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A packaged toolbox with stabilization and functionalization strategies for robust and multifunctional metal–organic framework (MOF) materials has been summarized and discussed.
In addition to the radiological characterization and control measures the materials composition has to be described and respective control measures need to be implemented. The approach to verify the materials composition depends on the status of the waste: - During conditioning of raw waste the control of the materials composition has to be taken into account. - For already conditioned waste a retrospective qualification of the process might be possible. - If retrospective process qualification is not possible, legacy waste can be qualified by spot checking according to the materials composition requirements The integration of the control of the material composition in the quality control system for radioactive waste is discussed and examples of control measures are given. With the materials-list and the packaging-list the Federal Office for Radiation Protection (BfS) provides an appropriate tool to describe the materials composition of radioactive waste packages. The control measures with respect to the materials composition integrate well in the established quality control framework for radioactive waste. The system is flexible enough to deal with waste products of different qualities: raw waste, qualified conditioned waste or legacy waste. Control measures to verify the materials composition can be accomplished with minimal radiation exposure and without undue burden on the waste producers and conditioners. (authors)
A set of obligations in the yearly storage container surveillance program, ensures that a retrospective approach is applied with loaded containers in inventory to confirm compliance with all “users” to ensure nuclear material storage container requirements are being followed. Validation that packaged containers are properly used, within a specific set of container types prescribed “bounding conditions of use,” is based on meeting requirements as identified in operating procedures associated with TA55-DOP-091, TA-55 Nuclear Material Packaging, and PA-RD-01022, Nuclear Material Packaging Requirements. The surveillance plan obligation requires the application of local area nuclear material accountability software or (LANMAS) to produce queries. The data queries are used to assess attributes of containers in storage these investigations are conducted on a bi-annual basis. This rules of use (ROU) compliance process ensures proper usage of storage containers as containment systems and therefore provides effective worker protection. It is structured as a supporting effort by implementing aspects of PA-AP-01207, Nuclear Material Container Safety Management at TA-55.
Highlights: Black-Right-Pointing-Pointer Study of the influence of materials in the pyrolysis of real plastic waste samples. Black-Right-Pointing-Pointer Inorganic compounds remain unaltered. Black-Right-Pointing-Pointer Cellulosic components give rise to an increase in char formation. Black-Right-Pointing-Pointer Cellulosic components promote the production of aqueous phase. Black-Right-Pointing-Pointer Cellulosic components increase CO and CO{sub 2} contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm{sup 3} reactor, swept with 1 L min{sup -1} N{sub 2}, at 500 Degree-Sign C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg{sup -1}). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO{sub 2}; their HHV is in the range of 18-46 MJ kg{sup -1}. The amount of CO-CO{sub 2} increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.
A set of obligations in the yearly storage container surveillance program, ensures that a retrospective approach is applied with loaded containers in inventory to confirm compliance with all “users” to ensure nuclear material storage container requirements are being followed. Validation that packaged containers are properly used, within a specific set of container types prescribed “bounding conditions of use,” is based on meeting requirements as identified in operating procedures associated with TA55-DOP-091, TA-55 Nuclear Material Packaging, and PA-RD-01022, Nuclear Material Packaging Requirements. The surveillance plan obligation requires the application of local area nuclear material accountability software or (LANMAS) to produce queries. The data queries are used to assess attributes of containers in storage these investigations are conducted on a bi-annual basis. This rules of use (ROU) compliance process ensures proper usage of storage containers as containment systems and therefore provides effective worker protection. It is structured as a supporting effort by implementing aspects of PA-AP-01207, Nuclear Material Container Safety Management at TA-55.
This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.
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The lifetime of a power electronics package is, to a large extent, determined by the reliability of its bonded interfaces under major loading conditions in an operational environment. Based on the application-level requirements of the package, bonded interface material is selected based on the results obtained from accelerated tests such as thermal cycling and power cycling. In addition to evaluating the reliability through accelerated tests, it is important to consider the impact of other component layers on the thermomechanical performance of the interface material, both from a material and geometric perspective. The co-efficient of thermal expansion (CTE) mismatch introduced by the use of different materials within a package and its structural design plays a critical role in determining the interface material reliability. In this paper, we present the results of a parametric modeling study of a power electronics package under thermal cycling in which the materials and geometric design of the different component layers were varied with respect to a baseline design to understand their impact on the reliability of the interface material. We chose the volume-averaged strain energy density per cycle computed at the corner region of the interface material as the metric for the reliability comparisons. Our results indicate that in addition to the CTE mismatch, the stiffness of individual component layers has a major impact on reliability. Among the different baseplates that we studied, aluminum silicon-carbide baseplates offered superior reliability over their copper and aluminum counterparts. We also found that the magnitude of the impact of stiffness variation—amongst the adjacent layers—on the reliability of the solder joint is dependent on the inherent CTE mismatch between the layers.
The lifetime of a power electronics package is, to a large extent, determined by the reliability of its bonded interfaces under major loading conditions in an operational environment. Based on the application-level requirements of the package, bonded interface material is selected based on the results obtained from accelerated tests such as thermal cycling and power cycling. In addition to evaluating the reliability through accelerated tests, it is important to consider the impact of other component layers on the thermomechanical performance of the interface material, both from a material and geometric perspective. The co-efficient of thermal expansion (CTE) mismatch introduced by the use of different materials within a package and its structural design plays a critical role in determining the interface material reliability. In this paper, we present the results of a parametric modeling study of a power electronics package under thermal cycling in which the materials and geometric design of the different component layers were varied with respect to a baseline design to understand their impact on the reliability of the interface material. We chose the volume-averaged strain energy density per cycle computed at the corner region of the interface material as the metric for the reliability comparisons. Our results indicate that in addition to the CTE mismatch, the stiffness of individual component layers has a major impact on reliability. Among the different baseplates that we studied, aluminum silicon-carbide baseplates offered superior reliability over their copper and aluminum counterparts. We also found that the magnitude of the impact of stiffness variation—amongst the adjacent layers—on the reliability of the solder joint is dependent on the inherent CTE mismatch between the layers.
In a radioactive material (RAM) packaging, the formation of eutectic at the Pu/SS (plutonium/stainless steel) interface is a serious concern and must be avoided to prevent of leakage of fissile material to the environment. The eutectic temperature for the Pu/SS is rather low (410 C) and could seriously impact the structural integrity of the containment vessel under accident conditions involving fire. The 9975 packaging is used for long term storage of Pu bearing materials in the DOE complex where the Pu comes in contact with the stainless steel containment vessel. Due to the serious consequences of the containment breach at the eutectic site, the Pu/SS interface temperature is kept well below the eutectic formation temperature of 410 C. This paper discusses the thermal models and the results for the extended fire conditions (1500 F for 86 minutes) that exist in a long term storage facility and concludes that the 9975 packaging Pu/SS interface temperature is well below the eutectic temperature.
High velocity munitions and kinetic penetrators experience monumental external forces, impulses, and accelerations. The hard target environment is immensely taxing on sophisticated electronic components and recorders designed to retrieve valuable data related to the systems performance and characteristics in the periods of flight, impact, and post-impact. Such electronic systems have upper limits of overall shock intensity which, if exceeded, will either shorten the operating life of the parts or risk destruction resulting in loss of both the data and the principal value of the recorder. The focus of this project was to refine the categorization of leading material types for encapsulation and passive shock isolation and implement them in a method useable for a wide variety of environments. Namely, a design methodology capable of being tailored to the specific impact conditions to maximize the lively hood of sensitive electronics and the information recorded. The results of the study concluded that the materials observed under consistent dynamic high strain rate tests, which include Conathane® EN-4/9, Slygard®-184, and Stycast™-2651, behaved well in certain aspects of energy transmission and shock when considering the frequency environment or package coupled with the isolation material’s application. Key points about the implementation of the materials in extreme shock environments is discussed with the connection to energy analysis, loss attributes, and pulse transmissibility modeling. However, attempts to model the materials solely based on energy transmissibility in the frequency domain using only external experimental data and simplified boundary conditions was not found to be consistent with that acquired from the pressure bar experiments. Further work will include the addition of further material experimentation of the encapsulants in other frequency and temperature states, confined and pre-load boundary states, and composite constructions.
The 9975 Type B shipping package is used within the DOE complex for shipping special nuclear materials. This package is re-certified annually in accordance with Safety Analysis Report for Packaging (SARP) requirements. The package is also used at the Savannah River Site as part of the long-term storage configuration of special nuclear materials. As such, the packages do not undergo annual recertification during storage, with uncertainty as to how long some of the package components will meet their functional requirements in the storage environment. The packages are currently approved for up to 15 years storage, and work continues to provide a technical basis to extend that period. This report describes efforts by the Savannah River National Laboratory (SRNL) to extend the service life estimate of Viton® GLT and GLT-S fluoroelastomer O-rings used in the 9975 shipping package. O-rings of both GLT and GLT-S compositions are undergoing accelerated aging at elevated temperature, and are periodically tested for compression stress relaxation (CSR) behavior. The CSR behavior of O-rings was evaluated at temperatures from 175 to 400 °F. These collective data were used to develop predictive models for extrapolation of CSR behavior to relevant service temperatures (< 156 °F). The predictive model developed from the CSR data conservatively indicates a service life of approximately 37 years for Viton GLT O-rings at the maximum effective service temperature of 156 °F. The estimated service life for Viton GLT-S O-rings is significantly longer.
The Savannah River Site (SRS) is storing plutonium (Pu) materials in the K-Area Materials Storage (KAMS) facility. The Pu materials were packaged according to the DOE-STD-3013 standard and shipped to the SRS in Type B 9975 packages. The robust 9975 shipping package was not designed for long-term product storage, but it is a specified part of the storage configuration and the KAMS facility safety basis credits the 9975 design with containment. Within the 9975 package, nested stainless steel containment vessels are closed with dual O-ring seals based on Viton{sup R} GLT or GLT-S fluoro-elastomer. The aging behavior of the O-ring compounds is being studied to provide the facility with advanced notice of nonconformance and to develop life prediction models. A combination of field surveillance, leak testing of surrogate fixtures aged at bounding service temperatures, and accelerated-aging methodologies based on compression stress-relaxation and oxygen consumption analysis is being used to evaluate seal performance. A summary of the surveillance program relative to seal aging behavior is presented. The aging behavior of fluoro-elastomer seals based on Viton{sup R} GLT and GLT-S is being studied to develop life prediction models in support of long-term storage of plutonium materials in the 9975 shipping packages at the Savannah River Site. Field surveillance data in combination with accelerated-aging data suggest a significant lifetime for the seals. Typical storage conditions are not anticipated to challenge the leak-tightness of the seals for many years. Early life prediction models based on compression stress relaxation indicate a seal lifetime of ∼12 years at the maximum service temperature predicted (93 deg. C). Seal lifetimes at lower, more realistic conditions are likely significantly longer. Service life predictions based on CSR data are thus far conservative relative to predictions based on time to leakage failure. Surveillance data on packages examined after 6 years in storage show only minor compression set of the O-rings and no significant degradation. Surveillance and testing will continue as needed to validate and refine life prediction models. (authors)
The 9975 Type B shipping package is used within the DOE complex for shipping special nuclear materials. This package is re-certified annually in accordance with Safety Analysis Report requirements. The package is also used at the Savannah River Site as part of the long-term storage configuration of special nuclear materials. As such, the packages do not undergo annual recertification during storage, with uncertainty as to how long some of the package components will meet their functional requirements in the storage environment. The packages are currently approved for up to 15 years storage, and work continues to provide a technical basis to extend that period. This paper describes efforts by the Savannah River National Laboratory (SRNL) to extend the service life estimate of Viton® GLT and GLT-S fluoroelastomer O-rings used in the 9975 shipping package. O-rings of both compositions are undergoing accelerated aging at elevated temperature, and are periodically tested for compression stress relaxation (CSR) behavior and leak performance. The CSR behavior of O-rings was evaluated at temperatures from 79 °C to 177 °C. These collective data were used to develop predictive models for extrapolation of CSR behavior to relevant service temperatures (< 75 °C). O-rings were also aged in Primary Containment Vessel (PCV) fixtures at temperatures ranging from 79 °C to 232 °C. The fixtures are helium leak tested periodically to determine if they remain leak-tight. The PCV fixture tests demonstrate that the 9975 O-rings will remain leak-tight at temperatures up to 149 °C for 3 years or more, and no leak failures have been observed with up to 8 years aging at 93 °C. Significantly longer periods of leak-tight service are expected at the lower temperatures actually experienced in the storage environment. The predictive model developed from the CSR data conservatively indicates a service life of more than 20 years at the bounding temperature of 75 °C. Although the relationship between CSR behavior and leak-tight performance has not been established for this design, the CSR predictions for this O-ring are conservative relative to leak-tight performance to date.
The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation has sponsored a suite of training courses that have been conducted since the early 1990's. More recently, PCP has developed an online tutorial - RAM Packaging Safety Tutorial, as well as several courses developed and taught by several U.S national laboratories. The courses developed and taught by the national laboratories are referred to as the DOE Packaging University. Since 2013 as part of the Packaging University, the DOE PCP has worked with the University of Nevada, Reno (UNR) Mechanical Engineering Department to develop an accredited graduate-level certificate in Nuclear Packaging. This paper provides an overview of the currently offered PCP courses, as well as the UNR graduate certificate in Nuclear Packaging. (authors)
The contents approved for shipment in a Type B radioactive material transportation package have historically been descriptions of discrete items, or groupings of well-defined similar items, in the package safety basis documentation. The need for a comprehensive functional content envelope of both gamma and neutron emitting nuclides compliant with regulatory limits has become necessary as the DOE complex requires shipments of unique mixtures of radionuclides and impurities. Recent publications have presented a calculational model and a corresponding content envelope intended to be compliant with federal regulatory external radiation limits as well as design decay heat limits based on the Model 9977 Packaging. The methodology used to develop this content envelope consisted of determining the external radiation dose rates based on one gram of a given isotope combined, in the case of actinides, with various levels of light element impurities and determining the allowable mass to the regulatory limits based on these dose rates. The method of ratioing from these calculations to the masses that meet the regulatory limits fails in the case of some combinations of actinides and light element impurities because of the effect of subcritical multiplication, resulting in non-conservative and non-compliant dose rates in some cases. This is particularly true for many actinides combined with beryllium, boron, fluorine, lithium, and sodium. In this study, the source was modeled as a sphere with the appropriate dimensions based on the actinide or gamma source density and placed at the bottom of the containment vessel for maximum conservatism. The results presented in this report rectify the above-noted deficiencies by adjusting the previously determined mass limits so that the resulting dose rates are compliant with regulatory limits. Some amount of iteration was required in some instances to adjust the masses for dose compliance. In some instances the masses were increased since they were too conservative. In addition, the method of ratioing the masses when applied to several gamma emitting isotopes also resulted in either non-conservative or overly conservative data. The mass imits presented here are for bare sources and are complaint with the design decay heat limit of 19W. The masses are based on package surface dose rates (the limiting dose rate) that fall in the range 185-195 mrem/h, thus giving an additional margin of between 7.5% and 2.5% to the regulatory limit of 200 mrem/h. This report presents a revised set of mass limits (i.e., content envelope) for several neutron emitting actinides with varying levels of light element impurities compliant with both external radiation and design decay heat limits. In addition, revised limits for gamma sources are also presented. The neutron emitter mass limits are in Tables 1 through 9 and the gamma emitter mass limits are in Table 10. The revised content envelope covering a wide range of materials present in the DOE complex, was developed for the Model 9977 Packaging, providing conservative limits for other commonly-used Type B Packagings such as the Models 9975 and 9978. It should be noted that for the Model 9977 package, shipment is also possible with three shielded containers – lead or tungsten for gamma sources and polyethylene for neutron emitters. The design decay heat limits for these are as follows: 19W for tungsten, 6W for lead and 3W for polyethylene. The limits presented in this report would need to be checked against the lower decay heat limits for the lead and polyethylene and adjusted as required for compliance. Criticality safety is not factored into these limits and should be evaluated on a case by case basis.
The purpose of this work was to discover a method to investigate the properties of interfaces as described by a numerical physical model. The model used was adopted from literature and applied to a commercially available multiphysics software package. By doing this the internal properties of simple structures could be elucidated and then readily applied to more complex structures such as valves and pumps in laminate microfluidic structures. A numerical finite element multi-scale model of a cohesive interface comprised of heterogeneous material properties was used to elucidate irreversible damage from applied strain energy. An unknown internal state variable was applied to characterize the damage process. Using a constrained blister test, this unknown internal state variable could be determined for an adherend/adhesive/adherend body. This is particularly interesting for laminate systems with microfluidic and microstructures contained within the body. A laminate structure was designed and fabricated that could accommodate a variety of binary systems joined using nearly any technique such as adhesive, welding (solvent, laser, ultrasonic, RF, etc.), or thermal. The adhesive method was the most successful and easy to implement but also one of the more difficult to understand, especially over long periods of time. Welding methods are meant to achieve a bond that is similar to bulk properties and so are easier to predict. However, methods of welding often produce defects in the bonds.. Examples of the test structures used to elucidate the internal properties of the model were shown and demonstrated. The real life examples used this research to improve upon current designs and aided in creating complex structures for sensor and other applications.
Radioactive materials are widespread in our everyday activities which are used in many different types of industries: Consumer products, Medicine, Industry, Energy, and Defense. In the United States there is about 3 million shipments of radioactive materials per year. With radioactive material being very widespread the number of new applications to ship these quantities increases daily. As such, ensuring the safe and secure transport of these materials is paramount, necessitating stringent packaging, transportation, and certification requirements. Shipments of radioactive materials have been made from Oak Ridge National Laboratory (ORNL) for the past 65 years. Since the 1940s ORNL has been at the forefront of the early development of regulations pertaining to the transport of radioactive shipments and associated regulatory testing. The early involvement resulted in ORNL being one of the leaders in the development of regulations and testing standards pertaining to the testing and design of packages. Today all package testing activities are conducted under the Package Testing Program (PTP) at ORNL at the Package Evaluation Facility (PEF). The PTP develops and evaluates testing solutions – ensuring that they are safe, efficient, and meet regulatory requirements ORNL PTP has extensive experience in the regulatory testing of radioactive material packages, supporting certification of Type A, Type, B, Type AF, and special form packages. The PEF serves as a one-stop-shop for design, testing, and certification of a wide variety of packages needed for shipment of radioactive materials. The facility provides the space, expertise, and equipment necessary to conduct rigorous performance testing under both normal and accident conditions or transport environments including free drop, dynamic crush, puncture, penetration, compression, vibration, water spray, water immersion, and thermal test. The PTP remains a key contributor to the safety and compliance of radioactive material transportation. This presentation will provide an overview of the program’s history, capabilities, and its ongoing role in advancing regulatory standards and transportation safety.