Tutorial : RELAP5/MOD3 - Reactor Excursion and Leak Analysis Program

RELAP is abbreviation from Reactor Excursion and Leak Analysis Program. This program has been developed by Idaho National Engineering Laboratory (INEL) for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. There are 2 kinds of RELAP from INEL, RELAP5/MOD3 and the newest is RELAP5-3D.RELAP5/MOD3 is a BE (best estimate) system code suitable for the analysis of all transients and postulated accidents in Light Water Reactor (LWR)systems, including both large- and small-break loss-of-coolant accidents (LOCAs) as well as the full range of operational transients.

The one dimensional RELAP5/MOD3 code is based on a non-homogeneous and non-equilibrium model for the two-phase system that is solved by a fast, partially implicit numerical scheme to permit economical calculation of system transients. As a part of the severe accident core meltdown code SCDAP/RELAP5, RELAP5/MOD3 describes the thermal hydraulic aspects of both LWR systems and of the degrading core.

RELAP5/MOD3 can work for these operational system :

• Cray X-MP (UNICOS),
• DEC station 5000 (ULTRIX),
• DEC Alpha Workstation (OSF/1),
• IBM Workstation 6000 (UNIX),
• SUN Workstation (UNIX),
• HP Workstation (UNIX).
• CDC Cyber (NOS/VE),
• IBM 3090 (MVS),
• IBM-PC (DOS)

The picture above is describing the RELAP5/MOD3 execution and process.

DOWNLOAD RELAP5/MOD3 code documentations (PDF file) :

• Volume I :presents modeling theory and associated numerical schemes; 
• Volume II :details instructions for code application and input data preparation; 
• Volume IV : discusses in detail RELAP5 models and corelations; 
• Volume V : presents guidelines that have evolved over the past several years through the use of the RELAP5 code
• Input Deck RELAP 5/MOD3 : describes data deck organization and data card requirements for all problem types allaowed in RELAP5/MOD3.2

Download Journals and Research Articles - RELAP5 Applications (PDF file):

• Natural Circulation in the ATUCHA-I PHWR...
• Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties...
• Validation of RELAP/SCDAPSIM/MOD3.4 for Research Reactor Applications 
• Developing and Validating Severe Accident Management Guidelines Using SAMPSON- RELAP/SCDAPIM/MOD3.4 
• Prediction of Reactor Coolant Pump Performance Under Two-Phase...
• RELAP5/MOD3 Analysis for Hydraulic Load Calculation of the SEBIM ...
• LOCA Analysis and Development of a Simple Computer Code for Refill ... 
• RELAP5 Simulation of the Small Inlet Header Break Test B8604 ... 
• Simulation of Multiple Steam Generator Tube Rupture (SGTR) Event ...

Another references from Amazon :

• Analysis of VTI Test Data on the Behavior of the Heated Rod Temperatures in the Partially Uncovered VVER-440 Core Model Using RELAP5/MOD3.2.2 Gamma (NUREG/IA-0208)
• RELAP5/MOD3 assessment using the semiscale 50% feed line break test S-FS-11 (SuDoc Y 3.N 88:53/0104)
• Assessment of Relap5/Mod2 and Relap5/Mod1-Eur Codes on the Basis of Lobi-Mod2 Test Results: Final Report
• RELAP5/MOD2 calculation of OECD LOFT test LP-FW-01 (SuDoc Y 3.N 88:53/0021)

Source/References:

• HINDAWI
• KAERI
• SCDAP/RELAP