RSICC CODE PACKAGE PSR-581
1. NAME AND TITLE
SCDAP/RELAP5/MOD3.3: A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident.
Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho through the Nuclear Regulatory Commission, Rockville, MD, USA.
3. CODING LANGUAGE AND COMPUTER
Package ID: P00581MNYCP00
Fortran; PC; Linux; Sun; SGI and Windows (Cygwin) - source code and executables.
Package ID: P00581MNYCP01
Executables only (no source code) for PC; Linux; Sun; SGI and Windows (Cygwin).
Export control regulations restrict the distribution of Fortran source code. If restrictions apply, RSICC will send the executable-only version. Please note that included executables run only on the machines listed below in Section 9 of this abstract.
4. NATURE OF PROBLEM SOLVED
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 and reactor core during severe accidents as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. The coolant system behavior is calculated using a two-phase model allowing for unequal temperatures and velocities of the two phases of the fluid, and the flow of fluid through porous debris and around blockages caused by reactor core damage. The reactor core behavior is calculated using models for the ballooning and oxidation of fuel rods, the meltdown of fuel rods and control rods, fission product release, and debris formation. The code also calculates the heat-up and structural damage of the lower head of the reactor vessel resulting from the slumping of reactor core material. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems.
5. METHOD OF SOLUTION
The code is the result of merging the RELAP5/MOD3 and SCDAP models. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and transport of non-condensable gases. A model is also included in RELAP5 to calculate flow losses in porous debris. Although previous versions of the code have included the analysis of fission product transport and deposition behavior using models derived from TRAP-MELT, this capability has been replaced through a data link to the detailed fission product code, VICTORIA, as a result of an effort to reduce duplicative model development and assessment. The SCDAP models calculate the heatup and damage progression in the core structures and the lower head of the reactor vessel. The calculations of damage progression include calculations of the meltdown of fuel rods and structures, the fragmentation of embrittled fuel rods, convective and radiative heat transfer in porous debris, the formation of a molten pool of core material, and the slumping of molten material to the lower head.
SCDAP/RELAP5 is capable of modeling a wide range of system configurations from single pipes to different experimental facilities to full-scale reactor systems. The configurations can be modeled using an arbitrary number of fluid control volumes and connecting junctions, heat structures, core components, and system components. Flow areas, volumes, and flow resistances can vary with time through either user control or models that describe the changes in geometry associated with damage in the core. System structures can be modeled with RELAP5 heat structures, SCDAP core components, or SCDAP debris models. The RELAP5 heat structures are one-dimensional models with slab, cylindrical, or spherical geometries. The SCDAP core components include representative light water reactor (LWR) fuel rods, silver-indium-cadmium (Ag-In-Cd) and B4C control rods and/or blades, electrically heated fuel rod simulators, and general structures. A two-dimensional, finite element heat conduction model based on the COUPLE code may be used to calculate the heat-up of the lower head of the reactor vessel and the slumped material supported by the lower head. This model takes into account the decay heat and internal energy of newly fallen or formed debris and then calculates the transport by conduction of this heat in the radial and axial directions to the wall structures and water surrounding the debris. The most important use of this model is to calculate the heat-up of the vessel lower head and the timing of its failure in response to contact with material that has slumped from the core region. Other system components available to the user include pumps, valves, electric heaters, jet pumps, turbines, separators, and accumulators. Models to describe selected processes, such as reactor kinetics, control system response, and tracking non-condensable gases, can be invoked through user control.
6. RESTRICTIONS OR LIMITATIONS
7. TYPICAL RUNNING TIME
Dependant on the input files, typically a few minutes.
8. COMPUTER HARDWARE REQUIREMENTS
SCDAP/RELAP5/MOD3.3 runs on Personal Computers, SGI and Sun Workstations and Windows systems under CYGWIN.
9. COMPUTER SOFTWARE REQUIREMENTS
A Fortran compiler is required for source compilation.
RELAP5/MOD3.3 Code Manual Volume I: Code Structure, System Models, and Solution Methods (January 2003).
RELAP5/MOD3.3 Code Manual Volume II: User’s Guide and Input Requirements (January 2003).
RELAP5/MOD3.3 Code Manual Volume III: Developmental Assessment Problems (December 2001).
RELAP5/MOD3.3 Code Manual Volume IV: Models and Correlations (December 2001).
RELAP5/MOD3.3 Code Manual Volume V: User’s Guidelines (December 2001).
RELAP5/MOD3 Code Manual Volume VI: Validation of Numerical Techniques in RELAP5/MOD3.0 (December 2001).
RELAP5/MOD3.3 Code Manual Volume VII: Summaries and Reviews of Independent Code Assessment Reports (December 2001).
RELAP5/MOD3.3 Code Manual Volume VIII: Programmers Manual (December 2001).
L. J. Siefken, E. W. Coryell, E. A. Harvego, and J. K. Hohorst, “SCDAP/RELAP5/MOD3.3 Code Manual Volume 2: Modeling of Reactor Core and Vessel Behavior During Severe Accidents,” NUREG/CR-6150, Revision 2, Volume 2 (September 2000).
L. J. Siefken, E. W. Coryell, E. A. Harvego, and J. K. Hohorst, “SCDAP/RELAP5/MOD3.3 Code Manual Volume 3: User’s Guide and Input Manual,” NUREG/CR-6150, Revision 2, Volume 3 (September 2000).
L. J. Siefken, E. W. Coryell, E. A. Harvego, and J. K. Hohorst, “SCDAP/RELAP5/MOD3.3 Code Manual Volume 4: MATPRO -- A Library of Materials Properties for Light_water_Reactor Accident Analysis,” NUREG/CR-6150, Revision 2, Volume 4 (September 2000).
L. J. Siefken, E. W. Coryell, E. A. Harvego, and J. K.
Hohorst, “SCDAP/RELAP5/MOD3.3 Code Manual Volume 5: Assessment of Modeling of
Reactor Core Behavior During Sever Accidents”, NUREG/CR-6150, Revision 2, Volume 5 (September 2000).
11. CONTENTS OF CODE PACKAGE
The executable-only (P00581MNYCP00) package includes the SCDAP/RELAP5/MOD3.3executables for PC Windows, Linux, Sun and SGI platforms along with the referenced documents above and test cases. The P00581MNYCP01 package includes everything listed above plus Fortran source files.
12. DATE OF ABSTRACT
KEYWORDS: LOCA; HEAT TRANSFER; THERMAL HYDRAULICS; NUCLEAR SAFETY