Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Forms”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Analysis Results for Sub-Samples of SRS Tanks 30, 32, and 39 to Support Evaluations of the I-129 Inventory

In order to appropriately model and predict the chemical integrity and performance of cementitious materials used for waste immobilization at the Savannah River Site (SRS), it is critical to understand the I-129 solubility and distribution within the tank farm. Iodine in radioactive waste and in environmental media is typically highly mobile and long lived. Iodine is ubiquitous in SRS tank waste and waste forms. The iodine is assumed to be soluble and present at low levels in Performance Assessments (PAs) for SRS Tank Farms, and is one of the dose drivers in the PAs for both the SRS Salt Disposal Facility (SDF) and the H-Area Tank Farm (HTF). Analysis of tank waste samples is critical to understanding the Tank Farm iodine inventory and reducing disposal uncertainty. Higher than expected iodine levels have recently been observed in residual solids isolated from some SRS tanks prior to closure, indicating uncertainty regarding the chemical species involved. If the iodine inventory uncertainty is larger than anticipated, future work may be necessary to reduce the uncertainty. This memorandum satisfies a portion of the work scope identified in Task Plan SRNL-RP-2016-00651. A separate memorandum issued previously, reported historical unpublished I-129 data, a significant portion of which was below detectable analytical limits. This memorandum includes iodine and general chemical analysis results for six archived SRNL samples which were previously reported to have I-129 concentrations below detectable limits. Lower sample dilution factors were used for the current analyses in order to obtain concentrations above detection. The samples analyzed included surface and depth samples from SRS tanks 30, 32, and 39.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

A Brief Review of Technologies for Separating Tritium from the Aqueous Effluent of Reprocessing Plants

The scope of this letter report encompasses a review of technologies applicable to separating tritiated water from light and heavy water with a focus on wastewater arising from the reprocessing of UNF. Reports considering flowsheet options for tritium management are also reviewed. The main objective of this letter report is to support consideration of programmatic opportunities for the Materials Recovery and Waste Form Development (MRWFD) Campaign within the Department of Energy’s Office of Nuclear Energy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Using Cosmic Ray Muons to Assess Geological Characteristics in the Subsurface

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Deep Borehole Disposal Safety Case

The safety case for deep borehole disposal of nuclear wastes contains a safety strategy, an assessment basis, and a safety assessment. The safety strategy includes strategies for management, siting and design, and assessment. The assessment basis considers site selection, pre-closure, and post-closure, which includes waste and engineered barriers, the geosphere/natural barriers, and the biosphere and surface environment. The safety assessment entails a pre-closure safety analysis, a post-closure performance assessment, and confidence enhancement analyses. This work outlines the assessment basis and safety assessment aspects of a deep borehole disposal safety case. The safety case presented here is specific to deep borehole disposal of Cs and Sr capsules, but is generally applicable to other waste forms, such as spent nuclear fuel. The safety assessments for pre-closure and post-closure are briefly summarized from other sources; key issues for confidence enhancement are described in greater detail. These confidence enhancement analyses require building the technical basis for geologically old, reducing, highly saline brines at the depth of waste emplacement, and using reactive-transport codes to predict their movement in post-closure. The development and emplacement of borehole seals above the waste emplacement zone is also important to confidence enhancement.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Retention of LDR Inorganics in Solidified/Immobilized Waste

This report provides a summary of the laboratory data and statistical analysis used to calculate retention factors in solidified/stabilized Hanford tank waste for the 13 LDR inorganic species associated with Hanford tank waste (SRNL-STI-2020-00228). The data presented is a summary of the results of research performed to build a correlation between the untreated waste concentration and the TCLP response of a solidified/stabilized waste form. This report establishes the efficacy of two treatment technologies found in 40 CFR 268.42, CHRED (chemical reduction) and STABL (stabilization), that are key to the Sample-and-Send strategy.

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

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗
DOE OSTI2021

Crystallization Constraints for WTP LAW Operations: Assessment of CCC Impacts on VHT and PCT

Much work has been done to expand the glass composition region available for operation of the Hanford Waste Treatment and Immobilization Plant. This includes the development of updated glass property-composition models as well as constraints. This report supports this effort by suggesting constraints for avoiding excessive, and likely detrimental, crystallization during slow cooling of the low-activity glass waste forms in their containers. The constraints target crystals in the Na-Al-silicate and Na-Ca-silicate families. These types of crystals were found to be potentially detrimental to glass durability as they remove Al and Si from the glass matrix, resulting in poor performance of the residual glass during testing such as the Product Consistency Test and the Vapor Hydration Test. Using previously acquired results and results from testing during this effort, the constraints described in the report were determined, and are suggested as options to reduce the risk of forming crystals of the types and concentrations that are likely detrimental to glass durability.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Verification and Validation of START: A Spent Nuclear Fuel Routing and Decision Support Tool

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Rev. 0)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis. The estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE 435.1 Chg 1 and its companion documents DOE M 435.1-1 Chg 1 and DOE G 435.1-1). All performance objectives are currently satisfied, and operational waste acceptance criteria and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years' evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. Within the active burial grounds, low-level waste and mixed low-level waste currently may be disposed only in two lined trenches in the 218-W-5 Burial Ground (Trenches 31 and 34) until they are either filled or a decision is made to close these trenches. Some mixed low-level waste is also disposed at the Environmental Restoration Disposal Facility in the 200 West Area (which is covered under a separate PA). During this reporting period (fiscal year 2018, from October 1, 2017, through September 30, 2018), waste was disposed to the 200 West Area LLBGs. Results from the multi-year diffusion experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., uranium-238, technetium-99, and iodine-123). The results of sorption experiments initiated in FY 2017 to evaluate sorption coefficients for iodine and technetium are also summarized. The test durations ranged from 1 to 6 months. Continued groundwater monitoring of the 200 West Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 is being maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds (Rev. 0)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis. The estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE O 435.1 Chg 1 and companion documents DOE M 435.1-1 Chg 1 and DOE G 435.1-1). All performance objectives are currently satisfied, and operational waste acceptance criteria and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years' evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. Within the active burial grounds, low-level waste and mixed low-level waste currently may be disposed only in the dedicated U.S. Navy reactor compartment trench in the 218-E-12B Burial Ground (Trench 94). Naval reactor compartment disposal at Trench 94 will continue. During this reporting period (fiscal year 2018, from October 1, 2017, through September 30, 2018), two reactor compartments were disposed in Trench 94. Results from multi-year diffusion experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., uranium-238, technetium-99, and iodine-129). The results of sorption experiments initiated in FY 2017 to evaluate sorption coefficients for iodine and technetium are also summarized. The test durations ranged from 1 to 6 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East Area and 200 West Area in the Central Plateau. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 is being maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Sampling for Confidence: Guidance for Evaluating the Quality of Canister Sampling Plans

There is a need for an inspection frequency that demonstrates reasonable assurance that the confinement function of dual-purpose-canisters (DPCs) is maintained as these DPCs are returned to storage service following transportation. The DPCs contain spent fuel from shutdown reactor sites. The currently-proposed 100% DPC inspection frequency is overly conservative and unnecessary to demonstrate reasonable assurance that the confinement function of all DPCs is maintained for return to storage service. Testing all DPCs does not appropriately reflect the very low risk of leakage (i.e., low probability based on design and low consequences based on analysis). Currently, there are 1066 DPCs forecast to be loaded at reactor sites which either are shutdown or plan to cease operations by 2025. These DPCs may be transported by rail in sets of 3 to 5 DPCs per shipment. Assuming each rail consist originates at a single reactor site, an alternative proposal to test one canister/consist/site would result in 356 to 214 canisters, respectively, being tested given the population of 1,066 canisters This Savannah River National Laboratory (SRNL) report outlines statistically-based, random sampling plans that provide guidance on the number of samples of these canisters that are needed to establish bounds on the quality of the outcome of such sampling. The quality of each proposed sampling plan is expressed in the form of a statement that the number of undetected, unacceptable canisters in the population is less than some number, call it D, at a high level of confidence. In addition, evaluations using a similar approach for random sampling at the proposed levels of 214 and 356 canisters are provided. This is a technical report that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed upon contract amendment. To the extent discussions or recommendations in this report conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Challenges with vitrification of Hanford High-Level Waste (HLW) to borosilicate glass – An overview

The Hanford site in Washington State houses ~56 million gallons of radioactive wastes stored in 177 underground tanks. The waste must be immobilized and permanently stored, and the plan is to separate the tank wastes into low activity waste (LAW) and high-level waste (HLW) streams. The U.S. Department of Energy is building a Waste Treatment and Immobilization Plant at Hanford site to separately vitrify these two waste streams in borosilicate glass using Joule-heated ceramic melters (JHCM). Although the process of nuclear waste vitrification seems to be well established, in practicality, it is faced with complex problems starting from the design of glass compositions, to processing in melters and long-term performance of the final vitrified waste forms. The article presents an overview of our current understanding of critical challenges related to the development and performance of HLW glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Performance Assessment for the E-Area Low-Level Radioactive Waste Disposal Facility at the Savannah River Site: Appendix D

This appendix to Chapter 5, Section 5.1 provides supplemental concentration profiles for radionuclide species in STs and ETs that contribute to at least 0.1% of the sum-of-fractions. All concentrations are reported as pCi L -1 per Ci parent buried. The following nomenclature is used for all radionuclides in all DUs: an uppercase letter suffix indicates a SWF (e.g., I-129G, C-14N, H-3F, etc.), while the absence of an uppercase letter denotes a generic waste form (e.g., I-129, C-14, H-3, etc.).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Understanding Radionuclide Interactions with Cementitious Materials for Radioactive Waste Management - 18298

In the UK, ILW is currently managed using cementitious encapsulation in Portland cement (PC)-based grouts. The long-term behaviour of radionuclides in ILW cementitious waste-forms under varying local conditions of extended interim storage and geological disposal is not well understood and the behaviour of cement minerals in the retention of various radionuclides is poorly constrained. This work focuses on the Ca-Si-[Al,Mg,S] mineral system and how this may evolve with the introduction of radionuclides into the system. Here, we discuss the experimental program aimed at underpinning the uptake and retention mechanisms of a range of key radionuclides (U, Tc, Ce/Pu), with selected cement phases: calcium-silicate-hydrate (C-S-H), hydrotalcite (Mg{sub 6}Al{sub 2}CO{sub 3}(OH)16.4H{sub 2}O) and ettringite (Ca{sub 6}(Al{sub 2}O{sub 6})(SO{sub 4}){sub 3}.32H{sub 2}O). Described here is the syntheses of C-S-H and radionuclide incorporation experiments performed on this phase. Further to this, thermodynamic geochemical modelling has been adopted to explore the Ca-Si-U mineral system, aimed at identifying the mineral phases likely to form on contact of C-S-H phases with aqueous uranium, as a function of pH. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Thermal conversion in air of rare-earth fluorides to rare-earth oxyfluorides and rare-earth oxides

Phase transformations of seven different rare-earth fluorides (i.e., REF3) where RE = La, Ce, Pr, Nd, Tm, Yb, Lu at temperatures ranging from 400–1400°C in air were investigated with X-ray diffraction. All of the REF3 compounds first transformed to oxyfluorides and then to oxides, with the exception of CeF3, which transformed directly to an oxide. This study focuses on the phase transitions of REF3 to REOx by simple heat-treatment processes in air and shows plausibility to remove RE elements from fluoride salt streams from molten salt reactors through fluoride-to-oxyfluoride or fluoride-to-oxide conversion mechanisms, which will result in precipitation. This could be used to remove fission product poisons from molten salt reactor waste streams. A waste form option for the resulting REOx products is lanthanide aluminoborosilicate (LABS) glass. To demonstrate this NdF3 was converted to Nd2O3 and immobilized in a LABS glass.

oxyfluoride, molten salt reactors, fluoride salt w↗