MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.
Aerosol particles that deposit on surfaces may be subsequently resuspended by air flowing over the surface. A review of models for this liftoff process is presented and compared to available data. Based on this review, a model that agrees with existing data and is readily computed is presented for incorporation into a system level code such as MELCOR. Liftoff Model for MELCOR July 2015 4 This page is intentionally blank
A fission product release and transport model for High Temperature Gas cooled Reactors (HTGRs) is being developed for the MELCOR code. HTGRs use fuel in the form of TRISO coated fuel particles embedded in a graphitized matrix. The HTGR fission product model for MELCOR is being developed to calculate the released amounts and distribution offission products during normal operation and during accidents. The fission product release and transport model considers the important phenomena for fission product behavior in HTGRs, including the recoil and release offission products from the fuel kernel, transport through the coating layers, transport through the surrounding fuel matrix, release into circulating helium coolant, settling and plate-out on structural surfaces, adsorption by graphite dust in the primary system, and resuspension. The fraction of failed particles versus time is input by a particle failure fraction response surface of particle failure fraction as a function offuel temperature, and potentially, fuel burn-up. Fission product release from the fuel kernel and transport through the particle coating layers is calculated using diffusion-based release models. The models account for fission product release from uranium contamination in the graphitized matrix, and adsorption of fission products in the reactor system. The dust and its distribution can be determined from either MELCOR calculations of the reactor system during normal operation, or provided by other sources as input. The distribution of fission products is then normalized using the OR1GEN inventory to provide initial conditions for accident calculations. For the initial releases during an accident, the existing MELCOR aerosol transport models, with appropriate modifications, are being explored for calculating dust and fission product transport in the reactor system and in the confinement. For the delayed releases during the accident, which occur over many hours, and even days fission product release is calculated by combining the diffusion-based release rate with the failure fraction response surface input via a convolution integral. The decay of fission products is also included in the modeling.
A simplified ESBWR MELCOR model was developed to track the transport of iodine released from damaged reactor fuel in a hypothesized core damage accident. To account for the effects of iodine pool chemistry, radiolysis of air and cable insulation, and surface coatings (i.e., paint) the iodine pool model in MELCOR was activated. Modifications were made to MELCOR to add sodium pentaborate as a buffer in the iodine pool chemistry model. An issue of specific interest was whether iodine vapor removed from the drywell vapor space by the PCCS heat exchangers would be sequestered in water pools or if it would be rereleased as vapor back into the drywell. As iodine vapor is not included in the deposition models for diffusiophoresis or thermophoresis in current version of MELCOR, a parametric study was conducted to evaluate the impact of a range of iodine removal coefficients in the PCCS heat exchangers. The study found that higher removal coefficients resulted in a lower mass of iodine vapor in the drywell vapor space.
Due to increasing concerns over the buildup of long-lived transuranic isotopes in spent nuclear fuel waste, attention has been given in recent years to technologies that can burn up these species. The separation and transmutation of transuranics is part of a solution to decreasing the volume and heat load of nuclear waste significantly to increase the repository capacity. A fusion neutron source can be used for transmutation as an alternative to fast reactor systems. Sandia National Laboratories is investigating the use of a Z-Pinch fusion driver for this application. This report summarizes the initial design and engineering issues of this ''In-Zinerator'' concept. Relatively modest fusion requirements on the order of 20 MW can be used to drive a sub-critical, actinide-bearing, fluid blanket. The fluid fuel eliminates the need for expensive fuel fabrication and allows for continuous refueling and removal of fission products. This reactor has the capability of burning up 1,280 kg of actinides per year while at the same time producing 3,000 MWth. The report discusses the baseline design, engineering issues, modeling results, safety issues, and fuel cycle impact.
A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems.
This report describes research and development of methods to couple vastly different subsystems and physical models and to encapsulate these methods in a Java{trademark}-based framework. The work described here focused on developing a capability to enable design engineers and safety analysts to perform multifidelity, multiphysics analyses more simply. In particular this report describes a multifidelity algorithm for thermal radiative heat transfer and illustrates its performance. Additionally, it describes a module-based computer software architecture that facilitates multifidelity, multiphysics simulations. The architecture is currently being used to develop an environment for modeling the effects of radiation on electronic circuits in support of the FY 2003 Hostile Environments Milestone for the Accelerated Strategic Computing Initiative.