This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2019. First, the current sodium pool fire model in MELCOR, which is adapted from CONTAIN-LMR code, is discussed. The associated sodium fire input requirements are also presented. A proposed model improvement developed at Sandia is discussed. Finally, the validation study of the sodium pool fire model in MELCOR carried out by a JAEA's staff is described. To validate this model, a JAEA sodium pool fire experiment (F7-1 test) is used. A preliminary calculation is performed using a modified MELCOR model from a previous experiment simulation. The results of the calculation are discussed as well as suggestions for improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2020.
This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 11932 and 14959. Revision 11932 represents the last official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 14959. Along with the newly updated MELCOR Users' Guide [2] and Reference Manual [3], users will be aware and able to assess the new capabilities for their modeling and analysis applications. Following the official release an addendum section has been added to this report detailing modifications made to the official release which support the accompanying patch release. The addendums address user reported issues and previously known issues within the official code release which extends the original Quick look document to also support the patch release. Furthermore, the addendums section documents the recent changes to input records in the Users' Guide applicable to the patch release and corrects a few issues in the revision 14959 release as well. This page left blank.
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.
Engineers performing safety analyses throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Limited experimental data on fragmentation of solids, such as ceramic pellets (i.e., PuO2), and container breach due to mechanical insults (i.e., explosion-induced fragmentation, drop and forklift impact), can be supplemented by modeling and simulation using high fidelity computational tools. This paper presents the use of Sandia National Laboratories' SIERRA Solid Mechanics (SIERRA/SM) finite element code to investigate the behavior of two widely utilized waste containers (Standard Waste Box and 7A Drum) subject to a range of free fall impact and puncture scenarios. The resulting behavior of the containers is assessed, and a methodology is presented for calculating bounding airborne release fractions from calculated breach areas for the various accident conditions considered. The paper also describes a novel multi-scale constitutive model recently implemented in SIERRA/SM that can simulate the fracture of brittle materials such as PuO2 and determining the amount of hazardous respirable particles generated during the fracture process. Comparisons are made between model predictions and simple bench-top experiments.
Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010,Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment.Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated.
This report describes the progress on the validation of the development of MELCOR Sodium Chemistry (NAC) package. The primary focus for this report is to ensure that the implementation of the CONTAIN-LMR sodium models into MELCOR is correctly done. Thus, the verification test is to conduct the code-to-code comparison with MELCOR and CONTAIN-LMR. Last year we had reported the development of NAC package which included three sodium models: spray fire, pool fire and atmospheric chemistry. The first 2 models were completed and additional improvement for these two models were done this year to allow upward spray capability and various functional capability for modeling the pool fire experiment better, respectively. This year, the atmospheric chemistry implementation has been progressed to a point for testing in the presence of water vapor (modeled as ideal gas) as a part of the two-condensable option model in the CONTAIN- LMR. The user's guide and reference manual for the NAC package including these improvements are described in a separate document being published as a part of the MELCOR 2.2 release. For this report, we would discuss the experimental validation using the implemented spray fire and pool fire models. A code-to-code comparison with CONTAIN-LMR is described for a spray fire experiment. Note that the atmospheric chemistry model has not fully implemented due to the absence of the two condensable option. Only the chemical reactions between the sodium aerosol and water vapor can be modeled. ACKNOWLEDGEMENTS This work was overseen and managed by Matthew R. Denman (Sandia National Laboratories). In addition, we appreciate that Chris Faucett for developing experimental data and provided the initial input decks as a part of the MELCOR assessment report development for U.S. Nuclear Regulatory Commission's project. This work is supported by the Office of Nuclear Energy of the U.S. Department of Energy work package number AT-17SN170204 and NT-185N05030102.
This report describes the status of the development of MELCOR Sodium Chemistry (NAC) package. This development is based on the CONTAIN-LMR sodium physics and chemistry models to be implemented in MELCOR. In the past three years, the sodium equation of state as a working fluid from the nuclear fusion safety research and from the SIMMER code has been implemented into MELCOR. The chemistry models from the CONTAIN-LMR code, such as the spray and pool fire mode ls, have also been implemented into MELCOR. This report describes the implemented models and the issues encountered. Model descriptions and input descriptions are provided. Development testing of the spray and pool fire models is described, including the code-to-code comparison with CONTAIN-LMR. The report ends with an expected timeline for the remaining models to be implemented, such as the atmosphere chemistry, sodium-concrete interactions, and experimental validation tests .
Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE-HDBK-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms from postulated accident scenarios. In calculating source terms, analysts tend to use the DOE Handbook’s bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes.