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DOE OSTI2020

Estimation of Sorption Behavior of Europium to Calcium Silicate Hydrate by Thermal Analysis - 20172

High-level radioactive waste generated from reprocessing processes will be disposed in a geological disposal that is deep underground. Calcium silicate hydrate is formed as a secondary mineral in the repository via the reaction of calcium ions leaching from the cement and the silicic acid dissolved from the host rocks. In this study, the sorption behavior of europium, which is also known as a chemical analog of americium, on calcium silicate hydrate was examined. The solid phase was analyzed via thermogravimetry- differential thermal analysis, differential scanning calorimetry, Raman spectroscopy, fluorescence spectroscopy, and fluorescence lifetime analysis. To understand the behavior of calcium silicate hydrate under conditions saturated with groundwater, the study samples were analyzed using the above methods, except for deferential scanning calorimetry, without a drying process. The results of the differential scanning calorimetry analyses showed that the temperatures for the endothermic peaks of calcium silicate hydrate for the lower Ca/Si ratios were larger than those for the higher Ca/Si ratios for coexistent conditions of europium ions. Hence, the relationships of the temperatures of the endothermic peaks and the Ca/Si ratios were reversed by the addition of europium ions. These study results suggest that calcium silicate hydrate can immobilize europium ions, thus preventing the production of colloidal forms. (authors)

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

The Center for Performance and Design of Nuclear Waste Forms and Containers (WastePD) Energy Frontier Research Center (Final Report)

The DOE Office of Environmental Management is responsible for high level nuclear waste that must be safely isolated from humans and the environment for extremely long periods. The waste forms and containers are made of glass, ceramics, and metals. Verified safe disposal requires understanding the fundamental mechanisms of waste form degradation and the design of new waste forms with improved performance, which comprise the goals of the Energy Frontier Research Center known as the Center for the Performance and Design of Nuclear Waste Forms and Containers, WastePD. WastePD was constructed to develop innovative approaches and solutions to those goals through the synergistic interactions of individuals who are experts in the degradation behavior, modeling, and design of glasses, ceramics and metal alloys. WastePD is the first center ever created to address this diverse group of materials in a comprehensive and coordinated manner. The science goals are grouped into three common topics: corrosion mechanisms via advanced characterization, environmental impacts, and materials design. Synergistic interactions in these areas were a key component of WastePD. The fundamental understanding of the degradation mechanisms of the waste forms and containers as well as the development of new materials with improved properties will allow DOE to prevent environmental contamination and to explore totally new repository concepts. WastePD was operational from August 2016 through July 2022, but the DOE support was drastically reduced for the last two years. This final technical report covers the full period of performance. However, much of what was accomplished in the first four years is nicely summarized in a review paper published in 2021, which is appended to this report. Therefore, this final report focuses on the technical findings from the last two years of WastePD activities. Considerable progress was made in the areas of a) the environmental and compositional impacts on the corrosion of borosilicate and aluminosilicate glasses, b) the mechanism of glass corrosion and the structure and evolution of the surface alteration layer, c) the effects of environment and composition on the corrosion of pyrochlore, perovskite, hollandite, and other oxide ceramics, d) the corrosion mechanism of multi-principal element metallic alloys, and e) a new framework for understanding the pitting corrosion of metals.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Evaluation of Nuclear Spent Fuel Disposal in Clay-Bearing Rock - Process Model Development and Experimental Studies

The DOE R&D program under the Spent Fuel Waste Science Technology (SFWST) campaign has made key progress in modeling and experimental approaches towards the characterization of chemical and physical phenomena that could impact the long-term safety assessment of heat-generating nuclear waste disposition in deep clay/shale/argillaceous rock. International collaboration activities such as heater tests and postmortem analysis of samples recovered from these have elucidated key information regarding changes in the engineered barrier system (EBS) material exposed to years of thermal loads. Chemical and structural analyses of sampled bentonite material from such tests has as well as experiments conducted on these are key to the characterization of thermal effects affecting bentonite clay barrier performance and the extent of sacrificial zones in the EBS during the thermal period. Thermal, hydrologic, and chemical data collected from heater tests and laboratory experiments has been used in the development, validation, and calibration of THMC simulators to model near-field coupled processes. This information leads to the development of simulation approaches (e.g., continuum vs. discrete) to tackle issues related to flow and transport at various scales of the host-rock and EBS design concept. Consideration of direct disposal of large capacity dual-purpose canisters (DPCs) as part of the back-end SNF waste disposition strategy has generated interest in improving our understanding of the effects of elevated temperatures on the EBS design. This is particularly important for backfilled repository concepts where temperature plays a key role in the EBS behavior and long-term performance. This report describes multiple R&D efforts on disposal in argillaceous geologic media through development and application of coupled THMC process models, experimental studies on clay/metal/cement barrier and host-rock (argillite) material interactions, molecular dynamic (MD) simulations of water transport during (swelling) clay dehydration, first-principles studies of metaschoepite (UO 2 corrosion product) stability, and advances in thermodynamic plus surface complexation database development. Drift-scale URL experiments provides key data for testing hydrological-chemical (HC) model involving strong couplings of fluid mixing and barrier material chemical interactions. The THM modeling focuses on heater test experiments in argillite rock and gas migration in bentonite as part of international collaboration activities at underground research laboratories (URLs). In addition, field testing at an URL involves in situ analysis of fault slip behavior and fault permeability. Pore-scale modeling of gas bubble migration is also being investigated within the gas migration modeling effort. Interaction experiments on bentonite samples from heater test under ambient and elevated temperatures permit the evaluation of ion exchange, phase stability, and mineral transformation changes that could impact clay swelling. Advances in the development, testing, and implementation of a spent nuclear fuel (SNF) degradation model coupled with canister corrosion focus on the effects of hydrogen gas generation and its integration with Geologic Disposal Safety Assessment (GDSA). GDSA integration activities includes evaluation of groundwater chemistries in shale formations.

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

Solubility of An(OH) 3 am and An(OH) 3 cr in the presence of Citrate and EDTA

The solubility of An(III), which pertains to Pu(III), Am(III), and Cm(III), in brine is an important contributor in WIPP (Waste Isolation Pilot Plant) performance assessment (PA) models to calculate the potential release of transuranic elements in the near-field environment of the WIPP repository. Americium, although present in much smaller quantities than plutonium (~ 336 kg on year 2033 is potentially WIPP-bound primarily as Am-241), has a significant impact on release due to its higher activity during the earlier times of repository history (~26% initially, decreasing to 17% and ~0% at 1000 and 10,000 years after emplacement) (SOTERM, 2019). Lanthanide analogues such as Nd 3+ possess physical and chemical characteristics that allow them to be used to examine the chemical behavior of trivalent actinides (Lucchini, et al., 2007).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Ten-Year Review of US/German Collaboration in Salt Repository Research - 20069

A multinational collaboration on salt repository research, design and operation has enjoyed remarkable success since reinvigorating efforts in 2010. Nations now engaged in the shared salt repository research agenda include Germany, the United States, the United Kingdom, the Netherlands, and Poland. The scientific basis for safe and permanent disposal of nuclear waste in salt formations has been strengthened by annual workshops that recognize and address contemporary research, including breakout sessions to stimulate open discussion and focus planning for ongoing investigations. Collaboration not only identifies pertinent technical issues, but facilitates timely, expert, and cost-effective consideration. Contemporary workshops have been held annually since 2010 and are documented in yearly state-of-the-art Proceedings, which summarize content and conclusions. The Proceedings help preserve scientific understanding and provide timely source references. These workshops often produce valuable joint publications coordinated with the Nuclear Energy Agency and other suitable external forums for dissemination. Nuclear waste management programs face growing challenges, while permanent disposal in salt formations provides a robust, safe option for several nations. Workshop format and publications provide a cost-effective insurance against loss of scientific expertise and institutional memory. This paper summarizes ten US/German workshops since formal reinitiation, reexamines key technical issues, discusses the evolving research agenda, and highlights successes and challenges. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Results from the US-German Benchmark Initiative for FY14

Sandia recently joined the third phase of and is a contributing partner to a U.S.-German “Joint Project” entitled “Comparison of current constitutive models and simulation procedures on the basis of model calculations of the thermo-mechanical behavior and healing of rock salt.” The first goal of the project is to check the ability of numerical modeling tools to correctly describe the relevant deformation phenomena in rock salt under various influences. Achieving this goal will lead to increased confidence in the results of numerical simulations related to the secure storage of radioactive wastes in rock salt, thereby enhancing the acceptance of the results. These results may ultimately be used to make various assertions regarding both the stability analysis of an underground repository in salt during the operating phase and the long-term integrity of the geological barrier against the release of harmful substances into the biosphere in the post-operating phase. Among the numerical modeling tools are constitutive models that are used in computer simulations for the description of the thermal, mechanical, and hydraulic behavior of the host rock under various influences and for the long-term prediction of this behavior into the future. A second goal of the project is to investigate and demonstrate the possibilities for further potential development and improvement of these constitutive models. This report summarizes the efforts undertaken during FY14 in support of these international benchmark calculations of field experiments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Summary Report for the NEPA Impact Analysis. Revision 1

The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of defense-related transuranic (TRU) waste. Containment of TRU waste at the WIPP facility is derived from standards set forth in Title 40 of the Code of Federal Regulations (CFR), Part 191. The DOE assesses compliance with the containment standards according to the Certification Criteria in Title 40 CFR Part 194 by means of Performance Assessment (PA) calculations performed by Sandia National Laboratories (SNL). WIPP PA calculations estimate the probability of radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. The DOE Carlsbad Field Office (CBFO) has initiated a National Environmental Policy Act (NEPA) action for a proposal to excavate and use additional transuranic (TRU) waste disposal panels at the WIPP facility. This report documents an analysis undertaken as part of an effort to evaluate the potential environmental consequences of the proposed action. Although not explicitly required for a NEPA analysis, evaluations of a dose indicator to hypothetical members of the public after final facility closure are presented in this report. The analysis is carried out in two stages: first, Performance Assessment (PA) calculations quantify the potential releases to the accessible environment over a 10,000-year post-closure period. Second, dose was evaluated for three hypothetical exposure pathways using the conservative radionuclide concentrations assumed to be released to the accessible environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

DR Argillite Disposal R&D at LBNL

Within the Natural Barrier System (NBS) group of the Used Fuel Disposition (UFD) Campaign at the Department of Energy’s (DOE) Office of Nuclear Energy, LBNL’s research activities have focused on understanding and modeling EDZ evolution and the associated coupled processes and impacts of high temperature on parameters and processes relevant to performance of a clay repository to establish the technical base for the maximum allowable temperature. This report documents results from some of these activities. These activities address key Features, Events, and Processes (FEPs), which have been ranked in importance from medium to high, as listed in Table 7 of the Used Fuel Disposition Campaign Disposal Research and Development Roadmap (FCR&D-USED-2011-000065 REV0) (Nutt, 2011). Specifically, they address FEP 2.2.01, Excavation Disturbed Zone, for clay/shale, by investigating how coupled processes affect EDZ evolution; FEP 2.2.05, Flow and Transport Pathways; and FEP 2.2.08, Hydrologic Processes, and FEP 2.2.07, Mechanical Processes, and FEP 2.2.09, Chemical Process—Transport, by studying near-field coupled THMC processes in clay/shale repositories. The activities documented in this report also address a number of research topics identified in Research & Development (R&D) Plan for Used Fuel Disposition Campaign (UFDC) Natural System Evaluation and Tool Development (Wang 2011), including Topics S3, Disposal system modeling – Natural System; P1, Development of discrete fracture network (DFN) model; P14, Technical basis for thermal loading limits; and P15 Modeling of disturbed rock zone (DRZ) evolution (clay repository).

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

Experimental and modeling studies of PR and ND oxalate solubility to high ionic strengths: Insight into actinide(III) oxalates

Actinide oxalates are chemical compounds important to nuclear industry, ranging from actinide separation in waste reprocessing, to production of specialty actinides, and to disposal of high level nuclear waste (HLW) and spent nuclear fuel (SNF). In this study, the solubility constants for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O by performing solubility experiments in HNO 3 and mixtures of HNO 3 and H 2 C 2 O 4 at 23.0 ± 0.2 °C have been determined. The targeted starting materials, Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O, were successfully synthesized at room temperature using PrCl 3 , NdCl 3 and oxalic acid as the source metrials. Then, we utilized the targeted solubility-controlling phases to conduct solubility measurements. There was no phase change over the entire periods of experiments, demonstrating that Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O were the solubility-controlling phases in our respective experiments. Based on our experimental data, we have developed a thermodynamic model for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O in the mixtures of HNO 3 and H 2 C 2 O 4 to high ionic strengths. The model for Pr 2 (C 2 O 4 ) 3 •10H 2 O reproduces well the reported experimental data for Pu 2 (C 2 O 4 ) 3 •10H 2 O, which are not utilized for the model development, demonstrating that Pr(III) is an excellent analog for Pu(III). Similarly, the model for Nd 2 (C 2 O 4 ) 3 •10H 2 O reproduces the solubility of Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The Pitzer model was used for the calculation of activity coefficients. Based on the published, well established model for dissociation constants for oxalic acid and stability constants for actinide-oxalate complexes [i.e., AmC 2 O 4 + , and Am(C 2 O 4 ) 2 - ] to high ionic strengths, we have obtained the solubility constants (log 10 K 0 ) for the following reactions at 25 °C, Pr 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Pr 3+ + 3C 2 O 4 2- + 10H 2 O(l). Nd 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Nd 3+ + 3C 2 O 4 2- + 10H 2 O(l). to be -30.82 ± 0.30 (2σ), and - 31.14 ± 0.35 (2σ), respectively. These values for can be directly applied to Pu 2 (C 2 O 4 ) 3 •10H 2 O, Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The model established for actinide oxalates by this study provides the needed knowledge with regard to solubilities of actinide/REE oxalates at various ionic strengths, and is expected to find applications in many fields, including the geological disposal of nuclear waste and the mobility of REE under the surface conditions, as Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O can be regarded as the pure Pr and Nd end-members of deveroite, a recently discovered natural REE oxalate with the following stoichiometry, (Ce 1.01 Nd 0.33 La 0.32 Pr 0.11 Y 0.11 Sm 0.01 Pb 0.04 U 0.03 Th 0.01 Ca 0.04 ) 2.01 (C 2 O 4 ) 2.99 •9.99H 2 O. Regarding its importance in the geological disposal of nuclear waste, Am 2 (C 2 O 4 ) 3 •10H 2 O/Pu 2 (C 2 O 4 ) 3 •10H 2 O/Cm 2 (C 2 O 4 ) 3 •10H 2 O can be the source-term phase for actinides, as demonstrated by the instance in the disposal in clay/shale formations. This is exemplified by the stability of Am 2 (C 2 O 4 ) 3 •10H 2 O in comparison with Am(OH) 3 (am), Am(OH) 3 (s) and AmCO 3 (OH)(s) under the relevant geological repository conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Panel Session 14: US DOE WIPP - Future Plans for the Utilization of a National Asset

This panel focused on the next phase of restoring the Waste Isolation Pilot Plant (WIPP: located in Carlsbad, New Mexico, USA) to full disposal operations. The US DOE Carlsbad Field Office (CBFO) is funded for two Capital Asset Projects that will facilitate the restoration of WIPP to full capability and allow for future repository space planning. The panel will include updates on the capital projects, infrastructure improvements to extend WIPP's operational life, regulatory actions in support of capital projects, and improvements in new processes to implement WAC and DSA enhancements. This panel is filled by CBFO and WIPP contractor leaders to provide the big picture leading to WIPP's future. Presentations proposed include: 1 - 2018 WIPP Year in Review (Bruce Covert, Nuclear Waste Partnership, LLC (USA)); 2 - TRU Waste Shipments and WIPP Operations (Mark Pearcy, Nuclear Waste Partnership, LLC (USA)); 3 - Projects and Activities at WIPP (Ronald Gill, US DOE (USA)); 4 - The Future of WIPP (J.R. Stroble, US DOE (USA), Todd Shrader, US DOE (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The Proposed Operations Master Plan of Cigeo - 17013

The proposed operations master plan of Cigeo describes the so-called 'reference progression' of the Cigeo project as envisaged by Andra at the end of the preliminary design studies. It aims at clarifying the objectives of the industrial pilot phase and present the choices offered by the reversibility for the management of the Cigeo project. The reference progression includes the inventory of the waste to be disposed of in Cigeo used for the dimensioning (73 600 m{sup 3} of ILW-LL and 10 100 m{sup 3} of HLW). Cigeo is designed to store the waste already produced and those that will be produced in the future by the existing nuclear installations and those whose creation was authorized, until the predictable end of their operation and their dismantling. All spent fuel is assumed to be reprocessed. The reference progression also includes the duration and the projected milestones of the construction and operation phases of Cigeo until termination. Across the operating life of Cigeo, the extensions of the Cigeo repository areas proceed by successive incremental phases, from the start of the industrial pilot phase (2025) to the final closure operations planned by 2150. Reversibility is defined as the ability to offer to the next generations different choices in terms of radioactive waste long-term management. The implementation of the reversibility principle is based on both governance and technical project management tools. (authors)

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

Radiological Characterisation of VLLW Produced at High-Energy Particle Accelerators - 19118

In this paper we describe the characterisation method of very-low-level radioactive waste (VLLW) produced at the European Organization for Nuclear Research (CERN). The method is based on the calculation of a preliminary radionuclide inventory using either analytical or Monte Carlo codes, on the measurement of γ-emitters and on the quantification of pure-α, pure-β and low-energy X-ray emitters, called difficult-to-measure (DTM) radionuclides, using the so-called scaling factor (SF) technique. The first stage of the characterisation process is the calculation of the radionuclide inventory, which is a list of radionuclides that can potentially be produced via the interaction of primary and secondary particles with the surrounding media. Once the preliminary radionuclide inventory is obtained, the γ-emitting radionuclides are measured via γ-ray spectrometry on each package of the waste population. The major γemitter, called key nuclide (KN), is also identified. The scaling factor method estimates the activity of DTM radionuclides by checking for a consistent and repeated relationship between the key nuclide and the activity of the difficult to measure radionuclides from samples. Between 2016 and 2018, several hundred samples were collected at CERN and over 1000 radiochemical analyses were performed to estimate robust statistics to quantify scaling factors for the most common DTM radionuclides, such as H-3, Fe-55 and Ni-63. This paper illustrates the sample collection strategy, the tests performed to choose the most appropriate statistical models describing the behaviour of scaling factors and a complete quantification of the uncertainties associated with the characterisation process. Finally, we present the results obtained during the characterisation of over 2000 m{sup 3} radioactive metals and cables disposed of at the French National repository for VLLW. (authors)

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

Comparative analysis of uranium bioassociation with halophilic bacteria and archaea

Rock salt represents a potential host rock formation for the final disposal of radioactive waste. The interactions between indigenous microorganisms and radionuclides, e.g. uranium, need to be investigated to better predict the influence of microorganisms on the safety assessment of the repository. Hence, the association process of uranium with two microorganisms isolated from rock salt was comparatively studied. Brachybacterium sp. G1, which was isolated from the German salt dome Gorleben, and Halobacterium noricense DSM15987T , were selected as examples of a moderately halophilic bacterium and an extremely halophilic archaeon, respectively. The microorganisms exhibited completely different association behaviors with uranium. While a pure biosorption process took place with Brachybacterium sp. G1 cells, a multistage association process occurred with the archaeon. In addition to batch experiments, in situ attenuated total reflection Fourier-transform infrared spectroscopy was applied to characterize the U(VI) interaction process. Biosorption was identified as the dominating process for Brachybacterium sp. G1 with this method. Carboxylic functionalities are the dominant interacting groups for the bacterium, whereas phosphoryl groups are also involved in U(VI) association by the archaeon H. noricense.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

DOE SFWST Campaign R&D Roadmap Update Rev.1.

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Departmentof Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel and Waste Disposition(SFWD) is conducting research and development (R&D) on deep geologic disposal of spentnuclear fuel (SNF) and high-level nuclear waste (HLW). R&D addressing the disposal ofSNF/HLW in the U.S. is currently generic (i.e., "non-site-specific") in scope, following thesuspension of the Yucca Mountain Repository Project in 2010. However, to prepare for theeventuality of a repository siting process, the former Used Fuel Disposition Campaign (UFDC) ofDOE-NE, which was succeeded by the SFWST Campaign, formulated an R&D Roadmap in 2012outlining generic R&D activities and their priorities appropriate for developing safety cases andassociated performance assessment (PA) models for generic deep geologic repositories in severalpotential host-rock environments in the contiguous United States. This 2012 UFDC Roadmap alsoidentified the importance of re-evaluating priorities in future years as knowledge is gained fromthe DOE's ongoing R&D activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

The iodine-129 paradox in nuclear waste management strategies

Nuclear energy has an important role in the low-carbon energy transition, but the safety of spent nuclear fuel (SNF) management remains a public concern. Here, in this study, we investigate the interplay between waste management strategies and their environmental impacts with a particular focus on a highly mobile and persistent radionuclide, iodine-129 (I-129), which is the dominant risk contributor from SNF disposal and at existing groundwater contamination sites. The results show that the current recycling practice releases more than 90% of I-129 in SNF into the present-day biosphere using an isotropic dilution strategy, whereas the direct disposal of SNF in geological repositories is likely to delay and reduce the release by 8 orders of magnitude. In addition, our data synthesis of surface water concentrations near four nuclear facilities shows that the release-dilution strategy results in lower concentrations than regulatory standards, while insufficient waste isolation in the past has resulted in locally high concentrations within one site. Our analysis suggests that it is essential to consider effluents more explicitly as a part of the waste, that as society moves from dilution to isolation of waste, the potential risks of waste isolation to local regions should be carefully evaluated, and that excessive burdens of proof could hinder or discourage waste isolation. Comprehensive waste management strategies—considering not just volume but also mobility, isolation technologies and ultimate fates—are needed for persistent contaminants. This study offers valuable insights for optimizing the management of SNF and other persistent contaminants.

Wainwright, Haruko M. [Massachusetts Inst. of Tech↗
DOE OSTI2023

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Analysis of the Needs for Long Term Memory and Knowledge Preservation Relating to Radioactive Waste Disposal Facilities - 17213

Proposals for a set of provisions for long-term memory and knowledge to be defined for a radioactive waste disposal facility, near surface or deep underground, address two primary motives, related to two ethical principles. The first motive is to prevent future generations from interfering involuntarily with the repository. This requires maintaining awareness of the repository, and addresses the ethical principle of protection of man and environment. The second motive is to provide future generations all the available relevant information which might help them make informed decisions about intentional actions, and assess the consequences. This requires transmitting detailed knowledge of the repository, and addresses the ethical principle of preservation of freedom of action. The set of provisions to be implemented with respect to each of these motives may not be the same. In order to define and assess the set of provisions, it is also useful to identify the various components of the process of transmission of a given message, or set of messages, to future generations. Three sub-processes have been identified: (i) 'memorization', at the producer stage, where a full set of information to be transmitted is identified, organized and expressed; (ii) 'preservation', where the potential durability of records is extended, the preservation conditions are controlled and where the records may be restored, if their status is degraded; (iii) 'access', at the receiver stage, where the receiver has to be notified of the existence of the information, to find it and to interpret it properly. As a failure of transmission to future generations would result from the failure of any of the subprocesses, a minimal set of provisions may be defined from this decomposition. (authors)

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

Effects of Borate and Organics on U(VI) Solubility in WIPP Brine

The solubility of uranium (VI) in Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities. This research was performed by the Los Alamos National Laboratory Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP). The WIPP Actinide Source Term Program (ASTP) did not develop a model for the solubility of actinides in the VI oxidation state. The solubility of UO 2 2+ , in the absence of WIPP specific data, is presently set to be equal to a conservatively high 1 mM within the WIPP Performance Assessment (PA) for all expected WIPP conditions (SOTERM, 2019) as selected at the recommendation of the Environment Protection Agency (EPA) (EPA, 2005). According to the current WIPP chemistry model assumptions and conditions, the expected pC H+ is about 9.5 and controlled by MgO buffering CO 3 2- . The goal of this study is to perform screening experiments that account for the contributions of organics and borate on uranium solubility. In this report, the solubility of U(VI) was determined at pC H+ 9 WIPP brine in the absence or presence of borate and organics at under-saturation approach. Experiments were equilibrated for about 135 days. Organic compounds present in WIPP waste can form strong complexes with actinides and can affect the oxidation states of actinides. The organic compounds addressed in WIPP performance assessment include EDTA (Ethylenediaminetetraacetic Acid), oxalate, citrate, and acetate (SOTERM, 2019). These data quantify the effects of WIPP-relevant concentrations of borate and organics effects on the solubility of U(VI) to challenge the predictions of the WIPP actinide model and inform decisions and recommendations made in the upcoming recertification of the WIPP (CRA-2024). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Experimental Strategy to Challenge Actinide Solubility Predictions” and designated LCO-ACP-26. All data reported were obtained under the LANL-CO Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD) (QAPD, 2017).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗