Overview of MELCOR 1.8.4: Modeling advances and assessment
- Sandia National Labs., Albuquerque, NM (United States). Modeling and Analysis Dept.
- Innovative Technology Solutions, Albuquerque, NM (United States)
The newly released MELCOR 1.8.4 reactor accident analysis code contains many new modeling features as well as improvements to existing models. New model additions to the MELCOR code include a model for predicting enhanced depletion rates for hygroscopic aerosols and a model for predicting the chemisorption of Cesium to the surfaces of piping. Improvements to existing models include: upgrading the core module (COR) to handle flow redistribution resulting from the formation of core blockages, improving the thermal hydraulics (CVH) coupling with COR to handle flow reversal situations, and upgrading the fission product scrubbing model to incorporate the SPARC90 code. Significant upgrading of the COR package core degradation modeling was also included in the new code release version. New and improved models are described in the following paper. In addition, a number of assessment analyses were recently performed, focusing on demonstrating the new and improved capabilities in the code. Results of assessment calculations demonstrating code performance for aerosol (pool) scrubbing, hygroscopic aerosol behavior, and core degradation and hydrogen production are presented. Finally, ongoing code developments activities beyond MELCOR 1.8.4 are described. These include models for treating iodine behavior in containment sumps, pools, and atmosphere, and plans for implementing reflood models and the attendant effects on accident progression. Further improvements and additions to the core degradation modeling in MELCOR are described, including the implementation of enhanced clad failure models to treat clad ballooning and eutectic interaction with grid spacers, and expansion of the COR package to allow for improved representation of UO{sub 2}-Zr eutectic behavior, improved melt relocation treatment, greater detail in describing aspects of BWR core degradation (fuel channel, bypass, and lower plenum), and more flexibility in modeling other structures in the core such as core plate structures (supporting) and PWR control elements (non-supporting).
- Research Organization:
- Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, DC (United States); Brookhaven National Lab. (BNL), Upton, NY (United States)
- OSTI ID:
- 305913
- Report Number(s):
- NUREG/CP-0162-Vol.1; CONF-9710101-Vol.1; ON: TI98007503; TRN: 99:002833
- Resource Relation:
- Conference: 25. water reactor safety information meeting, Bethesda, MD (United States), 20-22 Oct 1997; Other Information: PBD: Mar 1998; Related Information: Is Part Of Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 376 p.
- Country of Publication:
- United States
- Language:
- English
Similar Records
MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment
MELCOR 1.8.5 modeling aspects of fission product release, transport and deposition an assessment with recommendations.