[en] Hazards encountered during decommissioning and waste treatment involve different issues, emphasis, and scenarios than encountered during normal facility operations. This paper provides examples from experience in analysis and modeling of diverse facilities, framed in terms of custom phenomena models and their incorporation into integral facility analysis modeling. Models for entrainment of contamination and aerosol behavior are described and applied. The FATE^TM (Facility Flow, Aerosol, Thermal, and Explosion) computer program is described, and example calculations are given for contamination release due to a dust explosion in order to demonstrate the sensitivity to boundary conditions and the use of engineered safeguards. (authors)

[en] Industry-standard thermal hazard screening is an effective, cost efficient approach to quickly obtain the required data typically utilized for safe scale-up of chemical processes and to accommodate changes to process recipes. Such thermal hazard screening is directly relevant to the packaging, transport, and storage of radioactive waste that is or can become chemically reactive. For such waste streams it is vital to identify safe temperature and pressure conditions and quantify adiabatic heat and gas generation rates in order to safely accommodate (or preclude) thermal instability within the waste package or storage facility. This paper illustrates widely-used thermal hazard screening bench-scale techniques that lend themselves to quickly identifying reactive hazards while providing directly scalable data for package/storage facility design. Example data are presented with discussion of how the data are analyzed for application to safe packaging and storage. (authors)

[en] The FATE^TM code is a well-established software tool for simulation of processing, shipping, and interim storage of damaged spent nuclear fuel (SNF) that was used for process design and safety analysis for the Hanford Spent Nuclear Fuel Program (SNFP). This paper describes new model developments that allow FATE to be used for simulation of drying of damaged LWR fuel and, in particular, for simulation of processing, shipping, and interim storage of Fukushima fuel debris. The model is capable of quantifying important tradeoffs in the debris chunk size, process conditions such as temperature and pressure, container size, etc. to optimize process design. The model is also capable of demonstrating proof-of-dryness and to demonstrate the link between observed quantities such as pressure and the amount of residual water. (authors)

[en] This paper describes new capability of the Modular Accident Analysis Program (MAAP) to model cladding and fuel materials other than Zirconium (Zr) and Uranium Dioxide (UO₂) during a severe accident. MAAP is an EPRI developed software that predicts fuel and core responses to severe accidents involving core melt. MAAP can determine the time to reactor clad failure and the extent of its failure, time to core melt and the extent of core degradation, time of vessel failure, and eventual source term release to environment. In the case of Fukushima Daiichi, a station blackout accident (SBO) occurred in which all on and off site power was lost for an extended period. Due to loss of all power for an extended period, core fuel assemblies were uncovered and overheated by decay heat which caused the fuel rod cladding and channel box Zr to react with steam. This reaction accelerated core heat-up, and eventual core melt down, and core relocation, and hydrogen explosion in Fukushima Daiichi. Considering the Fukushima Daiichi severe accident, industry and US Department of Energy are developing and evaluating the advantages of nuclear fuel with cladding other than Zircaloy to minimize or altogether avoid the reactions that occurred at Fukushima Daiichi. The purpose of these new cladding materials is to provide good high temperature material properties and to prevent or reduce hydrogen generation during severe accidents. In addition to the new cladding materials, new fuel material (other than UO₂) is also being considered. In MAAP5.05, a capability to model new cladding and fuel materials is added so that the ATF developers can evaluate responses of their new materials during severe accident. In this paper, responses of SiC cladding material are presented for SBO and TMI-2 cases for PWR and SBO case for BWR and compared with those of Zircaloy cladding. SiC is known to have favorable high temperature properties, including a decomposition temperature of 2545degC. SiC does react with steam at elevated temperatures, however the reaction under the conditions tested is primarily oxidation generating silica (SiO₂) scale and volatilization of this SiO₂ scale. By simulating the TMI-2 case for PWR and a SBO for both PWR and BWR, the results presented in this paper provide a first approach at predicting the material behavior similar to what occurred at TMI-2 and Fukushima Daiichi and comparisons of accident progressions with two different cladding materials. (author)

[en] A simple turbulent diffusion model is described for predicting late term mixing due to Rayleigh-Taylor instability at an interface that initially separates two semi-infinite regions of fluid of different densities. The one-dimensional, transient species conservation equation is combined with an available vertical dispersion equation which relates the turbulent diffusivity to the local density gradient and a characteristic mixing length. The similarity solution of the species conservation equation yields the well-known long time result for the vertical extent of the mixing region and shows that the dimensionless growth constant that appears in the mixing thickness equation is simply a measure of the size of the characteristic mixing length. A simple polynomial relation is derived for the heavy fluid volume fraction profile within the mixing region which is in close agreement with the profile obtained from a previously reported numerical simulation of Rayleigh-Taylor mixing. It is demonstrated how the model can be readily applied to predict Rayleigh-Taylor mixing in a finite fluid region between top and bottom boundaries.

[en] Improvements to the MAAP 3.0B model for core-concrete interactions, DECOMP, are in progress to create a new version for the accident manage - ment code MAAP 4. This paper describes the first version for MAAP 4, its validation, and future plans. Since DECOMP must fit into an integral code, it contains simplified representations of many phenomena. Significant new features are independent sideward and downward erosion, capacity to incorporate complex pseudo-binary phase diagrams, and benchmark capability using the integral code itself. Code performance for the BETA, SURC, and ACE experiments is promising, but indicates that oxide systems are difficult to model because of the paucity of phase diagram data and the coupling of solidification behavior, viscosity, and heat transfer. excellent agreement for any single test may be obtained by varying the model parameters; the point of this exercise is that a consistent set can be found for all tests. Erosion is fairly well-predicted by the model in all cases. This indicates a good relative balance between upward losses and heat transfer to the walls and floors contacting the debris. For the ACE and SURC simulations, upward losses are sensitive to the imposed temperature boundary conditions because the debris is heated in place and the calculation is initialized when ablation starts. Errors in debris temperature can account for several centimeters error in erosion for any of these tests. The relative erosion error due to temperature error of course decreases for tests of longer duration. The temperature predictions for tests L2 and L5 are considered good, while those for L6 and L7 appear to be low. The temperature prediction for SURC-4 does not follow the increase upon Zr addition but is otherwise good, and the temperature prediction for BETA V5.1 is presumed to be too high because condensed phase Zr and Si does not affect the metal In summary: The core-concrete interaction model within MAAP 3.0B, called DECOMP, is being upgraded for MAAP 4. It contains simplified models to account for phenomena observed during current experiments and yet run fast within an integrated accident analysis. It allows independent sideward and downward erosion and can account for uncertainty in their relative magnitudes. Debris temperature for oxide systems is strongly influenced by solidification behavior which is the most important uncertainty for dry interaction cases. DECOMP contains general pseudo-binary phase diagram capabilities to account for this behavior, and input for this diagram must come from advanced codes. Validation of the model shows acceptable performance and provide s guidance to the parametric heat transfer coefficient which heuristically accounts for viscosity increase with solid fraction. It is expected that better phase diagram data for the oxides will yield better overall results. Future work will consider better data when available, a distributed parameter approach to the debris crusts, and consideration of results obtained in the MACE program

[en] Spent Nuclear Fuel scrap generated while washing the SNF in Hanford's K-Basins to prepare it for cold vacuum drying differed significantly from that envisioned during project design. Therefore, a technical review panel evaluated the new information about the physical characteristics of scrap generated during processing by characterizing it based on measured weights and digital photographic images. They examined images of the scrap and from them estimated the volume and hence the masses of inert material and of large fragments of spent fuel. The panel estimated the area of these particles directly from images and by fitting a lognormal distribution to the relative number particles in four size ranges and then obtaining the area-to-volume ratio from the distribution. The estimated area is 0.3 m2 for the mass of scrap that could be loaded into a container for drying, which compares to a value of 4.5 m2 assumed for safe operation of the baseline process. The small quantity of scrap genera ted is encouraging. However, the size and mass of the scrap depend both on processes degrading the fuel while in the basin and on processes catching the scrap during washing, the latter including essentially unintentional filtration as debris accumulates. Therefore, the panel concluded that the estimated surface area meets the criterion for loading scrap into an MCO for drying, but because it did not attempt to evaluate the criterion itself, it is not in a position to actually recommend loading the scrap. Further, this is not a sufficiently strong technical position from which to extrapolate the results from the examined scrap to all future scrap generated by the existing process

[en] For decommissioning of the Fukushima Daiichi Unit 1 - 3, it is expected to predict a core status and qualify the core debris distribution (mass, location and composition) in the reactor vessel and containment). However, there has not been any useful information from the analysis results, because of the analysis model limitation such as the complex lower core region geometry in the BWR reactor vessel that any analysis codes don't incorporate it in details, in addition to the uncertainty of the boundary conditions in the accident progression. The MAAP enhancements has been planned to obtain the profitable information for the decommissioning of the plants, based on the lessons learned from the analysis results using MAAP4 conducted by TEPCO, which includes the model improvements of core melt progression, the reactor vessel failure and the core melt behavior in the containment such as core melt spreading and MCCI. The improved model will be incorporated to MAAP5 and validated by benchmarking with state-of-art knowledge. Finally, the enhanced MAAP5 will be applied to the Fukushima-Daiichi accident progression analysis. This paper summarizes the progress on this project. This project is funded by METI (Ministry of Economy, Trade and Industry) in Japan for Fukushima-Daiichi NPPs decommissioning. (authors)