Kyle W. Ross

MELCOR 1.8.5 Simulation of TMI-2 Phase 2 With an Enhanced 2-Dimensional In-Vessel Natural Circulation Model

Phase 2 of the TMI-2 accident (core uncovery and melting) is revisited with the latest release of MELCOR (i.e., Version 1.8.5). An enhanced multi-ring multi-level hydro nodalization of the reactor core and upper plenum was developed to permit calculation of natural convection heat transfer between the core and upper internals. Uncertainties in boundary and initial conditions are investigated with particular attention given to the distribution of liquid in the RCS at the beginning of Phase 2 and to the histories of pressure and level in the steam generators. Special attention given to modeling the boiler sides of the once-through steam generators is presented. The modeling is designed to capture the rapid condensation of RCS vapor that would result from spraying cold auxiliary feedwater directly onto the upper portion of a voided steam generator tube bundle. Presentation of key Phase-2 accident signatures including fuel temperatures and hydrogen generation are presented. Comparisons are made between MELCOR calculations, TMI-2 data, and SCDAP/RELAP simulations. A largely improved MELCOR simulation of TMI-2 Phase 2 is obtained.Copyright

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.