Kh. Mousavian
doaj   +1 more source

Simulations of Core Damage Progression for TMI‐2 Severe Accident Using CINEMA Computer Code

Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023
As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant ...
Rae-Joon Park, Donggun Son, Jun Ho Bae, Sung Won Bae, Bub-Dong Chung, Kwang Soon Ha, Han Zhang   +6 more
wiley   +1 more source