Fluid inclusions are found within mineral crystals or along grain boundaries in many sedimentary rocks, notably in evaporite formations, and can migrate along a thermal or hydro-mechanical gradient. Shale and salt rocks have been considered potential host rocks for radioactive waste disposal, due to their low permeability. Previously stagnant inclusions may become mobilised by a perturbation of the in situ state by a geotechnical installation or the emplacement of heat-generating waste. The migration of fluid inclusions can thus have important impacts on the long-term performance of a geologic repository for high-level radioactive waste disposal. As a part of the international research project DECOVALEX-2019, two aspects of fluid inclusion migration in rock salt are currently investigated under different boundary conditions: a) altered hydro-mechanical conditions as a consequence of tunnel excavation or borehole drilling and b) coupled thermo-hydro-mechanical-chemical conditions during the heating period of the post-closure phase of a repository. To obtain a mechanistic understanding of underlying physical processes for fluid inclusion migration, a multi-scale modelling strategy has been developed. Microscale hydraulic and time-dependent mechanical conditions related to the creep behaviour of rock salt are constrained by considering the macroscale stress evolution of an underground excavation. An analysis using a coupled two-phase flow and elasto-plastic model with a consideration of permeability variation indicates that a pathway dilation along the halite grain boundary may increase the permeability by two orders of magnitude. The calculated high flow velocity may explain the fast pressure build-up observed in the field. In addition, a mathematical model for the migration and morphological evolution of a single fluid inclusion under a thermal gradient has been formulated. A first-order analysis of the model leads to a simple mathematical expression that is able to explain the key observations of thermally driven inclusion migration in salt. Finally, numerical methods such as a phase field method for solving a moving boundary problem of fluid inclusion migration have also been explored.
The catastrophic nuclear reactor accident at Fukushima damaged public confidence in nuclear energy and a demand for new engineered safety features that could mitigate or prevent radiation releases to the environment in the future. We have developed a novel use of sacrificial material (SM) to prevent the molten corium from breaching containment during accidents as well as a validated, novel, high-fidelity modeling capability to design and optimize the proposed concept. Some new reactor designs employ a core catcher and a SM, such as ceramic or concrete, to slow the molten corium and avoid the breach of the containment. However, existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials (current designs are limited to water). The SM proposed in this Laboratory Development Research and Development (LDRD) project is based on granular carbonate minerals that can be used in existing light water reactor plants. This new SM will induce an endothermic reaction to quickly freeze the corium in place, with minimal hydrogen explosion and maximum radionuclide retention. Because corium spreading is a complex process strongly influenced by coupled chemical reactions (with underlying containment material and especially with the proposed SM), decay heat and phase change. No existing tool is available for modeling such a complex process. This LDRD project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. We have demonstrated small-scale to large-scaled experiments using lead oxide (Pb0) as surrogate for molten corium, which showed that the reaction of the SM with molten Pb0 results in a fast solidification of the melt and the formation of open pore structures in the solidified Pb0 because of CO 2 released from the carbonate decomposition. Our modeling simulations show that Sierra Mechanics/Aria code can be used to model a molten corium spreading experiment and the PbO/carbonate experiment. A simplified carbonate decomposition model has been developed to predict thermal decomposition of carbonate mineral in contact with corium. This model has been incorporated into an input model for MELCOR, a severe accident nuclear reactor code developed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. A full-plant MELCOR simulation suggests that the ex-vessel accident progression, e.g., core-concrete interaction and core spreading on the containment floor, could be significantly delayed by the introduction of SM to the reactor cavity prior to the reactor pressure vessel failure. Delays of one and half day are suggested with limited SM. Filling the cavity with SM might delay progression by days. Additionally, the modeling suggests that the relative concentration (molar fraction) of hydrogen in containment could be substantially reduced by the non-condensable gas (CO 2 ) generation associated with the SM reaction effectively making the hydrogen concentration below its flammable limit. ACKNOWLEDGEMENTS This research was supported by the Laboratory Directed Research and Development Program of Sandia National Laboratories (Sandia). The authors would like to express thanks to all Sandia staff who helped with this research, including Ms. Denise Bencoe for assisting with the performance of the small-scaled experiments at Advanced Material Laboratories, Ms. Amanda Sanchez and Ms. Lydia Boisvert for grinding all natural carbonate materials and sieving, Dr. Anne Grillet for measuring the microstructure of the samples using X-ray micro CT Scan (SKYSCAN 1272), Dr. Clay Payne for the XRD measurement, Dr. Eric Lindgren for assisting the selection of crucible materials, Dr. Larry Humphries for review this report and Dr. Randall O. Gauntt for reviewing this research, who has retired from Sandia at the time of this publication. The authors like to thank Ms. Laura Sowko for editing this report. Additionally, the authors appreciated the use of the FARO L-26S data information described in Section 4.2.2.1 of this report downloaded from STRESA, Joint Research Centre, European Commission (c) Euratom, 2019.