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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Unveiling the Role of Critical Impurities in Spent LiFePO 4 Cathodes for Scalable Direct Regeneration

Direct regeneration offers a promising alternative to recycling End-of-Life (EoL) batteries by restoring metal elements and preserving the material structure, yet scaling these technologies to handle practical cathode black mass (CBM) with impurities remains challenging. Here, this study investigates the evolution of impurities, including aluminum (Al), polyvinylidene difluoride (PVDF) binder, and residual carbon (C), during direct recycling of spent LiFePO 4 (LFP) cathodes and their impact on electrochemical performance. Using various ex situ and in situ analyses, it is shown that the formation of lithium fluoride (LiF) during the traditional direct recycling process hinders lithium diffusion and deteriorates the reversible capacity. To address this major challenge, the combination of pH-controlled hydrothermal purification and the two-step sintering process is proposed effectively to regenerate spent LFP cathodes, eliminating the negative effect of Al and fluorine (F) impurities while mitigating any potential impacts of carbon residuals. The regenerated LFP from spent CBM achieves superior performance, retaining 152.5 mAh g −1 at 0.1 C and 133 mAh g −1 at 1 C with 98.7% capacity retention after 200 cycles. This approach is further validated using three distinct waste feedstocks from battery modules, enhancing impurity management and scalability in direct recycling. These findings present a sustainable and economically viable solution for large-scale LFP regeneration.

25 ENERGY STORAGE↗
DOE OSTI2025

Biotransformation of Phenolics in Spent Liquor from Aqueous Ammonia Pretreatment

Spent liquors of biomass pretreatment provide a source for renewable chemical production. These liquors require treatment before being discharged; otherwise, they negatively impact the environment. Herein, spent liquors from aqueous ammonia pretreatment of poplar wood are characterized for phenolic content via liquid chromatography–mass spectrometry and nuclear magnetic resonance spectroscopy. The main phenolics are phenol, p-hydroxybenzamide (pHBAm), and p-hydroxybenzoic acid (pHBA), of which pHBAm and pHBA are produced from the ester-linked p-hydroxybenzoates in poplar wood. Phenol is produced from pHBA via decarboxylation. The potential biotransformation of the extracted phenolics into 2-pyrone-4,6-dicarboxylic acid (PDC) is assessed using an engineered strain of Novosphingobium aromaticivorans DSM12444 (PDC strain). Biotransformation of pHBAm to PDC is shown to be possible in the presence of pHBA, but not when pHBAm is the sole phenolic substrate, this is the first reported observation of N. aromaticivorans producing PDC from an aromatic amide. The phenol present is not transformed to PDC and does not inhibit PDC production. This study demonstrates that the phenolic amide in spent liquor from ammonia pretreatment can be valorized via biotransformation using N. aromaticivorans, which adds to the growing versatility of N. aromaticivorans as a microbial chassis for converting plant-derived compounds to useful products.

biomass↗
DOE OSTI2016

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2014

Determining heavy metals in spent compact fluorescent lamps (CFLs) and their waste management challenges: Some strategies for improving current conditions

Highlights: • Heavy metals in spent compact fluorescent lamps (CFLs) determined. • Current waste management condition of CFLs in Iran assessed. • Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. • We propose extended producer responsibility (EPR) for CFLs waste management. - Abstract: From environmental viewpoint, the most important advantage of compact fluorescent lamps (CFLs) is reduction of green house gas emissions. But their significant disadvantage is disposal of spent lamps because of containing a few milligrams of toxic metals, especially mercury and lead. For a successful implementation of any waste management plan, availability of sufficient and accurate information on quantities and compositions of the generated waste and current management conditions is a fundamental prerequisite. In this study, CFLs were selected among 20 different brands in Iran. Content of heavy metals including mercury, lead, nickel, arsenic and chromium was determined by inductive coupled plasma (ICP). Two cities, Tehran and Tabriz, were selected for assessing the current waste management condition of CFLs. The study found that waste generation amount of CFLs in the country was about 159.80, 183.82 and 153.75 million per year in 2010, 2011 and 2012, respectively. Waste generation rate of CFLs in Iran was determined to be 2.05 per person in 2012. The average amount of mercury, lead, nickel, arsenic and chromium was 0.417, 2.33, 0.064, 0.056 and 0.012 mg per lamp, respectively. Currently, waste of CFLs is disposed by municipal waste stream in waste landfills. For improving the current conditions, we propose by considering the successful experience of extended producer responsibility (EPR) in other electronic waste management. The EPR program with advanced recycling fee (ARF) is implemented for collecting and then recycling CFLs. For encouraging consumers to take the spent CFLs back at the end of the products’ useful life, a proportion of ARF (for example, 50%) can be refunded. On the other hand, the government and Environmental Protection Agency should support and encourage recycling companies of CFLs both technically and financially in the first place.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries

Highlights: • Selective precipitation and solvent extraction were adopted. • Nickel, cobalt and lithium were selectively precipitated. • Co-D2EHPA was employed as high-efficiency extraction reagent for manganese. • High recovery percentages could be achieved for all metal values. - Abstract: Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe–Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC{sub 2}O{sub 4}⋅2H{sub 2}O and Li{sub 2}CO{sub 3} using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor.

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DOE OSTI2017

Use of glucose as reductant to recover Co from spent lithium ions batteries

Highlights: • Glucose was innovatively used as alternative to H{sub 2}O{sub 2} in leaching of spent LiCoO{sub 2} cathode material. • Over 98% Co and nearly 100% Li are leached with 1.5 mol/L phosphoric acid and 0.02 mol/L glucose at 80 °C in 2 h. • A molecular structure diagram was made to show dissolution process of reductive of Co(III) to Co(II) with glucose oxidation. • Cobalt oxalate (purity) was obtained as the precipitating product. • There are no reports on the leaching solution of glucose and phosphate acid. - Abstract: A hydrometallurgical leaching process has been developed for recovery of Co and Li from cathode material (LiCoO{sub 2}) collected from spent LIBs using a mix solution of glucose and phosphoric acid. The spent LiCoO{sub 2} before and after leaching process are analyzed by scanning electron microscopy. A leaching rate of about 98% Co and nearly 100% Li is presented with 1.5 mol/L phosphoric acid and 0.02 mol/L glucose at 80 °C in about 2 h. During leaching process, glucose was oxidized into monocarboxylic acid with reduction of Co(III) to Co(II). Co in solution was recovered as Co-oxalate after leaching process. Using glucose as reductant to dissolve LiCoO{sub 2} with chelating agent of phosphoric acid is achieved here.

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DOE OSTI2017

Recovery of zinc and cadmium from spent batteries using Cyphos IL 102 via solvent extraction route and synthesis of Zn and Cd oxide nanoparticles

Highlights: • First time Cyphos IL102 explored for recovery of Zn/Cd from spent batteries. • Recovery of metals with high purity was achieved at room temperature. • Quantitative extraction was achieved in two stages for Zn and Cd. •Loaded organic phases were used to synthesize ZnO and CdO nanoparticles. •Simple strippants for stripping isotherm was used. - Abstract: The overall aim of this study is to separate and recover zinc and cadmium from spent batteries. For this purpose Cyphos IL 102 diluted in toluene was employed for the extraction and recovery of Zn and Cd from Zn-C and Ni-Cd batteries leach liquor. The influence of extractant concentration for the leach liquors of Zn-C (0.01–0.05 mol/L) and Ni-Cd (0.04–0.20 mol/L) batteries has been investigated. Composition of the leach liquor obtained from Zn-C/Ni-Cd spent batteries is Zn - 2.18 g/L, Mn - 4.59 g/L, Fe - 4.0 × 10{sup −3} g/L, Ni - 0.2 × 10{sup −3} g/L/Cd - 4.28 g/L, Ni - 0.896 × 10{sup −1} g/L, Fe - 0.148 g/L, Co - 3.77 × 10{sup −3} g/L, respectively. Two stage counter current extraction at A/O 1:1 and 3:2 with 0.04 mol/L and 0.2 mol/L Cyphos IL 102 for Zn and Cd, respectively provide more than 99.0% extraction of both the metal ions with almost negligible extraction of associated metal ions. A stripping efficiency of around 99.0% for Zn and Cd was obtained at O/A 1:1 using 1.0 mol/L HNO{sub 3} in two and three counter current stages, respectively. ZnO and CdO were also synthesized using the loaded organic phase and characterized using XRD, FE-SEM and EDX techniques. XRD peaks of ZnO and CdO correspond to zincite and monteponite, respectively. The average particle size was ∼27.0 nm and ∼37.0 nm for ZnO and CdO, respectively. The EDX analysis of ZnO and CdO shows almost 1:1 atomic percentage.

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DOE OSTI2017

Electrochemical process for electrode material of spent lithium ion batteries

Highlights: • Electro-chemical leaching and separation of Co and Mn from spent LIBs are studied. • Overall recovery (>96%) for Co, Mn and Cu was observed. • Electrolyte pH was critical for the production of good quality Co-metal and EMD. • Environmental friendly and commercially feasible process for recycling spent LIBs. - Abstract: Electrochemical method for recovering cobalt and manganese from electrode materials of spent lithium ion batteries was studied. Electrochemical leaching of cobalt and manganese from electrode material was optimized by varying different process parameters such as time, acid concentration and current density. Both cobalt and manganese could effectively be leached out at a current density of 400 A/m{sup 2} in 3 h using 2 M sulphuric acid. In the subsequent study, the metallic cobalt and electrolytic manganese dioxides was recovered from the leach liquor at 200 A/m{sup 2}, pH 2–2.5 and 90 °C after removing aluminum. The commercial feasibility of the study was tested in pilot scale. Overall recovery of Co, Cu and Mn was above 96%, 97% and 99%, respectively.

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DOE OSTI2020

MECHANICAL SHOCK AND VIBRATION ANALYSIS OF SPENT NUCLEAR FUEL CARRIED BY THE ATLAS RAILCAR

Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.

Klymyshyn, Nicholas A.↗
DOE OSTI2021

THERMAL STRESS ANALYSIS OF A SPENT NUCLEAR FUEL CANISTER

The potential for chloride induced stress corrosion cracking (CISCC) in spent nuclear fuel dry storage canisters is a current topic of research by the US Department of Energy Spent Fuel and Waste Science and Technology program. One of the important prerequisites for CISCC is a tensile residual stress state. This study utilizes computational models to provide an initial analysis of thermal stresses in a generic spent nuclear fuel canister. A STAR-CCM+ thermal fluid model provides a temperature profile analysis of the canister based on four different ambient temperature conditions. An ANSYS APDL finite element model incorporates the temperature profiles to analyze the thermal stresses in the canister. The calculated thermal stress magnitudes are in the range of 10 MPa to 80 MPa, which could be significant for crack propagation through the canister wall. The maximum thermal stress is not a concern for the structural failure of the canister, but it is high enough that it is a significant contribution to the total stress state of the canister wall, which also includes residual stress from fabrication, internal pressure from helium cover gas, and potential transient mechanical loads, such as earthquakes. This paper describes the initial finite element thermal stress analysis that was completed in 2020, reports the initial findings, and identifies areas where model refinement is still needed to complete this work in the future.

Jensen, Ben J.↗
DOE OSTI2022

Large-scale parametric modeling of spent nuclear fuel dynamics in the 30 cm package drop scenario

Packages used to transport spent nuclear fuel (SNF) are required by the U.S. Code of Federal Regulations 10 CFR 71.71 to demonstrate satisfactory performance during a drop scenario. While the CFR is meant to ensure safe package function, it does not evaluate survival of the SNF within. The U.S. Department of Energy Spent Fuel and Waste Science and Technology program is working on closing the knowledge gap related to the response of SNF to external mechanical loads, including the hypothetical 30 cm package drop scenario in the CFR. In support of this effort, LS-DYNA finite element simulations were developed by Pacific Northwest National Laboratory (PNNL) to model generic drop scenarios at both the package and fuel assembly level. The models were validated against one-third scale package and full scale fuel assembly drop test data and were exercised to predict fuel cladding strains in a narrow range of model configurations. This work describes a large-scale parametric study conducted by PNNL using the previously developed and validated PWR finite element model, with the addition of a new generic BWR assembly model. The motivation for the parametric study was to characterize the broad range of SNF responses in the 30 cm package drop scenario. This was accomplished by varying the drop orientation, fuel assembly type (17x17 PWR and 10x10 BWR), burnup, cladding temperature, spacer grid buckling load, package mass, impact limiter stiffness, and mechanical gap conditions within the basket. A MATLAB framework was developed to automate LS-DYNA model generation and execution on PNNL institutional computing resources. In total, over 2000 simulations were performed. For each simulation, the SNF response was quantified in terms of permanent grid deformation, fuel rod contact pressure, and strains within the fuel rods, guide tubes, and water rods. The results provide valuable insight into the range of responses that could be reasonably expected from SNF in the hypothetical drop scenario, as well as the sensitivity to each input parameter. The results of this parametric study are a key component of the testing and modeling strategy the Spent Fuel and Waste Science and Technology program is using to close the external loads knowledge gap.

Kadooka, Kevin↗
DOE OSTI2013

Direct Measurement of U235 and Pu239 in Spent Fuel Rods with Gamma-Ray Mirrors

The amounts of fissile Pu and U in spent nuclear fuel are of primary concern to the safeguards community. In particular, there are issues when safeguards transitions from an item accountancy basis (such as fuel bundles) to a fissile material mass basis as occurs when spent fuel enters a reprocessing plant. Discrepancies occur because item accountancy requires estimating the content of fissile material using indirect techniques such as the fuel burn-up and item-level measurements of radiation emissions from fission by-products. Direct measurement of the fissile content by monitoring line emissions from fissile species themselves is impossible because the lines are much weaker than those emitted by shorter-lived isotopes in the fuel. The goal of this project is to develop a technique to directly measure these weaker lines despite the presence of overwhelming radiation from other isotopes. This is achieved by using gamma-ray mirrors as a narrow band-pass filter. The mirrors reflect only energies of interest toward a HPGe detector that is shielded from direct view of the spent fuel and its fierce emissions. This can significantly improve the reliability with which the mass of fissile material is tracked.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Unreviewed Disposal Question Evaluation: Disposal of the Portsmouth Gaseous Diffusion Plant Spent Trap Media and Other Uranium Bearing Particulate Wastes at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Portsmouth (PORTS) Gaseous Diffusion Plant (GDP) Spent Trap Media and Other Uranium Bearing Particulate Wastes is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the PORTS GDP Spent Trap Media and Other Uranium Bearing Particulate Wastes meets all performance objectives in DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P. The PORTS GDP Spent Trap Media and Other Uranium Bearing Particulate Wastes are recommended for acceptance with the condition that the total technetium-99 (99Tc) activity disposed under this waste profile shall not exceed 2.1E13 Bq.

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DOE OSTI2016

Final Report for FY15 Spent Fuel Project Task 12: Data Authentication

The data authentication task is concerned with the ability of a user to trust the output from an unattended measurement of spent nuclear fuel intended for disposal. Five high-level requirements that are driven from a data authentication perspective need to be considered for instrument development: The instrument must be secured. All data emerging from the instrument must be signed. Instrument inputs may also need to be signed and authenticated. The instrument environment needs to be controlled. A vulnerability assessment will be necessary for the full system before deployment. Additional requirements may be necessary, especially if operational scenarios include arrangements for joint use. None of the prototype instruments currently under development yet satisfies any of these requirements. From a data authentication perspective, the selection of a preferred instrument from among the candidates under development should consider the relative ability to secure the instrument, its relative needs for ancillary instrumentation, and ability of the instrument to self-interrogate its state of health. Further work is recommended to develop advanced technologies, methods and approaches for tamper indication. Beyond the development of any particular instrument for the nondestructive assay of spent fuel, efforts should be devoted to developing the system for spent fuel verification prior to long term disposal, including likely operational scenarios for routine use of the verification system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of Mockups and Instrumentation for Spent Fuel Drying Tests

The purpose of this report is to provide updates on the experimental components, methodology, and instrumentation under development for use in advanced studies of realistic drying operations conducted on surrogate spent nuclear fuel. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. Water remaining in canisters upon completion of drying procedures can lead to cladding corrosion, embrittlement, and breaching, as well as fuel degradation. Additional information is needed on the drying process efficacy to help evaluate the potential impacts of water retention on extended longterm dry storage. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations. Smaller-scale tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report describes the implementation of moisture monitoring equipment on a pressurized, submersible system employing a single waterproof, electrically heated spent fuel rod simulator as a demonstration of analytical capabilities during a drying process. A mass spectrometer with specially designed inlets was used to monitor moisture and other gases at 150 kPa to 800 kPa for a test simulating a forced helium dehydration procedure and below 1 torr for tests mimicking a vacuum drying process. The dew point data from the mass spectrometer was found to be in good agreement with a solid-state moisture probe. A distinct advantage of the mass spectrometer system was the capability to directly sample from the hightemperature (>200 °C) head space expected in a prototypic scale experiment where a solid-state moisture probe would suffer considerable loss of accuracy or fail altogether. The operational and analytical experiences gained from this test series are poised to support an expansion to assembly-scale tests at prototypic length. These assemblies are designed to feature prototypic assembly hardware, advanced diagnostics for in situ internal rod pressure monitoring, and failed fuel rod simulators with engineered cladding defects to challenge the drying system with waterlogged fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

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DOE OSTI2019

Initial Neutronics and Thermal-Hydraulic Coupling for Spent Nuclear Fuel Canister

This report documents work performed supporting the US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.05, “Direct Disposal of Dual Purpose Canisters.” In particular, this appendix fulfills the M3 milestone, M3SF-19OR010305015, “Multiphysics criticality consequence analysis capability development status report,” as Revision 1 to M3SF-19OR010305016, “Initial neutronic and thermal hydraulic coupling for waste package” within work package SF-19OR01030501, “Direct Disposal of Dual Purpose Canisters–ORNL.” This report presents the initial development status of a multiphysics criticality consequence simulation framework, Terrenus. Terrenus currently couples two physics codes: the Monte Carlo radiation transport code Shift and the thermal-hydraulics code COBRA-SFS. The coupling capability has been demonstrated using a simplified 3 x 3 fuel pin lattice. Terrenus will be developed further to include (1) general geometry package support, allowing modeling of a spent nuclear fuel (SNF) cask with all structural details, (2) a depletion solver to determine the change in nuclide composition at the end of each time step, and (3) a mechanics code to determine any structural impact due to a criticality event. Future research and development works include identifying/developing a modern two-phase thermal-hydraulics code and developing an approach to solve fast neutronic transients. The goal of this multiphysics framework is to determine the feasibility of direct disposal of currently loaded SNF canisters by including or excluding a criticality event from a repository performance analysis framework in terms of consequences. The revision 1 to this report includes extending Terrenus to a 17 x 17 pressurized water reactor (PWR) fuel assembly. We also investigated the phenomena necessary to quench a critical reaction, and investigated the water level necessary to achieve criticality in a typical canister. Finally, we describe the initial implementation of a critical power search capability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

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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