Results 111 to 120 of about 1,033,988 (155)

On the use of RELAP5-3D as a subchannel analysis code

Nuclear Engineering and Design, 2010
Abstract The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries.
Matthew Memmott, Jacopo Buongiorno, Pavel Hejzlar   +2 more
openaire   +1 more source

RELAP5-3D User Tools

Nuclear Technology, 2016
To support the functionality of RELAP5-3D for best-estimate reactor simulation code, a variety of utility programs were developed at Idaho National Laboratory.
openaire   +1 more source

Subchannel Analysis of LFR Wire-Wrapped Fuel Bundle with RELAP5-3D

Nuclear Technology, 2023
A computational campaign was carried out at the Department of Astronautical, Electrical and Energy Engineering of Sapienza University of Rome aiming at the assessment of RELAP5-3D© capabilities for subchannel analysis. More specifically, the investigation involved a lead-bismuth-eutectic–cooled wire-spaced fuel pin bundle and compared simulation ...
Cristiano Ciurluini, Vincenzo Narcisi, Ivan Di Piazza, Fabio Giannetti   +3 more
openaire   +1 more source

RELAP5-3D transient modelling for NGNP integrated plant

International Journal of Nuclear Energy Science and Technology, 2014
The High-Temperature Gas-cooled Reactor (HTGR) is designed with outlet temperatures ranging between 750°C and 800°C. These high outlet temperatures enhance the power production efficiency and facilitate a variety of industrial applications. The objective of this study is to understand the response of the primary system to potential transients in the ...
Nolan A. Anderson, Piyush Sabharwall
openaire   +1 more source

RELAP5-3D Analysis of OECD-NEA/NRC BFBT Benchmark

Volume 3: Thermal Hydraulics; Instrumentation and Controls, 2008
OECD-NEA and the NRC organized and sponsored the BWR Fuel Bundle Test (BFBT) Benchmark with the main purpose of assessing sub-channel and Computational Fluid Dynamic codes capabilities in estimating relevant thermal-hydraulics parameters like void fraction and critical power for a BWR boiling channel.
KOVTONYUK A., PETRUZZI A., PARISI C., D'AURIA, FRANCESCO SAVERIO   +3 more
openaire   +2 more sources

Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

ASME 2017 Nuclear Forum, 2017
Condensation of steam on the primary side of a steam generator in a pressurized water reactor (PWR) is one means of removing decay heat during some accident scenarios, including small break loss of coolant accident (SBLOCA). However, when noncondensable gasses mix with steam, it impairs condensation.
Nolan A. Anderson, George L. Mesina
openaire   +1 more source

Assessment of RELAP5-3D for Application on In-Pool Passive Power Removal Systems

Proceedings of the 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference, 2020
This research investigates the capability of RELAP5-3D© to reproduce relevant thermal-hydraulic phenomena which characterize the operation of an in-pool passive power removal system. The activity has been carried out in the framework of a validity benchmark proposed within the OECD/NEA/CSNI/WGAMA “Thermal-hydraulics of passive safety systems in water ...
Narcisi V., Melchiorri L., Giannetti F., Caruso G.   +3 more
openaire   +2 more sources

RELAP5-3D Code Validation for RBMK-1500 Reactors

12th International Conference on Nuclear Engineering, Volume 3, 2004
This paper deals with RELAP5-3D code validation through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A successful best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data.
Evaldas Bubelis, Algirdas Kaliatka, Eugenijus Uspuras   +2 more
openaire   +1 more source

Foreword: Special Issue on the RELAP5-3D Computer Code

Nuclear Science and Engineering, 2016
(2016). Foreword: Special Issue on the RELAP5-3D Computer Code. Nuclear Science and Engineering: Vol. 182, No. 1, pp. iii-iii.
openaire   +1 more source

VVER-1000 Model Assessment Using RELAP5-3D

Nuclear Technology, 2005
Detailed thermal-hydraulic analysis of commercial power reactors requires modeling of complex multidimensional thermal, fluids, and neutronic behavior. One code that has three-dimensional (3-D) thermal-hydraulic and neutronic capabilities is RELAP-3D.
Emilian L. Popov, Graydon L. Yoder, Valeri Velichkov   +2 more
openaire   +1 more source