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DOE OSTI2026

Effect of Radical Initiators on Polypropylene Thermal Deconstruction

A new route for polypropylene (PP) deconstruction through radical pathways based on thermodynamic and kinetic considerations has been investigated. Radical polypropylene (PP) deconstruction, activated through small quantities of initiators, can enable the deconstruction of waste POs into unsaturated products. We found that stirring was detrimental to the β-scission extent because it accelerates radical termination reactions through mixing. The best results were achieved by using a semibatch process that included mechanical mixing during the temperature ramp, static heating of the polymer/initiator mixture at the final temperature, and volatile product removal with N 2 . Dicumyl peroxide and alkylated dicumene initiators produced similar liquid and solid products after a 30 min thermal treatment. Terminal alkenes were formed in the liquid products of reactions at 375 and 400 °C, with about 5% of the protons in the liquid product ascribed to terminal alkenes. Yields of liquid products increased with temperature, reaching 60% w/w at 400 °C; at the same time, yields of solid products decreased with temperature to 16% w/w at 400 °C. Although radical initiators decrease PP molecular weight during the temperature ramp, at 400 °C, initiators only marginally increased the liquid product fraction compared to control experiments. Finally, the terminal double bonds in the liquid product mixture (C8–C36) provide multiple pathways for upgrading to surfactants, plasticizers, lube oils, and other valuable products.

liquids↗
DOE OSTI2024

Chapter 7.1: Microalgae

Microalgae is a unique biomass resource that does not need to compete for land and water with other biomass feedstocks because it can be cultivated on low-quality unencumbered land using noncompetitive water types including saline and wastewater. It is included as a complementary resource alongside other biomass feedstocks reported in this study, albeit at higher biomass production costs reflective of more capital-intensive farming operations than typical for terrestrial biomass. Higher biomass costs can be offset by the potential to produce value-added coproducts unique to compositional constituents of microalgae. Relative to BT16, this chapter reflects the latest analysis from the 2022 Algae Harmonization Update, which uses the latest parameterized and high-performing saline algal strain, second-generation carbon capture of point-source waste CO2 and high-pressure pipeline transport resolved to specific point-source types, saline water sourcing up to 40,000 mg/L total dissolved solids for source and makeup water salinity, blowdown water treatment and recycle, and brine disposal handling. National-scale algal biomass availability potential was calculated at 152 million tons/yr ash-free dry weight (AFDW) (191 million tons/yr dry weight) at an average biomass productivity of 26.2 g/m2 -day AFDW (about 50 tons/acre/yr dry weight).1 The algal biomass was cultivated on 3.9 million acres of multi-criteria screened and potentially available land for CONUS and fixed 268 million tons of waste CO2 based on biomass uptake. The algae biomass can be produced at an average MBSP of $674/ton AFDW ($536/ton dry weight) in 2020 dollars,2 corresponding to a total energy potential of 3.3 quads/yr at an average MBSP of $31.2/MMBtu (higher-heating-value [HHV] basis).

algae↗
DOE OSTI2021

Characterization of CST Post-Processing AP-105 Hanford Tank Waste

The primary goal of the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions, LLC (WRPS), is to remove entrained solids and 137 Cs from the Hanford tank waste supernate to expedite production of low-activity waste. Ion exchange (IX) testing of 10.9 L of waste from Hanford tank 241-AP-105 (AP-105), performed by Pacific Northwest National Laboratory, used a lead-lag-polish column format, with a bed volume of 10 mL per column, to decontaminate tank waste supernate using crystalline silicotitanate (CST) as the IX media. The AP-105 Cs IX processing test, discussed elsewhere, resulted in a shorter transition zone (i.e., steeper load curve) than those defined by wastes from tanks 241-AP-107 and 241-AW-102.1 The shorter transition zone was indicative of a matrix effect retarding Cs capacity. Therefore, aliquots of spent CST from the lead, lag, and polish columns were subjected to a digestion protocol to quantify analytes retained by the CST and extrapolate the impact on Cs capacity. The spent CST was digested using a combination of 5 M HNO 3 and H 2 O 2 with vigorous heating and stirring. Due to the radiation dose accompanying the 137 Cs on the CST columns, a secondary Cs separation by ammonium molybdophosphate embedded in polyacrylonitrile (AMP-PAN) was performed to separate the 137 Cs from the CST so the samples could be contact-handled for analysis outside of a shielded facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Simple but tricky: Investigations of terephthalic acid purity obtained from mixed PET waste

In this study, we report for the first time, the basic depolymerization of mixed waste-polyethylene terephthalate (PET) by hydrolysis and subsequent terephthalic acid monomer recovery at high purity using benign reaction conditions. Several conditions were tested for depolymerization such as PET chips size, concentration of aqueous sodium hydroxide (20 or 30%), organic co-solvent (ethylene glycol or ethanol), temperature at which the reaction was run and duration of the heating. More importantly, several batches of PET were utilized as starting materials including a commercial PET, chopped PET obtained from clean bottles whose caps and labels were removed and the purity of the product from each condition was evaluated via nuclear magnetic resonance (1HNMR and 13CNMR), differential scanning calorimetry (DSC) and powder Xray diffraction (XRD). Unsurprisingly, the conversion of PET is dependent on the particle size varying from 100% conversion for fine powder to 73% conversion for (300 µm, mesh 6 – 20 or mesh 14-20). Ethanol appears to be more efficient as a co-solvent than ethylene glycol, with higher PET depolymerization conversions (94% versus 75-80%), shorter reaction times (2h versus 6h) and lower temperatures (80 °C versus 110 °C). The terephthalic acid (TPA) recovered appeared to have only subtle differences among the batches, most notably a pink color when the reaction was run in ethanol/base. The DSC of the compounds produced in ethylene glycol water appear to display a melting point (280-288 °C) while the samples prepared in ethanol as well as a commercial sample did not. Overall, the purity of the various TPA batches is comparable, and similar to commercial TPA, demonstrating the utility of the method to depolymerize realistic waste streams. The method is simple, demonstrated on multigram scale (15-30g) and allows for the complete removal of waste other than PET unaffected by alkaline conditions.

Cosimbescu, Lelia↗
DOE OSTI1988

Nuclear Safety [Vol. 29, No. 1, January-March 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. THE CHERNOBYL ACCIDENT: 1 Core History and Nuclide Inventory of the Chernobyl Core at the Time of Accident by Gerald Kirchner and Cornelius C. Noack; GENERAL SAFETY CONSIDERATIONS: 6 Nonprescriptive Nuclear Safety Regulation: The Example of Loss of Offsite Power by M. W. Golay, V. P. Manno, and C. Vlahoplus, Jr., 20 Nuclear Power Safety Goals in Light of the Chernobyl Accident by C. Whipple and C. Starr; ACCIDENT ANALYSIS: 29 Containment Loads from Severe Accidents—U.S. Program by W. Kerr and M. K. Dey; PLANT SAFETY FEATURES: 36 Safety Characteristics of Modern High-Temperature Reactors: Focus on German Designs by W. Kröger, H. Nickel, and R. Schulten; ENVIRONMENTAL EFFECTS: 49 Radiation Hormesis and Nuclear Safety by C. C. Congdon; OPERATING EXPERIENCES: 58 Evaluation of Nonradiological Water Chemistry at Power Reactors by Harvey Zibulsky, James J. Kottan, Walter J. Pasciak, Mary Ann Castrogivanni, and Sujit Banerjee, 64 Reactor Shutdown Experience Compiled by J. W. Cletcher, 67 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 85 General Administrative Activities Compiled by E. G. Silver, 94 Reports, Standards, and Safety Guides by D. S. Queener, 100 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 103 Proposed Rule Changes as of Sept. 30, 1987; ANNOUNCEMENTS: 5 Northwestern University Short Course on Radiation Safety, 28 International Approach to Nuclear Safety (After Three-Mile Island and Chernobyl), 57 RPI Short Course on Modern Developments in Boiling Heat Transfer and Two-Phase Flow, 107 20th DOE/NRC Nuclear Air Cleaning Conference, 110 Third International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, 108 The Authors, 111 Indexes to Nuclear Safety, Volume 28.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1992

Nuclear Safety [Vol. 33, No. 4, October-December 1992]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 499 Technical Note: Nuclear Regulatory Commission’s Examination Question Bank System, C. L Jensen and N. L. Skinner; 506 Report on the Specialists’ Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5-7,1991, S. R. Paranjpe; ACCIDENT ANALYSIS: 514 A Review of Hydrogen Production During Melt/Water Interaction in LWRs, D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; 535 A Review of the U 3 0 8 -Al Reaction as a Potential Heat Source in Research and Test Reactor Accidents, J. L. Snelgrove; CONTROL AND INSTRUMENTATION: 543 Multilevel Interfaces for Power Plant Control Rooms II: A Preliminary Design Space, K. J. Vicente; DESIGN FEATURES: 549 Twenty-Second DOE/NRC Nuclear Air Cleaning and Treatment Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 562 Book Review: Safety Assessment of Unit 5 (WWER-440/W-213) of the Greifswald Nuclear Power Station, M. H. Fontana; WASTE AND SPENT FUEL MANAGEMENT: 576 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 586 Analysis of Corrective Actions Applied to Nuclear Power Plant Operations, A. E. Cross; 593 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 596 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; 609 Non-Commercial Power Operating Experience; 614 World Nuclear Performance, Compiled by M. D. Muhlheim; RECENT DEVELOPMENTS: 623 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 635 Reports, Standards, and Safety Guides, D. S. Queener; 639 Proposed Rule Changes as of June 30, 1992; ANNOUNCEMENTS: 513 Fourth International Conference on Nuclear Fuel Reprocessing and Waste Management RECOD '94; 585 International Topical Meeting on Advanced Reactors Safety; 622 International Workshop on Implementation of ALARA at Nuclear Power Plants; 648 Short Course on Nuclear Criticality Safety; 648 8th Topical Meeting on Problems in Reactor Physics MEPhi-VOLGA-93; 643 The Authors; 646 Reviewers of Nuclear Safety, Vol. 33

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Coupled chemical–CFD modeling of unsealed dry storage of advanced test reactor spent fuel

We report given the current lack of long-term disposal for spent nuclear fuel storage in the United States, the current inventory of advanced test reactor (ATR) fuel at Idaho National Laboratory (INL) is expected to be stored in a dry storage facility for an extended storage period. As part of an effort to understand the conditions during a 50-year period, a three-dimensional computational fluid dynamics (CFD) of the unsealed canisters packed with spent fuel in the storage facility was constructed, this CFD model is coupled with radiolytic chemical reactions associated with water vapor and other species. The coupled model is simulated over a range of sensitivity parameters for the fuel decay heat, relative humidity, and oxyhydroxide film thickness for undried and fully-dried fuel. The most important parameter was the initial fuel decay heat. Over the range of cases, the hydrogen and nitric acid concentrations are low enough no significant impact is expected.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Additively Manufactured Solid-State Luminaire

This project addresses several key barriers to wide-spread adoption of additive manufacturing (AM) technology as applied to solid state lighting luminaires. The solution will utilize cutting edge AM approaches for integrating structure with thermal management solutions, electronic functionality, and optics. The research team (Eaton, Lighting Research Center (LRC) at Rensselaer Polytechnic Institute, Xerox Research Centre of Canada (XRCC)) utilize their AM and lighting expertise to investigate breakthrough manufacturing approaches that will significantly reduce cost, eliminate manufacturing process waste, and improve luminaire efficacy. The team has identified critical areas of research and proposed novel technical approaches to achieve these goals. Key areas of focus in Budget Period 1 (BP1) of the project quantified the impact of applying AM methodologies to the main, discrete subsystem components (Heat Sink, Housing, Optics, Electronics). Budget Period 2 (BP2) research explored similar impact on a fully integrated, AM modular luminaire concept. Final Achievement of the Target Metrics for the project are as follows: Material Reduction: achieved > 57.45% (target is 50%) Manufacturing Process: achieved > 51% reduction (target is 50%) Application Efficacy: achieved 126 lm/W (target is 130 lm/W) First Cost vs Baseline: demonstrated 49% improvement in project timing, 59% improvement in man hour savings and 89% worse BOM costs (due to deficiencies in current “state of the art” equipment). The BOM costs improve to 53% savings if state of the art processes and equipment could have been used.

3D Printing↗
DOE OSTI2022

Thermochemical Conversion of Sugarcane Bagasse: A Comprehensive Analysis of Ignition and Burnout Temperatures

The Brazilian sugarcane industry generates a significant amount of waste each year, which should be properly analyzed and studied to allow an adequate recovery and application supported by the best understanding of its properties. The present work reports the ignition and burnout temperatures of sugarcane bagasse (SCB) obtained after performing a thermal analysis using four different heating rates. The intersection method (IM) and deviation method (DM) were employed to approach the ignition and burnout temperatures of the sugarcane bagasse. The ignition temperatures of the SCB measured from IM are between 250 and 263 °C, and their burnout temperatures are between 357 and 377 °C. The ignition temperature was in the range of 205 °C to 236 °C for the DM. IM is recommended for determining the ignition and burnout temperatures. In TGA, the heating rates in the range of 10 °C·min−1 and 15 °C·min−1 are suggested due to their accuracy and the contribution to timesaving in the analysis.

Morais, Leandro C. de (ORCID:0000000314685453)↗
DOE OSTI2020

Design, Development and Testing Plan for Energy Efficiency Algorithms Related to Building-Integrated Cooling, Heating, and Power Systems

Building-integrated cooling, heating and power (CHP) systems are more efficient than conventional systems at providing local power and thermal energy, and favorable fuel prices are bound to spur their increased adoption. It is estimated that the total CHP technical potential in the commercial building sector is approximately 76 GW, but only a small fraction of it is actually deployed. Advanced control and monitoring systems that realize the full potential of these CHP systems can transform this capacity into an asset for both the owner and the grid and also accelerate the adoption of these systems in the building sector. However, to realize the full benefit of the CHP systems, we must ensure persistence of energy efficient operations. It has been demonstrated by several studies that as much as 30% of building energy use is wasted. Much of the waste results from the inability to anticipate load variations (design day vs. non-design day operations; diurnal variation of loads), identify improper operation, and efficiently operate, control and maintain building systems. Moreover, many buildings, even those with adequate control infrastructure, lack knowledgeable staff to operate and maintain energy systems. Operators of buildings with CHP systems will constantly need to make tradeoffs (economic dispatch) between onsite power generation and purchasing power, between onsite heat production from distributed generation versus heating/cooling from conventional systems, and different passive and active load management measures. Much of the inefficiency in the current building operations can be eliminated by use of automated performance monitoring (PM), real-time commissioning verification (CxV) and automated fault detection and diagnostic (AFDD) tools. Automation can help system operators make intelligent decisions. Remote and continuous monitoring of system conditions and performance will enable better management and integration of CHP with existing building systems. Continuous PM, real-time CxV and AFDD could alleviate burdens for operations staff, enhance operations and maintenance (O&M), and improve reliability of building and CHP systems. To address the O&M challenges and to provide a means to maximize the rate-of-return of building-integrated CHP systems, the Building Technologies Office (BTO) within the U.S. Department of Energy’s (DOE’s) Office of Energy Efficiency and Renewable Energy (EERE) initiated a project to design, develop, and field test a VOLTTRON™-based supervisory controller and associated open-source algorithms. These algorithms will ensure real-time optimal operation of a building-integrated CHP system, support electric grid reliability, and lead to achieving the goal of a clean, efficient, reliable and affordable next-generation integrated energy systems. This report lists the components for which PM, real-time CxV and AFDD algorithms will be developed, how the algorithms will be tested, and metrics that will be used to validate the algorithms and easy of deployment. Deployment of these algorithms in the field will result in savings reduction in energy consumption of between 10% and 20%.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2020

Novel Membrane and Electrodeposition-Based Separation and Recovery of Rare Earth Elements from Coal Combustion Residues (Final Report)

This project developed a hydrometallurgical-based technology to extract and concentrate rare earth elements (REEs) from a representative group of coal combustion residuals (CCRs) from major coal sources in the United States. The approach for REE recovery comprises two general components: (1) Leaching of REEs from CCRs and (2) Separation of REEs from other major ions in the CCR leachate. The leaching step entails acid leaching with nitric or hydrochloric acid, and in some cases, pretreatment of CCRs with alkaline roasting prior to acid leaching. The follow-up REE recovery steps involve a series of membrane-based methods to separate and concentrate rare earth ions from other major ions in acid leachates of CCRs. These methods include liquid emulsion membranes, supported liquid membranes and electrodeposition with carbon nanotube electrochemical filters. The advantage of these separation approaches (relative to conventional approaches such as solvent extraction) is a reduction in the usage of solvents and other costly chemicals and a reduction of hazardous waste products that are challenges in conventional REE separation methods. In the testing of our process, we selected a representative set of CCR samples from our collection and performed a suite of characterization measurements (total REE content and major mineralogy). The major results of this work demonstrated that heated acid leaching could efficiently extract REEs from PRB-based coal fly ashes while the Appalachian and Illinois Basin CCRs required alkaline roasting with NaOH prior to acid leaching at room temperature. The supported liquid membrane (SLM) configuration and electrochemical deposition (ED) system yielded the most promising methods of REE recovery from the acid leachates of fly ash. Our most concentrated product to date comprised 1 wt.% (dry basis) REEs produced from an Appalachian-based fly ash leached and purified by SLM. The SLM separation process was more selective for heavy REEs over the light REEs. When SLM was combined with ED, preliminary results demonstrated that a final product of >2 wt.% REEs was possible.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2021

Modeling of ATR fuel in DOE Standard Canisters with Helium Backfilled Condition

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel dictates storage within helium backfilled sealed DOE standard canisters. These sealed canisters are intended for extened (>50 year) dry storage). The typical packaging configuration for the 15-foot DOE canisters places 10 ATR elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water from the surface. As testing to the effectiveness of the drying procedure is still underway, this modeling will include results at fully saturated and fully dried conditions. In previous modeling efforts, the G-value for the production of hydrogen from the oxyhydroxide layers was assumed to be in argon environments as measured by Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution. Previous experimental testing showed differences in the hydrogen generated based on the gaseous environment. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values are 28% and 58% higher than values for argon. In addition, a change to the modeling of the oxyhydroxide radiolysis has been made from previous reports. This change assumes the dependency of the dose rate on the overall reaction rate is applied to the total weight of the sample, rather than to just the weight of the oxyhydroxide layer. This decreases the dependency of the radiolytic reaction on the thickness of the oxyhydroxide layer. For a nominal scenario of stored ATR fuel, assuming the chemi-/physio- sorbed water have been fully removed, the internal canister pressure increases to 1.61 atm over a 50 year period, with a hydrogen mole percentage of 21%. As in previous modeling, any oxygen present is in negligible amounts (<1 ppt). If a small amount residual air is present, nitric acid can form up to 1300 ppm. For a scenario with high fuel decay heat, the model shows internal pressure increasing to 2.1 atm, with 39.3 mole percentage of hydrogen. In a scenario where significant chemi-/physio- sorbed water is present within the corrosion layer, the nominal scenario shows a pressure increase to 2.54 atm, with 21.1 mole percent hydrogen. The high decay heat case shows a pressure increase to 3.18 atm with 39.9 mole percent hydrogen.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Studies on the Vitrification Potential of High Risk Spent Nuclear Fuels

This research explored vitrification of U-Zr nuclear fuel that exhibits an explosion hazard when processed by common nitric acid dissolution. Glass frits were chosen based on properties that had been previously measured as well as being demonstrated successfully in production at the Defense Waste Process Facility (DWPF). Zirconium and Aluminum metal powders were chosen to represent the spent fuels after literature demonstrated that the potential phase that causes the issue for explosive reaction in the effluent stream is tied to the Zirconium-III phase field. DSC measurements were carried out on various simulants and waste forms to better understand thermal behavior of these materials. Zirconium (Zr) was tested with both nitric acid and the new effluent flowsheet that is under development by the SRNL Chemical Process Control (CPC) group. The nitric acid test did show the expected exothermic behavior as well as how the behavior is affected by particle size of the powders used. In contrast, the effluent samples did not show an exothermic peak. However, the concentration of metal in the samples was orders of magnitude lower than in the nitric acid testing and further testing of the effect of concentration on the exothermicity that could be measured confirmed a dampened signal. Tests were run that included frit, both DWPF Frit 510 and Iron Phosphate frit, exhibiting complex curves that included multiple endothermic and exothermic peaks during heating and cooling. The final crucible samples appear similarly to the bulk glasses pointing to these peaks being a part of the dissolution of the simulant into the base glasses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Calculating Potential Radiological Emissions for Waste Management Activities at INL - 20068

At Idaho National Laboratory (INL), work involving radionuclides is evaluated for potential emissions from a project in order to comply with the National Emission Standards for Hazardous Air Pollutants (NESHAP) regulations, 40 CFR 61 Subpart H. Emission calculations are documented in an Air Permitting Applicability Determination (APAD) to analyze unmitigated and mitigated emissions and determine if an Application to Construct (ATC) or continuous monitoring is required. To calculate the unmitigated and mitigated emissions, a spreadsheet was developed to provide ease in determining potential emissions by providing the maximum operating temperature and the material being used. The spreadsheet aids in determining the potential emissions for research projects and waste management activities at Materials and Fuels Complex (MFC) and other locations across the INL site. Furthermore, it can also be used for periodic confirmatory measurements (PCM) to justify low emissions. Elements that factor into the unmitigated and mitigated calculations include the amount of each radionuclide used (in curies or grams), specific activity (if amount is given in grams), the temperature the material is heated to in Celsius, the dose conversion factor which is derived from Clean Air Act Assessment Package - 1988 (CAP-88) modeling, and the number of HEPA filters used for mitigated measures. The main drivers for calculating the unmitigated emissions for a project are the amount used per radionuclide, the maximum operating temperature, and the location of the work. The maximum operating temperature determines the airborne release factor which is dependent on the physical state of the radionuclide. Prior to October 2017, if the radionuclide was heated to greater than 100 deg. C, the radionuclide was assumed to be a gas, which has the highest airborne release factor. This assumption would be overly conservative for radionuclides with high melting and boiling points, which provided a challenge to demonstrate low emissions. In October 2017, the Environmental Protection Agency (EPA) approved an alternative method for INL. This method allows the airborne release factor to be determined by using the melting point and 90% of the boiling point of the radionuclide. This methodology was included in the spreadsheet to allow unmitigated emission calculations for APADs to be completed more efficiently and effectively. Results show a reduction in time completing air emission calculations as well as lower total emissions across all facilities at INL. MFC annual emissions were reduced by 62% from the previous year and Research and Education Campus (REC) facilities were reduced by 38% due to implementation of the approved alternative method. Time spent on APADs, PCMs, and documentation for the annual NESHAP report was also reduced significantly. The spreadsheet provided in Table I provides the potential emission calculations for the 'Advanced Retrieval and Disposition Techniques for Remote Handled Mixed Low Level Waste (RH MLLW) at the Radioactive Scrap and Waste Facility (RSWF)' project. Calculations show the Potential Effective Dose Equivalent (PEDE) at RSWF to be 7.27 E-04 mrem/yr (7.27 E-09 Sv/yr) which is well below the 0.1 mrem/yr threshold. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 1, January-March 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Coplnger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.

05 NUCLEAR FUELS↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 1, January-March 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Copinger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Scoping Thermal Response Calculations of RNS Waste During Transport to and Disposal at the WIPP

Sandia National Laboratories (SNL) was contracted by the United States Department of Energy Environmental Management (DOE-EM), Los Alamos Field Office to perform mechanical and thermal scoping calculations as part of a study seeking to understand the ignitability risk of the Remediated Nitrate Salts (RNS) waste drums during transportation from the Waste Control Specialists (WCS) facility to Waste Isolation Pilot Plant (WIPP) and permanent disposal of the waste at WIPP. The scoping thermal simulations described in this report pertain to thermal calculations performed with a packaging system consisting of one Standard Waste Box (SWB) loaded with drums placed inside a Standard Large Box 2 (SLB2). During transportation, the SLB2 is inside Transuranic Package Transporter Model III (TRUPACT-III), which provides the third layer of the packaging. Once at the WIPP, it is assumed the SLB2 is extracted from the TRUPACT-III and maintained above ground, and then subsequently placed underground for permanent disposal. In these proposed configurations, the space between the SLB2 and the SWB is always filled by a layer of insulation consisting of air-filled glass microbubbles except for the bottom which rests directly on the SLB2. The thermal scoping calculations described in this report specifically address whether the introduction of external heat inputs, combined with the contributions from the internally generated radiolytic decay heat and chemical reactions, lead to an unstable thermal state during the time of its movement and placement in the permanent disposal location. The external heat inputs are of two forms: 1) ambient thermal irradiation (e.g., solar and ambient storage/disposal temperatures) and 2) accident-induced fire. Three scoping calculation scenarios were derived as representative, conservative scenarios: 1A) TRUPACT-III transient transportation, 1B) SLB2 48-hour outdoor storage with solar radiation, and 2) fully-engulfing fire during SLB2 handling or emplacement following a steady-state analysis in a 38 °C environment. All the simulated scenarios are conservative relative to the operational conditions expected for handling the waste package during transportation and placement in the WIPP underground disposal unit. The predictions obtained from simulating the three exposure scenarios revealed that adding the SLB2 and the air-filled glass microbubbles to the transport and storage/disposal configurations provides additional thermal protection of the drums beyond what the SWB provides alone, both during long-term above ground insolation and underground during a fire accident. Under the current transportation/storage/disposal concepts, the degree of protection provided by the packaging concept is sufficient to prevent the waste from being ignitable. The simulation results demonstrate that there is adequate margin to safely transport and place the RNS waste from WCS to the WIPP under the current operational concept.

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DOE OSTI2024

Oxygen Limit for Flammability of a Gas Mixture of Helium – Hydrogen – Oxygen in the U.S. Department of Energy Standard Canister

The radiolytic breakdown of residual free, physisorbed, and chemisorbed waters is anticipated in the sealed dry storage of spent nuclear fuel (SNF), releasing H 2 gas into the initial cover gas of helium. For aluminum-clad SNF (ASNF) in particular, the expected total hydrogen in the canister is enough to dominate the cover gas mixture if fully released, and based on experimental rates of H 2 generation, H 2 percentages exceeding the lower flammability limit appear likely. However, no evidence of radiolytic generation of O 2 has been observed to date, so the potential for flammability is averted due to the lack O 2 . Lack of oxygen has been credited for avoiding flammability in some existing storage and transportation approaches for SNF and other waste. This report reviews literature on flammability limits with a focus on the limiting oxygen concentration to provide technical underpinnings to refine a flammability criterion for gas mixtures of helium, hydrogen, and oxygen where the oxygen is expected to be the limiting species. While there is general consensus on H 2 flammability limits at standard temperature and pressure, the decay heat and gas generation within ASNF sealed dry storage canisters could push them to elevated temperatures and pressures, so the goal is to identify an oxygen threshold that will bound the behavior over the temperature and pressure range of the U.S. Department of Energy (DOE) Standard Canister. The information reviewed in this document supports the criterion that the oxygen concentration in a He-H 2 -O 2 mixture should be less than 1.5% to avoid flammability at temperatures and pressures relevant to sealed dry storage of ASNF in the DOE Standard Canister, i.e., up to 343°C (650°F) and 500 psig, respectively. Note that while data covering the full range of temperature at atmospheric pressure and nearly the full range of pressure at room temperature were available, data for the combined effects was available only at moderate temperatures and pressures; the recommended 1.5% O 2 threshold is based on the breadth of available data reported here and apparent trends therein.

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