Numerical Simulation of Container Breach and Airborne Release of Solids Due to Mechanical Insults

2021 ◽  
Vol 7 (3) ◽  
Author(s):  
David L. Y. Louie ◽  
San Le ◽  
Lindsay N. Gilkey
Keyword(s):  
Solid Mechanics ◽  
Source Term ◽  
Free Fall ◽  
Department Of Energy ◽  
High Fidelity ◽  
Computational Tools ◽  
Respirable Fraction ◽  
Waste Container ◽  
Model Predictions ◽  
Accident Conditions

Abstract Throughout U.S. Department of Energy (DOE) complexes, safety engineers employ the five-factor formula to calculate the source term (ST) that includes parameters of airborne release fraction (ARF), respirable fraction (RF) and damage ratio (DR). Limited experimental data on fragmentation of solids, such as ceramic pellets (i.e., PuO2), and container breach due to mechanical insults (i.e., drop and forklift impact), can be supplemented by modeling and simulation using high fidelity computational tools to estimate these parameters. This paper presents the use of Sandia National Laboratories' SIERRA solid mechanics (SM) finite element code to investigate the behavior of the widely utilized waste container (such as 7A Drum) subject to a range of free fall impact and puncture scenarios. The resulting behavior of the container is assessed, and the estimates are presented for bounding DRs from calculated breach areas for the various accident conditions considered. This paper also describes a novel multiscale constitutive model recently implemented in SIERRA/SM that simulates the fracture of brittle materials such as PuO2 and determines ARF during the fracture process. Comparisons are made between model predictions and simple bench-top experiments.

scholarly journals Nuclear Facility Safety Enhancement Using Sandia National Laboratories’ Computer Codes

2018 ◽  
Author(s):  
David L. Y. Louie
Keyword(s):  
Solid Mechanics ◽  
Source Term ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Nuclear Facility ◽  
Nuclear Facilities ◽  
Safety Research ◽  
Central Registry ◽  
Computer Codes ◽  
Safety Enhancement

This paper describes the ongoing study of nuclear facility safety enhancement using Sandia National Laboratories’ (SNL) computer codes, supported by U.S. Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program. Continued DOE NSR&D support, since 2014 has allowed the use of the SNL engineering code suite (SIERRA Mechanics) to further substantiate data in the DOE Handbook published in 1994: DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities.” The use of SIERRA codes allows for a better understanding of the mechanics, dynamics, chemistry and overall physics of airborne release scenarios. SIERRA codes provide insights into the contributing phenomena of source term releases from events such as liquid fires. The 1994 Handbook documents small-scaled, bench-top and limited experiments involving liquid fires, powder spills, pressurized releases, and mechanical insult-induced fragmentation scenarios. Data recorded from these scenarios has been substantiated using SIERRA solid mechanics and fluid mechanics codes. Data passing among multi-physics SIERRA codes predicted the contaminant release from a drum rupture due to fire even though there is no experimental data available. In the anticipated revision effort of the Handbook by DOE, these computational capabilities could enhance the data in a broader usage and could provide confidence in the safety analysis SIERRA codes can provide the initial source term to be used in the leak path factor (LPF) analyses, which predicts the ST release out of the facility. Typical LPF analysis is done using the MELCOR code, developed at SNL for the U.S. Nuclear Regulatory Commission. Widely used in nuclear reactor applications, MELCOR is a toolbox safety code in the DOE’s central registry for LPF applications. A recent LPF guidance study done by SNL indicated that MELCOR 2.1, along with updated guidance, should replace the obsolete MELCOR 1.8.5 guidance. This new guidance is significantly improved over the previous guidance, utilizing extensive MELCOR validation, including applicable reactor experiments and experiments described in the DOE-HDBK-3010-94 for LPF applications. The latest version of MELCOR should be included in DOE’s central registry, and should be used by safety analysts for LPF analyses.

scholarly journals Experimental and Mathematical Tools to Predict Droplet Size and Velocity Distribution for a Two-Fluid Nozzle

Fluids ◽  
2020 ◽  
Vol 5 (4) ◽  
pp. 231
Author(s):  
Sadegh Poozesh ◽  
Nelson K. Akafuah ◽  
Heather R. Campbell ◽  
Faezeh Bashiri ◽  
Kozo Saito
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Droplet Size Distribution ◽  
Integral Form ◽  
Accurate Estimation ◽  
Ambient Conditions ◽  
Computational Tools ◽  
Model Predictions ◽  
Model Fluids ◽  
Two Fluid

Despite progress in laser-based and computational tools, an accessible model that relies on fundamentals and offers a reasonably accurate estimation of droplet size and velocity is lacking, primarily due to entangled complex breakup mechanisms. Therefore, this study aims at using the integral form of the conservation equations to create a system of equations by solving which, the far-field secondary atomization can be analyzed through predicting droplet size and velocity distributions of the involved phases. To validate the model predictions, experiments are conducted at ambient conditions using water, methanol, and acetone as model fluids with varying formulation properties, such as density, viscosity, and surface tension. Droplet size distribution and velocity are measured with laser diffraction and a high-speed camera, respectively. Finally, an attempt is made to utilize non-scaled parameters to characterize the atomization process, useful for extrapolating the sensitivity analysis to other scales. The merit of this model lies in its simplicity for use in process control and optimization.

Friction Reduction Due to Lubrication Oil Changes in a Lean-Burn 4-Stroke Natural Gas Engine: Experimental Results

2007 ◽  
Author(s):  
Kris Quillen ◽  
Rudolph H. Stanglmaier ◽  
Victor Wong ◽  
Ed Reinbold ◽  
Rick Donahue ◽  
...  
Keyword(s):  
Natural Gas ◽  
Department Of Energy ◽  
Experimental Results ◽  
Friction Reduction ◽  
Massachusetts Institute Of Technology ◽  
Test Bed ◽  
Computational Tools ◽  
Gas Engine ◽  
Natural Gas Engine ◽  
Lubrication Oil

A project to reduce frictional losses from natural gas engines is currently being carried out by a collaborative team from Waukesha Engine Dresser, Massachusetts Institute of Technology (MIT), Colorado State University (CSU), and ExxonMobil. This project is part of the Advanced Reciprocating Engine System (ARES) program led by the US Department of Energy. Changes in lubrication oil have been identified as a way to potentially help meet the ARES goal of developing a natural gas engine with 50% brake thermal efficiency. Previous papers have discussed the computational tools used to evaluate piston-ring/cylinder friction and described the effects of changing various lubrication oil parameters on engine friction. These computational tools were used to predict the effects of changing lubrication oil of a Waukesha VGF 18-liter engine, and this paper presents the experimental results obtained on the engine test bed. Measured reductions in friction mean effective pressure (FMEP) were observed with lower viscosity lubrication oils. Test oil LEF-H (20W) resulted in a ∼ 1.9% improvement in mechanical efficiency (ηmech) and a ∼ 16.5% reduction in FMEP vs. a commercial reference 40W oil. This improvement is a significant step in reaching the ARES goals.

The Department of Energy Replacement for the 110-Gallon Specification 6M Shipping Container for Radioactive Contents

2008 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Paul T. Mann
Keyword(s):  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Department Of Transportation ◽  
Safety Requirements ◽  
Shipping Container ◽  
Regulatory Testing ◽  
Hypothetical Accident ◽  
Accident Conditions ◽  
Regulatory Commission ◽  
Energy Replacement

The Department of Energy (DOE) has been shipping university reactor fuels and other fissile materials in the 110-gallon Department of Transportation (DOT) Specification 6M container for over 20 years. The DOT 6M container has been the workhorse for many DOE programs. However, packages designed and used according to the Specification 6M (U. S. Code of Federal Regulations, 49 CFR 178.354; 2003) do not conform to the latest package safety requirements in 10 CFR 71, especially performance under hypothetical accident conditions. For that reason, the 6M specification containers are being terminated by the DOT. Packages designed to the 6M specification will no longer be allowed for in-commerce shipments after October 1, 2008. To meet on-going transportation needs, DOE evaluated several different concepts for replacing the 110-gallon 6M. After this evaluation, DOE selected the Y-12 National Security Complex for the project. The new Y-12 container, designated the ES-4100 shipping container, will have a capacity of four times the current 6M and will be certified by the Nuclear Regulatory Commission (NRC). The ES-4100 project began in September 2006 and prototypes of the new container are now being fabricated. Details on the design features and the upcoming regulatory testing of this new container are discussed in this paper.

The Importance of Criticality in the Safety Analysis of the Spent-Fuel Waste Container

1992 ◽  
Vol 294 ◽  
Author(s):  
William G. Culbreth ◽  
Paige Zielinski
Keyword(s):  
Thermal Effects ◽  
Federal Regulations ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Waste Container ◽  
Neutron Multiplication ◽  
Accident Conditions ◽  
High Level ◽  
Internal Configuration ◽  
Selection Of

ABSTRACTThe storage of high-level spent reactor fuel in a proposed national geologic repository will require the construction of containers to be placed in boreholes drilled into the host rock. Federal regulations require that the fuel be maintained subcritical under normal or accident conditions. This is determined through the calculation of a neutron multiplication factor, keff, that must remain below 0.95. Criticality will play an important role in the container design, the internal configuration of the fuel, and the selection of neutron poisons. An analysis of keff should be a normal step in the conceptualization of new waste container designs. Unlike thermal effects in a proposed repository, criticality will remain a problem long after the 10,000 year lifetime of the facility.

Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials

2005 ◽  
Author(s):  
V. N. Shah ◽  
B. Shelton ◽  
R. Fabian ◽  
S. W. Tam ◽  
Y. Y. Liu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Fissile Material ◽  
Radioactive Materials ◽  
Asme Code ◽  
Material Transportation ◽  
Accident Conditions ◽  
Regulatory Commission ◽  
High Level

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.

scholarly journals Comparison of High-Fidelity Computational Tools for Wing Design of a Distributed Electric Propulsion Aircraft

2017 ◽  
Author(s):  
Karen A. Deere ◽  
Sally Viken ◽  
Melissa Carter ◽  
Jeffrey K. Viken ◽  
Joseph M. Derlaga ◽  
...  
Keyword(s):  
Electric Propulsion ◽  
High Fidelity ◽  
Wing Design ◽  
Computational Tools

scholarly journals FEATURES OF THE SOURCE TERM CALCULATION CAPABILITY OF THE NECP-X CODE

2021 ◽  
Vol 247 ◽  
pp. 10024
Author(s):  
Xingjian Wen ◽  
Zhouyu Liu ◽  
Kai Huang ◽  
Liangzhi Cao
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
High Fidelity ◽  
Memory Requirement ◽  
Compression Method ◽  
Heat Measurement ◽  
Neutron Spectra ◽  
Decay Heat ◽  
Heat Calculation ◽  
Full Activation

The source term calculation capability is developed for the high-fidelity neutronics code NECP-X. Generally, a full activation library is used, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement during the calculation with very strict conditions, a new generalized activation chain compressed method is proposed based on the influence qualification. One basic compression element is a reaction channel or an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems, which utilizes the neutron spectra and neutron flux, is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and the influence qualification-based activation library compression method, a new compressed activation library is generated. The VERA-3A problem and the KAIST problem are used to assess the accuracy and the efficiency of the new activation library. 85 measurements of decay heat from decay heat measurement facilities GE-Morris and CLAB are used to validate the decay heat calculation in NECP-X. The results show good accuracy of NECP-X in predicting radiation source term of the spent nuclear fuel and significant memory saving when using new compressed activation library.

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