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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

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DOE OSTI2020

Shaker Table Test Plan

Currently, spent nuclear fuel (SNF) is stored in onsite independent spent fuel storage facilities (ISFSIs), which is a dry storage facility, at 55 nuclear power plant sites. The majority of SNF in dry storage is in welded metal canisters (2,917 canisters at the end of 2019). The canisters are loaded for storage in storage overpacks (vertical casks or horizontal storage modules) and placed on outdoor concrete pads. Because the SNF will be stored at ISFSIs for an extended period of time, there is growing concern with regards to the behavior of the SNF within these dry storage systems during earthquakes. To address these concerns, the SFWST program is considering conducting an earthquake shaker table test. The goal of this test is to determine the strains and accelerations on fuel assembly hardware and cladding during earthquakes of different magnitudes to better quantify the potential damage an earthquake could inflict on spent nuclear fuel rods. The seismic integrity of the storage system has been addressed in the past by the US Nuclear Regulatory Commission and is not the focus of this potential test. Instead the DOE would benefit from knowing the condition of the fuel cladding from storage, transportation, to disposal so that it can ascertain repository performance for the fuel and packaging in its final state. A seismic event is part of the possible loading events that the fuel could experience in its lifetime. This report proposes several earthquake shaker table tests with different degrees of complexity. Alternative 1 was defined in the FY20 work scope. Alternatives 2 and 3 were recently developed to take advantage of the NUHOMS 32PTH dry storage canister that may be available in FY21 for this test at a minimum cost to the project. The selection of the alternative(s) will depend on the available budget and the SFWST program priorities for the near future.

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DOE OSTI2021

SB10 Frit Recommendation, and Evaluations of the Glass Variability Study and Cs-137 Concentrations in Strip Effluent Based on May 2021 Projections

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, and 26F, Alternate Feed Stock-2 (AFS-2) and Sodium Reactor Experiment (SRE) material from H-Canyon. In support of SB10 qualification, frit development using 2020 Tank 40 blend projections and experimental work for the glass variability study were previously conducted. Frit 473 and Frit 209 were identified as candidate frits and both were included in the development of the variability study test matrix; however, a final frit recommendation was postponed until more information could be determined about the composition of SB10 after washing. In May 2021, Savannah River Remediation (SRR) reprojected SB10 based on the analytical results from the Tank 51 qualification sample that was washed in the Savannah River National Laboratory (SRNL) Shielded Cells Facility.

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DOE OSTI2022

Idaho National Laboratory CY 2021 National Emission Standards for Hazardous Air Pollutants Analysis, Methodology and Results for Radionuclides

This report details calculations of potential dose at public receptor locations surrounding the Idaho National Laboratory (INL) Site boundary, and INL in-town facilities, from radionuclides reported to be in use and potentially emitted from INL facilities during calendar year (CY) 2021. All calculations were performed in accordance with the requirements in Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H). Modeling methodology, model input parameters, and contribution to dose by facility, source, and radionuclide at the maximally exposed individual (MEI) location are also discussed. The information in this report supports the “National Emission Standards for Hazardous Air Pollutants - Calendar Year 2021 INL Report for Radionuclides” (INL 2022). In CY 2021, the estimated annual potential dose at the INL Site MEI location was 6.67E-02 mrem/yr, up slightly from the previous year, but far less than the regulatory standard of 10 mrem/yr (CFR 40 Part 61, Subpart H). Approximately 97% of the total dose to the INL Site MEI originated from MFC sources. Emissions from INL in-town facilities resulted in an estimated annual potential dose of 6.21E-03 mrem/yr to the MEI, down 40% from the CY 2020 estimated dose. Year-to-year variations in estimated annual dose can be attributed to adjustments in laboratory operations, changes to facility infrastructure, and variation in meteorological conditions.

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DOE OSTI2023

Idaho National Laboratory CY 2022 National Emission Standards for Hazardous Air Pollutants Analysis, Methodology and Results for Radionuclides

This report details calculations of potential dose at public receptor locations surrounding the Idaho National Laboratory (INL) Site boundary, and INL in-town facilities, from radionuclides reported to be in use and potentially emitted from INL facilities during calendar year (CY) 2022. All calculations were performed in accordance with the requirements in Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H). Modeling methodology, model input parameters, and contribution to dose by facility, source, and radionuclide at the maximally exposed individual (MEI) location are also discussed. The information in this report supports the “National Emission Standards for Hazardous Air Pollutants – Calendar Year 2022 INL Report for Radionuclides” (DOE-ID 2023). In CY 2022, the estimated annual potential dose at the INL Site MEI location was 1.78E-02 mrem/yr, down from the previous year, and far less than the regulatory standard of 10 mrem/yr (CFR 40 Part 61, Subpart H). Approximately 87% of the total dose to the INL Site MEI originated from Materials and Fuels Complex sources. Emissions from INL in-town facilities resulted in an estimated annual potential dose of 4.03E-03 mrem/yr to the MEI, down 35% from the CY 2021 estimated dose. Year-to-year variations in estimated annual dose can be attributed to adjustments in laboratory operations, changes to facility infrastructure, and variation in meteorological conditions.

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DOE OSTI2024

Project Report for Cold Spray Phase 2 (Tasks 9-20)

In support of the ongoing environmental protection mission at the Hanford Site, led by Washington River Protection Solutions (WRPS), the cold spray team has continued development of the tank refurbishment technology. Orano Federal Services, along with subcontractors VRC Metal Systems (VRC) and Robotic Technologies of Tennessee (RTT), have worked with WRPS experts to refine the tools demonstrated in Phase 1 of this project, with the potential to refurbish double shell tanks (DST)s on the Hanford Site, extending the service lives of these vital waste storage facilities.

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DOE OSTI2021

On the dissolution of a borosilicate glass with the use of isotopic tracing – Insights into the mechanism for the long-term dissolution rate

An understanding of the mechanisms responsible for controlling the long-term corrosion rate of nuclear waste glass is paramount if reliable glass dissolution models are to be used to calculate the controlled release of radionuclides from nuclear waste glass under geologic disposal conditions. Additionally, understanding silicate glass dissolution is also important for natural glasses to understand the role from the dissolution of these glasses has on the composition of natural aquatic systems. Two general mechanisms appear to be responsible for the elemental release from glass to the surrounding biosphere - ion exchange (release principally of alkalis) and matrix dissolution (release of the structural components of the glass). Key unknowns related to these mechanisms are the impact of surface layers on the altered glass and how these layers participate in corrosion. To better understand the impact of these layers on glass corrosion, we made two glasses of the same elemental composition, but with selected elements enriched in specific isotopes, e.g. more 10 B than 11 B. We reacted these glasses for 1 y then the solutions from each were exchanged and the chemistry of the solutions and the solids were followed for another 2 y and in some cases 3 y. This allowed us to follow changes in the glasses with respect to these elements and how these elements were transported through the alteration layers with time and, for at least one element, into the glass. Results from these experiments demonstrate that the release behavior of different elements is strongly dependent on their structural role in the glass (e.g. network formers and modifiers), but, more importantly, the role of water transport and subsequent ion exchange has in the long-term dissolution of the glass studied here. In this article, we highlight the behavior of four elements: lithium, sodium, silicon, and boron. Lithium and sodium, network modifying elements, have similar chemistries. The behavior of Li could be tracked in more detail by following both 6 Li and 7 Li; Li penetrates through the gel layer in both directions without hinderance and into and released from the glass deeper than sodium. Silicon, a network forming element, reacts with the silica-rich alteration layer. Boron, a network forming element, does not accumulate in the gel or the pristine glass and has a very sharp elemental profile between the pristine glass and alteration layer. Boron, lithium, and sodium elemental profiles suggest that there is little if any transport control within the alteration layer. The correlation between the Li and Na profiles and that of a water species suggest that the limiting release is a steady state between the diffusion of a water species and matrix dissolution that results in a low release rate for alkali and this provides a steady driving force for the long-term dissolution of glass.

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DOE OSTI2022

Sensitivity study of coupled chemical-CFD simulations for analyzing aluminum-clad spent nuclear fuel storage in sealed canisters

We report the United States Department of Energy (DOE) manages over 50 Metric Tons Heavy Metal (MTHM) of aluminum-clad spent nuclear fuel. One main source for DOE’s Aluminum-clad spent nuclear fuel (ASNF) inventory is the advanced test reactor (ATR) at the INL site, which makes this fuel of particular interest for storage scenarios. Road-ready and final disposition packaging configurations for the ATR fuel dictates storage within helium-backfilled, sealed DOE standard canisters. The conditions within these sealed canisters for extended (greater than50 year dry) storage is of interest. To further this goal, a three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model is developed of the sealed DOE standard canisters. This 3D CFD model is one-way coupled with bulk gas radiolysis reactions considering sealed canisters with inert gas and possible trace amounts of air and water vapor. This study looks at the evolution of the thermal history of the canisters over a 50 year time period with a coupling to the chemical reactions occurring from radiolytic breakdown of residual water. A sensitivity study is then carried out over the parameters of the model including the fuel decay heat, residual water content, sealed pressure, canister external temperature, and canister emissivity. In pure helium, hydrogen generation rates are low, under 10 ppm, but hydrogen generation rates are affected greatly by the presence of even 1% residual air, increasing by 50-plus-fold, and nitric acid generation with residual air also occurs ranging from 500 to 4000 ppm after 50 years. The fuel decay heat and the residual water content show the most importance in the generation of hydrogen gas in pure air, and for nitric acid with a residual air condition. External temperature, canister emissivity and sealed pressure all show minor sensitivity effects to the generation of potentially harmful species.

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DOE OSTI2022

The Thermodynamics of Crystallization and Phase-Separation in Melt-Derived Nuclear Waste Forms

This project aimed to gain detailed and fundamental understanding of the stability of glass waste forms to be used for safe and reliable encapsulation of nuclear waste. The focus was to understand thermodynamic and structural properties involved in crystallization and phase separation in melt-derived waste form glasses and glass-ceramics by studying a baseline glass composition with varying additives. This was achieved by combining neutron total scattering, advanced calorimetric measurements, and extensive modeling study. This project harnessed diverse and complementary approaches and expertise from four collaborating interdisciplinary institutions (two universities, one national laboratory, and one UK university partner). Samples were prepared at the national laboratory and exchanged with the two university institutions. The thermodynamic data obtained by various advanced calorimetry techniques were linked to the underlying short-range structure of waste forms measured by state-of-the-art neutron total scattering experiments. An important element of this project was extensive modeling to interpret and fundamentally understand the experimental results.

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DOE OSTI2022

CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.

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DOE OSTI2023

Analysis of Tank 38h (HTF-38-23-19, -20) and Tank 43H (HTF-43-23-21, -22) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most soluble species in the Tank 38H surface sample decreased significantly from the previous surface sample. The Tank 38H sub-surface sample shows changes in concentration for soluble species in the solution with some increasing and some decreasing. The current Thank 38H sub-surface sample contains visible sludge solids similar to the previous sample, i.e., less than 1%, based on visual appearance. The significant differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species between these two locations within the tank. Savannah River Mission Completion personnel indicate that ~ 150,000 gallons of Tank 22 supernate were received in Tank 38H since its last analysis during which time the 2H evaporator was not operated, so the observed stratification is expected. The Tank 43H surface and sub-surface samples are similar in composition to the previous samples. The similar solution compositions measured in the Tank 43H surface and sub-surface samples indicate a minimal stratification within the tank. The total uranium and plutonium in the current Tank 38H surface sample remains similar to the previous analysis. The Tank 38H sub-surface sample shows an increase in uranium and plutonium concentrations compared to the previous sample likely because of an increase in sludge solids in the current sample. The total uranium concentration in the Tank 43H surface sample is similar to the previous sample results while the plutonium in the Tank 43H sub-surface sample increased relative to the previous analysis. The sum of the major cations versus the sum of the major anions shows a difference of <10% for both samples from Tank 38H sub-surface sample (174 mg/L) increased compared with the previous sample. The Tank 38H surface sample silicon concentrations (27.5 mg/L) is one-half of the previous sample results. The Tank 43H surface sample silicon concentrations compared to the previous sample results indicate Si in the concentration range of 60 to 83 mg/L. Thus, these current samples analyzed from Tanks 38H and 43H show overall silicon concentrations ranging from 27.5 to 174 mg/L.

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DOE OSTI2023

Modeling DOE Standard Canister Configurations with Updated Surface Chemistry

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. These models have coupled the temperature conditions to both the gas phase radiolysis chemistry and the surface chemistry associated with the oxyhydroxide layer. Continued experimental work has identified trends for aluminum surrogate samples with oxyhydroxide layers for tests done at higher dose rates. At low initial doses a fast generation rate of hydrogen occurs which starts to roll over to a lower generation rate as the total dose applied in increased. A previous study created a small-scale chemical model was built to replicate a mini-canister surrogate system at SRNL as well as for the smaller capsule tests performed at INL. The previous iteration of the model utilized step function for its G-values for the hydrogen generation over a 200-year period. This model makes an update to the surface chemistry to account for theorized surface chemistry reactions allowing for oxygen to remain bounded to the oxyhydroxide layer. As additional experiments continue, the kinetic fits may be adjusted to adapt to new data. Three canister configurations are modeled – the base 18-inch, 15-foot DOR standard canister with 30 ATR fuel elements, an 18-inch, 10-foot DOE standard canister with 32 ATR fuel elements and a 24-inch, 10-foot DOE standard canister with 40 ATR fuel elements. In previous reports, the primary sensitivity of the canister conditions was identified as the decay heat of the fuel and the dried condition of the fuel. Only these parameters are studied in the report. For the nominal case with dried fuel, the hydrogen generation is only about 2%, so it is even less than the flammability limit if it were exposed to oxygen. For the densely packed 18-inch case the total hydrogen concentration is only around 4% with the nominal decay heat. For the densely packed 24-inch case the total hydrogen concentration is only 2.5%, roughly 20% higher than the original packing design, and still under flammability conditions if exposed to oxygen.

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DOE OSTI2023

Initial Site-Specific De-Inventory Report for Crystal River

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the Crystal River 3 Nuclear Power Plant (CR-3) independent spent fuel storage installation (ISFSI) site.

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DOE OSTI2023

Initial Site-Specific De-Inventory Report for La Crosse

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the La Crosse Boiling Water Reactor (LACBWR) independent spent fuel storage installation (ISFSI) site.

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DOE OSTI2023

Initial Site-Specific De-Inventory Report for Rancho Seco

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the independent spent fuel storage installation (ISFSI) site at the former Rancho Seco Nuclear Generating Station (RSNGS).

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DOE OSTI2024

Initial Site-Specific De-Inventory Report for San Onofre

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the San Onofre independent spent fuel storage installation (ISFSI) site.

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DOE OSTI2024

Study of Mechanical Properties, Microstructure, and Residual Stresses of AISI 304/304L Stainless Steel Submerged Arc Weld for Spent Fuel Dry Storage Systems

The confinement boundaries of spent nuclear fuel (SNF) canisters are typically fusion welded. Welded microstructures, strain hardening, and residual stresses combined with a chemically aggressive, chloride-rich environment led to concerns that the welded canister may be susceptible to chloride-induced stress corrosion cracking (CISCC). A comprehensive understanding of the modification of stainless steel (SS) metallurgical and mechanical properties by fusion welding could accelerate the predictive analysis of CISCC susceptibility. This paper describes a submerged arc welding (SAW) procedure that was developed and qualified on 12.7 mm (0.5 in.) thick AISI 304/304L SS to produce joints in a way similar to actual SNF canister manufacturing. This procedure has the potential to reduce the production cost and weld CISCC susceptibility by using fewer welding passes and lower heat input than current industrial applications. Global and local mechanical behaviors and properties, as well as residual stress distributions on the welded joint, were studied. The results indicate that hardness values in the fusion zone (FZ) and heat-affected zone (HAZ) are slightly higher than that of the base metal. Strain localization was presented in the HAZ before the tensile stress reached its maximum value, and then it shifted to the FZ. The specimen finally broke in the FZ. High tensile residual stresses exhibited in the FZ and the nearby HAZ suggest the highest CISCC-susceptible spots. The maximum tensile residual stresses were along the welding direction, indicating that if cracks occur, they would be perpendicular to the welding direction. This study involved developing and qualifying a SAW procedure for SNF canister production. The new procedure yielded cost savings (SAW working efficiency increased by about 80%), improved mechanical properties, and presented moderate residual stresses. Analysis revealed that the welded joint’s low-stress and high-stress damage assessments may be affected by shifts in the strain localization spot under loading.

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