Results 71 to 80 of about 1,536,451 (138)

SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

, 1995
The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a ...
Berna, G. A., Allison, C. M., Coryell, E. W.   +2 more
core   +1 more source

Developmental assessment of the SCDAP/RELAP5 code [PDF]

, 1997
The development and assessment of the late-phase damage progression models in the current version (designated MOD3.2) of the SCDAP/RELAP5 code are described.
Slefken, L. J., Coryell, E. W., Harvego, E. A.   +2 more
core  

International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B. International Agreement Report

, 1993
Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and
Schultz, R. R.
core   +1 more source

Experimental and Numerical Simulation Investigation of Single-Phase Natural Circulation in a Large Scale Rectangular Loop

Atom Indonesia, 2019
In order to anticipate station blackout, the use of safety system based on passive features is highly considered in advanced nuclear power plant designs, especially after the Fukushima Dai-ichi nuclear power station accident.
A.R. Antariksawan, S. Widodo, M. Juarsa, S. Ismarwanti, D. Saptoadi, M.H. Kusuma, T. Ardiyati, T. M.I. Mahlia   +7 more
doaj   +1 more source

Analysis of Density Wave Oscillations in Helically Coiled Tube Once-Through Steam Generator

Science and Technology of Nuclear Installations, 2016
Helically coiled tube Once-Through Steam Generator (H-OTSG) is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of ...
Junwei Hao, Yaoli Zhang, Jianxiang Zheng, Zhiwei Zhou, Xuanyu Sheng, Gang Hong, Kai Ye, Ning Li   +7 more
doaj   +1 more source

RELAP5-3D Code Includes Athena Features and Models [PDF]

, 2006
Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen,
Riemke, Richard A., Davis, Cliff B., Schultz, Richard R.   +2 more
core  

Design and thermal-hydraulic simulations of SBO and LOCA on a Nuward-like SMR using RELAP5

, 2023
LAUREA MAGISTRALEIl seguente elaborato tratta in fase introduttiva dei parametri di progettazione dell’E-SMR (modello di un piccolo reattore modulare proposto da un consorzio di aziende nell'ambito del progetto ELSMOR ai sensi del trattato Euratom dell ...
LAMPERTI, ALESSIO
core  

Simulation of Small-Break Loss-of-Coolant Accident Using the RELAP5 Code with an Improved Wall Drag Partition Model for Bubbly Flow

Energies
The RELAP5 code is a computational tool designed for transient simulations of light water reactor coolant systems under hypothesized accident conditions.
Young Hwan Lee, Nam Kyu Ryu, Byoung Jae Kim   +2 more
doaj   +1 more source

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Nukleonika, 2015
The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European ...
Skrzypek Maciej, Laskowski R.
doaj   +1 more source

ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature

Nuclear Engineering and Technology, 2018
An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection ...
Takeshi Takeda
doaj   +1 more source