Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear waste form”

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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
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 OSTI2024

New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗
DOE OSTI2020

Brine Availability Test in Salt, a Heated Borehole Experiment at the Waste Isolation Pilot Plant, New Mexico, USA - 20233

The Brine Availability Tests in Salt is a US Department of Energy Office of Nuclear Energy supported activity as part of the Spent Fuel and Waste Disposition Campaign where we are performing simultaneous heated and control experiments in adjacent horizontal boreholes in the WIPP Salt Disposal Investigation drifts as part of experimental Phase 1. Phase 1a of BATS, a low-cost, small spatial-scale experiment performed in order to design instrumentation and methods for further experimentation in the URL, has been completed (June 2018 - May 2019) and is described elsewhere. In coincidence with multiple physical parameters that will be measured over the duration of the experiment, we will be measuring the liquid and vapor phases of H{sub 2}O as well as its isotopic content (not measured in Phase 1a). Three different forms of water occur in natural salts: as fluid in macro scale (mm size) inclusions, as intergranular free H{sub 2}O and as H{sub 2}O incorporated in hydrous minerals. Each of these forms has potentially different stable isotope signatures and will be influenced differentially by heat induced transport and by interaction with introduced tracers. Continuous measurement of vapor phase H{sub 2}O isotopes during borehole heating will be impacted by several effects: 1) evaporation as liquid brine flows into the borehole and evaporates completely or partially into the dry N{sub 2} stream; 2) steam from dehydration of hydrous minerals, that may enter the borehole; and 3) hydrous minerals that may form in the borehole (preferentially incorporating heavier isotopes). The continuous time series collected over the course of the heating test will be combined with analyses of pre-test and post-test collected samples of test site salt to investigate the contributions of each of the three water types as H{sub 2}O is mobilized due to the effects of heating. (authors)

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

Model Package Report: Composite Analysis Solid Waste Release Model (CASWR Model)

This document describes the implementation of a solid waste form release model in GoldSim for the Hanford Site Composite Analysis (CA) Update. This Composite Analysis Solid Waste Release model (CASWR model) was designed to generate deterministic radionuclide release rates for Hanford’s Central Plateau solid waste disposal sites using single realizations of release model coefficients. Five generalized waste form types are used for the conceptual model of waste release: surplus reactor block, cement, soil-debris, grouted residual waste, and ancillary equipment. The surplus reactor block waste form consists of radionuclide leaching from surplus production graphite reactor core blocks via unspecified processes (White et al., 1984 as cited in PNNL-15965). The cement waste form represents solidified wastes whose permeability is much lower than that of the surrounding soil. The soil-debris waste form type is defined as unconsolidated waste mixed with soil material. Tanks and canyon complexes comprise the grouted residual waste form such that their solid waste will be grouted and capped with a surface barrier at the completion of their cleanup. Finally, the ancillary equipment waste form constitutes contaminant releases from ancillary and auxiliary waste form residues associated with tank farms at closure. Individual sub-models are implemented to numerically represent a respective waste form within the CASWR Model: Surplus Reactor Block Sub-model, Cement Sub-model, SoilDebris Sub-model, Grouted Residual Waste Sub-model, and Ancillary Equipment Submodel. Advection is assumed to be the primary transport process governing the release of radionuclides in the Ancillary Equipment and Soil-Debris Sub-models. Diffusion is assumed to be the primary release process in the Grouted Residual Waste and Cement Sub-models. An unspecified zero-order release process is considered in the Surplus Reactor Block Sub-model due to the lack of information regarding actual processes involved in irradiated graphite leaching. The Surplus Reactor Block, Cement, and SoilDebris Sub-models were compared against analytical solutions (PNNL-11800, Composite Analysis for Low-Level Waste Disposal in the 200 Area Plateau of the Hanford Site). The agreement between results of these analytical solutions and the corresponding waste form models verified their correct implementation in GoldSim. A 1-D modeling abstraction approach for the Grouted Residual Waste and Ancillary Equipment Sub-models was adopted from existing Performance Assessment (PA) models. Despite the simplifications made in these sub-models, they were found to be appropriate representations of the waste forms, similar to what was used in the Waste Management Area C PA model (RPP-ENV-58782, Performance Assessment of Waste Management Area C, Hanford Site, Washington, Rev. 0). A sensitivity analysis was conducted to identify the most influential parameters in each waste form sub-model. Suggestions for considering pH-dependent and redox-dependent release mechanisms are formulated through the development of a conditional constant approach in GoldSim.

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

Effects of He Ion Irradiation in Simple Borosilicate Glasses for HLW Immobilisation - 20324

High-level radioactive waste (HLW) consists of fission products and minor actinides. Borosilicate glass is the host material of choice in most countries to safely immobilize HLW, followed by eventual geological disposal. HLW glasses will experience both self-heating and self-irradiation effects due to the decay of fission products and minor actinides present therein. During the first 500 years of geological disposal, beta-gamma decay will be the main source of radiation due to short half-lives of fission products, as compared to very long-lived actinides which will be responsible for irradiation over much longer timescales. High-energy alpha particles and alpha recoil nuclei can potentially have significant impact (stored energy, volume change, accumulated dose, chemical durability) on the performance of the wasteform over longer timescales. Thus, to ensure safe and economic disposal over geological periods (10{sup 4} - 10{sup 6} years), the assessment of performance and properties in terms of radiation effects becomes indispensable. This work aims to simulate the effects of alpha decay in radioactive waste glasses. Two international glasses, namely UK 'MW' lithium-sodium borosilicate and Indian sodium-barium borosilicate, which are being used as HLW immobilization matrices by the UK and India respectively, were prepared by a melt-quench method and annealed below their glass transition temperatures. Both glasses were irradiated with 650 KeV He{sup 2+} ions with doses ranging from 10{sup 15} to 10{sup 17} ions/cm{sup 2}. Structural analysis was carried out using electron paramagnetic resonance (EPR) and Raman spectroscopies. First derivative continuous wave EPR spectra were recorded for all pristine and irradiated glass monoliths at room temperature using an X-band (frequency 9.6 GHz) EPR Bruker spectrometer. Results from EPR analyses showed the formation of irradiation induced paramagnetic centers at g ∼ 2.00 in all irradiated glasses. Raman spectra of pristine and irradiated samples were acquired, and showed notable changes in glass structure post-irradiation, with the emergence of a broad Raman band at 1560 cm{sup -1} upon irradiation with high fluences. (authors)

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

Design and Application of Materials for Sequestration and Immobilization of 99 Tc

99 Technetium ( 99 Tc) is a hazardous radionuclide generated by the nuclear industry that poses a serious environmental threat. The wide variation and complex chemistries of nuclear waste streams containing 99 Tc often create unique, site specific challenges when sequestering and immobilizing the waste in a matrix suitable for long-term storage and disposal. Therefore, an effective management plan for 99 Tc containing radioactive waste will likely require a variety of suitable materials/matrices capable of adapting to and addressing these challenges. In this review, we discuss and highlight the key developments for effective removal and immobilization of 99 Tc in inorganic waste forms. Specifically, we review the synthesis, characterization, and application of materials for targeted removal of 99 Tc from (simulated) waste solutions under various experimental conditions. These materials include: i) layered double hydroxides (LDHs), ii) metal-organic frameworks (MOFs), iii) ion-exchange resins (IERs) as well as cationic organic polymers (COPs), and iv) surface modified natural clay materials (SMCMs). Secondly, we discuss some of the major and recent developments towards 99 Tc immobilization in i) glass, ii) cement, and iii) iron mineral waste forms. Finally, we present future challenges that need to be addressed for the design, synthesis, and selection of suitable matrices for the efficient sequestration and immobilization of 99 Tc from targeted wastes. Finally, the purpose of this review is to inspire the researches on the design and application of various suitable materials/matrices for selective removal of 99 Tc present globally in different radioactive wastes and its immobilization in stable/durable waste forms.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Fundamental Radiation Chemistry

An overview presentation for the Fundamental Radiation Chemistry component of the Material Recovery and Waste Form Development campaign to be delivered to the campaign team. The overview includes all major findings from FY19 research projects, current state of FY20 research projects, additional projects completed between FY19 and now, and finally our vision for the future direction of the Fundamental Radiation Chemistry component.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2025

Stabilization of Preternatural Barium Oxidation States as an Unexpected Byproduct of β-Decay: Discovery of a New Halide Semiconductor Alloy

137 Cs has a wide range of roles in the nuclear industry. The solid material, safely encapsulated in CsCl as 137 CsCl, is stored as fission product waste from nuclear power production and legacy waste from nuclear weapons production; it has also served as a radiation source in food and sewage irradiators as well as medical devices. However, because of the solubility of the chloride salt and the relatively high specific activity of 137 Cs, damaged or broken capsules can lead to severe radiological accidents. Safe capsule design and material recycling are complicated by the unclear structural evolution during β-decay, which remains ambiguous due to the differing oxidation states of Cs (1+) and Ba (2+). Here, in this study, we use first-principles calculations to investigate the evolving structure–property relationships of Cs 1–x Ba x Cl during β-decay. Despite the well-established 2+ formal oxidation state of alkali-earth metals, we find that Ba 1+ can be stabilized in the form of a mixed-valence alloy at low concentrations. Specifically, we identify three regimes for the β-decay of 137 Cs into CsCl: Ba-doped CsCl (Ba ≤ 14%), wherein Ba has the expected 2+ oxidation state; Cs–Ba–Cl alloys, where Ba has a mix of the usual Ba 2+ and highly unusual Ba 1+ oxidation state in the form of a quasi-disordered mixed-valence alloy (Ba = 25%); and phase separation into a CsCl + BaCl 2 + Ba (m) mechanical mixture, where Ba reverts to its expected 2+ oxidation state (Ba > 25%). Surprisingly, the Cs 0.75 Ba 0.25 Cl mixed-valence alloy is a narrow indirect band gap semiconductor (1.05 eV) despite the insulating nature of both CsCl and BaCl 2 . It also exhibits strongly excitonic polarized optical properties, has glass-like ultralow thermal conductivity (directional average of 0.21 W/mK at 300 K), and shows greater resistance to deformation under both tensile and volumetric strengths compared with the original CsCl structure (e.g., shear and Young’s modulus of 9.04 and 31.62 GPa, respectively). These findings imply that transmutation of 137 Cs leads to highly unusual chemical bonding that stabilizes Ba 1+ in local regions of the quasi-disordered Cs 0.25 Ba 0.75 Cl, resulting in anomalous physical properties. Moreover, this discovery provides valuable insight for safe nuclear waste capsule design, which can aid in preventing environmental or human exposure to radioactive materials.

Fuhr, Addis S. [Oak Ridge National Laboratory (ORN↗
DOE OSTI2024

Finding ways to reduce nuclear waste: searching for the unknown one step at a time

In my home country, Venezuela, research has been stagnant. Due to the political turmoil and the crisis, many educated people have left the country in search for a better life. This has caused a deficit in any technological and scientific advances, making Venezuela one of the first South American countries to have its rate of publications decline by 29% in 2013. Currently, Venezuela lacks the infrastructure and the means to keep up with the research progress as compared to other countries in South America, such as Brazil. Since coming to the United States (US), and currently working for a national laboratory, the active research environment endorses a wide range of careers and engineering programs that allow researchers to thrive at any given field. Researchers have access to funds and tools to succeed in developing materials for the future. There are 17 national laboratories in the US, and all of these have a different research focus/objective. As examples, Los Alamos National Laboratory and Sandia National Laboratory focus is on national homeland security, weapon science, radiation effects, among others. Argonne National Laboratory focuses on nuclear energy, energy storage, high performance computing, etc. At Idaho National Laboratory (INL) the research focuses on innovating nuclear energy and clean energy resources, critical infrastructure materials, along with fuel cycle solutions to manage, dispose and find ways to recycle current and future radiological waste. Compared to other national laboratories, INL focuses slightly more on applied processes and how nuclear energy can be innovated to next reactor design and technologies. The research being conducted at INL made me apply for a Seaborg distinguished postdoctoral position. For the position itself, the researcher must submit a proposal related to actinide chemistry on a research field area. In this position, 50% of my time will be focused on my own proposal. The proposal that I am working on is focused on the innovation of nuclear energy and fuel cycle recycling, which is why I was mainly interested on working at this national laboratory. To give a bit more context of what my proposal is about, a little bit of background is necessary: After the nuclear fuel (UO2) is used in a reactor, the fuel matrix is then characterized by various fission products (FP). Among these FP (including rare earth elements, alkali/alkaline earths, and actinides), many can potentially be recovered through nuclear reprocessing technologies. In pyroprocessing, the used nuclear fuel undergoes electrochemical dissolution into a molten chloride salt mixture in an electrorefiner. Initially, uranium is reduced onto an inert cathode by applied potentials. However, numerous remaining FPs accumulate in the melt and pose challenges for recovery by an inert electrode, particularly the rare earth elements (e.g., Nd, Gd, Pr, Sm) due to their multivalent oxidation states and tendencies toward side reactions, leading to their dissolution in the electrolyte. These recovery challenges result in inefficiencies and necessitate the continual discarding of the molten chloride salt, thereby generating additional waste. Furthermore, the presence of rare earth elements and other fission products in the molten salt electrolyte alters its physical and chemical properties, affecting both uranium recovery efficiency and the longevity of the molten chloride salt. To improve the recovery efficiency of the FP, specifically rare earth elements, I am investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form. The kinetic pathways and the chemical reactions of these elements will give insights on how the recovery efficiency can be improved. The interactions and speciation of these elements are being studied by spectro-electrochemistry at high temperature environments in quartz and other ceramic materials (e.g., alumina crucibles). Some of the challenges I am facing specifically relates the reactivity of some of these elements with different glass and crucible materials. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for an efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Examining Thermolytic Production of Hydrogen from Lubrication Oil

Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to conduct testing via Technical Task Request (HR) to determine the thermolytic HGR of Mobil SHC™ 630, a lubrication oil. Currently, 35 gallons of contaminated Mobil SHC 630 is proposed for release into the recycle stream from the Defense Waste Processing Facility (DWPF) to Tank 22 and then to the 242-16H (2H) Evaporator system. Inhibited recycle waste in the Recycle Collection Tank (RCT) is transferred to Recycle Pump Tank (RPT) in the Low Point Pump Pit (LPPP) and then to the Concentration, Storage and Transfer Facilities (CSTF) H-area. The lubrication oil would be added directly to the RPT, bypassing the RCT. The current DWPF waste compliance plan for liquid transfers from the RCT to the CSTF limits the concentration of Mobil SHC 630 to <1,100 ppm which is equal to <9.3 gallons of Mobil SHC 630 when considering a 7,500 gal RCT batch with an initial Mobil SHC 630 concentration of 42 mg/L. Mobil SHC 630 is expected to be largely immiscible in the caustic aqueous waste stream. It is a blend of base oils including polyalphaolefin (PAO) base oil and additives such as triphenylphosphate and cresyl diphenyl phosphate at various concentrations (<0.25 wt%). While the base oils are expected to be largely unreactive in CSTF waste, the triarylphosphates additives would be expected to hydrolyze in the caustic waste, forming diarylphosphates and phenol. The tests described herein were governed by a single Run Plan and will determine thermolytic HGR from the caustic aqueous solution, as well as from any organic phases present.

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

Glass Crystallization Constraints for WTP LAW Operations: Assessment of Isothermal Treatments on Crystal Formation

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 melter operation while processing advanced low-activity glass waste forms. The constraints target SnO 2 and ZrO 2 crystals that can form when melter temperatures drop below 1100 °C. These types of crystals were found to be potentially detrimental during processing as they are denser than low-activity waste glass melts. SnO 2 , density of 6.95 g/cm3, and ZrO 2 , density of 5.68 g/cm3, have the potential to form during melter idling and settle to the bottom of the less dense glass melt (approximate density 2.65 g/cm3). If the crystals are present in appreciable amounts, they can result in blockages of the pour-spout riser, which impacts glass pouring and melter operation. Using previously acquired results and results from testing during this effort, constraints were determined and are suggested as options to reduce the risk of forming crystals of the types and concentrations that are likely detrimental to melter operation.

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

Saltstone First Quarter Calendar Year 2021 (1QCY21) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled quarterly for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z00023, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the first quarter of calendar year 2021(1QCY21). One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample”. Results from this technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Remediation (SRR). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The 1QCY21 saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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

Saltstone Third Quarter Calendar Year 2021 (3QCY21) Toxicity Characteristic Leading Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled quarterly for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00023, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the third quarter of calendar year 2021(3QCY21). One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample”. Results from this technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Remediation (SRR). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The 3QCY21 saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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

January 2022 Semiannual Saltstone Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00025, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the January 2022 semiannual Toxicity Characteristic Leaching Procedure (TCLP) samples. One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample.” Results from this technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The January 2022 semiannual saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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

July 2022 Semiannual Saltstone Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Salstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00025, Revision 0, two SDF waste form (salstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the July 2022 Semiannual Toxicity Characteristic Leaching Procedure (TCLP) samples. One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the "Cement-Free grout sample." Results from this technical report support Task 2: 'Grout Leaching Analyses' of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using Toxicity Characteristic Leaching Procedure (TCLP). The July 2022 semiannual salstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for the non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

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

Evolution of Radicals from the Photolysis of High Ionic Strength Alkaline Nitrite Solutions

Reactive nitrogen species (RNS), along with reactive oxygen species (ROS), are significant products from radiolysis in solution. While much research has been focused on biological systems, these species are also important products in the autoradiolysis that occurs in nuclear waste. In this work, we determine the correlation between solution constituents, particularly nitrite, and radical products in highly alkaline solutions relevant to liquid waste. Because these radicals tend to be very short-lived, we employ spin trapping in conjunction with electron paramagnetic resonance (EPR) to detect them and quantify their production. Most spin traps do not function in these conditions (>1 M NaOH); however, nitroalkanes such as nitromethane will act as spin traps in their aci form, which is dominant at high pH. To restrict the products to those originating from nitrite, we use 280–480 nm UV light to generate radicals, avoiding products from the photolysis of water. Under these circumstances, nitric oxide, nitrite radicals, and hydroxyl radicals are detected, and the trends with the concentration of the constituents of the solutions are tracked. These include nitrite, nitrate, hydroxide, and carbonate. We find that, while the equilibrium shifts with increasing pH from hydroxyl radicals to the more slowly reacting oxide radicals, the production of nitrite radicals does not decrease.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗