SCDAP/RELAP5/MOD 3.3 Code Manual:Assessment of Modeling of Reactor Core Behavior During Severe Accidents (NUREG/CR-6150, Volume 5, Revision 2)

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Publication Information

Manuscript Completed: August 2000
Date Published: January 2001

Prepared by:
L. J. Siefken, E. W. Coryell, E. A. Harvego,
J. K. Hohorst

Idaho National Engineering and Environmental Laboratory
P.O. Box 1625
Idaho Falls, ID 83415-3129

S. A. Arndt, NRC Project Manager

Prepared for:
Division of Systems Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code W6095

Availability Notice

Abstract

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 bockages 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 heatup 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.

The assessment of the MOD3.3 version of SCDAP/RELAP5 showed that its new models for severe fuel damage and thermal hydraulic behavior result in calculated behavior of fuel assemblies under severe accident conditions in agreement with measurements. The new models that were assessed using a wide range of fuel damage experiments and the TMI-2 accident included; (1) integral diffusion model for oxidation of fuel rod cladding, (2) stress-based model for failure of oxide layer containing melted metallic cladding, (3) model for re-slumping of previously slumped and frozen cladding, (4) model for cracking of cladding oxide layer during reflood conditions and affect of cracking on oxygen transport, (5) models for flow losses and heat transfer in porous debris, (6) model for heat transfer in molten pool that has stratified into oxidic and metallic parts, and (7) model for break-up of jets of molten material slumping into a pool of water. The assessment showed that the SCDAP/RELAP5 calculations of the axial distribution in oxidation and meltdown of fuel assemblies and the behavior of fuel assemblies under reflood conditions were improved by the new models in MOD3.3. The assessment showed that MOD3.3 calculates a more rapid onset of severe fuel damage under severe accident conditions than that calculated by the previous version of the code. The assessment also showed that MOD3.3 can analyze both severe fuel damage experiments and nuclear power plants with a single set of models for the phenomena causing damage to fuel assemblies.