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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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

A Review of Global Radioxenon Background Research and Issues

Among the most important problems for the world-wide nuclear explosion monitoring is the interference of naturally occurring and man-made radionuclides. The International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) frequently detects these interferences using sensitive radionuclide measurement equipment. We commonly refer to the presence of radionuclides that are relevant to the CTBT but do not originate from a nuclear explosion as “background”. Backgrounds are highest near the sources but are known to have regional and global effects on the IMS. This review paper summarizes much of the relevant work in the area of background and discusses issues of interest for nuclear explosion detection.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Chemical and structural characterization of particulate fallout isolated from air-filters

We report particulate nuclear fallout is the radioactive byproduct of a nuclear event formed by the mixture of proximate environmental materials with vaporized bomb debris. The fallout debris can be transported into the atmosphere during cloud rise, raining out locally and dispersing globally constituting a radiation hazard and contributing to the distribution of anthropogenic radionuclides in the environment. Questions remain on how entrainment of environmental material in the fireball affects fallout formation processes and radionuclide incorporation and distribution during cooling. To inform the characterization of fallout including the development of fallout size distributions and how radionuclides are incorporated into fallout debris where entrainment plays a role, we analyzed an archived historic US air-filter collected by aircraft in the aftermath of ground interacting nuclear tests. Particulate fallout collected on the filter was isolated and analyzed using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and NanoSIMS (Secondary Ion Mass Spectrometry) to determine chemistry, structure, morphology, and size of the particles. Results demonstrate that the particles isolated from the filter have spherical shape, display complex internal structures, and are mainly composed of Fe and Si oxides. In these spherical particles, Pu is preferentially associated with Fe-rich composition. The characterization of fallout particulate samples can provide information on nucleation and particle growth from the vapor phase to improve modeling and simulation of fallout hazards.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Current Status of Radiopharmaceutical Therapy

In radiopharmaceutical therapy (RPT), a radionuclide is systemically or locally delivered with the goal of targeting and delivering radiation to cancer cells while minimizing radiation exposure to untargeted cells. Examples of current RPTs include thyroid ablation with the administration of {sup 131}I, treatment of liver cancer with {sup 90}Y microspheres, the treatment of bony metastases with {sup 223}Ra, and the treatment of neuroendocrine tumors with {sup 177}Lu-DOTATATE. New RPTs are being developed where radionuclides are incorporated into systemic targeted therapies. To assure that RPT is appropriately implemented, advances in targeting need to be matched with advances in quantitative imaging and dosimetry methods. Currently, radiopharmaceutical therapy is administered by intravenous or locoregional injection, and the treatment planning has typically been implemented like chemotherapy, where the activity administered is either fixed or based on a patient’s body weight or body surface area. RPT pharmacokinetics are measurable by quantitative imaging and are known to vary across patients, both in tumors and normal tissues. Therefore, fixed or weight-based activity prescriptions are not currently optimized to deliver a cytotoxic dose to targets while remaining within the tolerance dose of organs at risk. Methods that provide dose estimates to individual patients rather than to reference geometries are needed to assess and adjust the injected RPT dose. Accurate doses to targets and organs at risk will benefit the individual patients and decrease uncertainties in clinical trials. Imaging can be used to measure activity distribution in vivo, and this information can be used to determine patient-specific treatment plans where the dose to the targets and organs at risk can be calculated. The development and adoption of imaging-based dosimetry methods is particularly beneficial in early clinical trials. In this work we discuss dosimetric accuracy needs in modern radiation oncology, uncertainties in the dosimetry in RPT, and best approaches for imaging and dosimetry of internal radionuclide therapy.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2022

Ultra-low-level measurements of airborne fission products from the Fukushima Daiichi reactor accident using high volume collection systems at Savannah River National Laboratory

For this study, ultra-low-level measurements of radionuclides in air have been conducted at the Savannah River National Laboratory (SRNL) to determine the atmospheric concentration of fission products released following the Fukushima Daiichi reactor accident on March 11, 2011. Air filter samples were acquired from two high-volume collection systems (a traditional filter-based system and an electrostatic precipitator-based system) to monitor airborne radionuclide concentrations in the period covering from 2 weeks to 3 years after the disaster. The world-wide spread of low-level concentrations of airborne fission products from the Fukushima event provided a unique opportunity to demonstrate SRNL's electrostatic particle collection technology and other improvements in environmental monitoring developed at the Savannah River Site (SRS). Detecting and analyzing the release allowed a comprehensive test of SRS systems for monitoring environmental radioactivity. Gamma-ray-emitting fission products ( 131,132 I, 134,136,137 Cs, and 129,132 Te) and cosmogenic isotopes ( 7 Be and 22 Na) in air were detected and quantified by high-resolution gamma-ray spectroscopy at concentrations as low as 0.07 μBq per standard cubic meter (SCM) (50 mBq total 137 Cs), while plutonium content was quantified by thermal ionization mass spectrometry (TIMS) at concentrations as low as 6.5 × 10 –21 g/SCM (3.0 fg 239+240 Pu). Isotope concentrations measured at SRNL from gamma-ray spectroscopy were compared to independent measurements from Chapel Hill, NC, located approximately 370 km (230 mi) NE of SRNL. Meteorological modeling was also used to predict radionuclide transport from the location of release to both measurement locations.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2025

A statistical approach to screening isotopic signatures in monitoring for underground nuclear explosions

The ability to differentiate between atmospheric radionuclide signatures from underground nuclear explosions (UNEs) and signals from other sources, such as medical isotope-production facilities and nuclear reactors, can be critical to the detection and monitoring of unannounced, low-yield nuclear events. Signatures having anomalously high amplitudes, compared to background levels, remain the best indicator in screening for a UNE. However, isotopic composition can further validate a suspected UNE signature, but separation from any atmospheric background composition is first necessary. To date, evaluating the challenges of performing this separation has typically involved comparing an observed background with a highly idealized deterministic model of radioxenon signature production by a UNE that does not consider the influence of post-detonation chemical/physical processes in the detonation cavity or the subsequent gas transport mechanisms that can also affect the isotopic composition of the detected gas signature. In addition, purely deterministic models, as previously employed, overlook the uncertainty inherent in estimating critical parameters characterizing the UNE and its detonation environment. In this paper, we create detailed, multi-parameter models of radionuclide evolution using the widely accepted England and Rider post-detonation radionuclide decay-chain network coupled to detailed models simulating physical production and transport processes affecting the gas signature. Because these models are governed by uncertain parameters including barometric fluctuations, realistic ranges of variation for each of the parameters influencing isotopic composition are then defined. A Latin-Hypercube sampling approach is used to obtain a random distribution of isotopic production and gas transport results associated with a given value of each parameter. We apply these results to background histories of two stations, one providing 4-isotope background measurements and the other providing two-isotope measurements associated with the 2013 DPRK announced UNE.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

A laboratory-scale process for producing dilithium beryllium tetrafluoride (FLiBe) with dissolved uranium tetrafluoride

Flibe Energy, Incorporated (FEI)'s conceptual Lithium Fluoride Thorium Reactor (LFTR) incorporates a chemical processing facility aimed at recovering uranium and other valuable volatile radionuclides while managing harmful radionuclides from the used fuel. The fuel utilized in this reactor is a combination of dilithium beryllium tetrafluoride (Li 2 BeF 4 or FLiBe) and uranium tetrafluoride (UF 4 ), (FLiBe/U). FEI's plan involves extracting the uranium and other valuable volatile fluoride-forming radionuclides using nitrogen trifluoride (NF 3 ). To facilitate laboratory-scale testing of uranium extraction using NF 3 and address the toxicity and physical hazards associated with beryllium and beryllium fluoride (BeF 2 ), we used a two-step process to prepare the simulated fuel salt. The first step entailed thermally decomposing ammonium beryllium tetrafluoride [(NH 4 ) 2 BeF 4 ] (ABeF) through a nominal 3-step process, combined with appropriate amounts of lithium fluoride (LiF) and UF 4 , resulting in the formation of beryllium fluoride (BeF 2 ). In the second step, the mixture was repeatedly melted and frozen at the melting point of FLiBe to prepare the eutectic FLiBe with dissolved UF 4 . Although the concept appears straightforward, the production of FLiBe/U involved various challenges. These challenges included transporting the gaseous decomposition products of ABeF, hydrogen fluoride (HF) and ammonia (NH 3 ), while preventing the formation of ammonium fluoride (NH 4 F). Additionally, it was necessary to control the reaction between the higher-than-anticipated water content in the commercial ABeF with NH 3 , HF, and the condensed NH 4 F, protect UF 4 from forming an unknown black compound, select suitable structural materials to mitigate fluoride corrosion, address the risks associated with beryllium toxicity through equipment design and operational protocols, and monitor process conditions. This article provides an account of the thermal decomposition chemistry observed in the commercial ABeF, describes the FLiBe/U production apparatus, describes the experiences and process refinements developed to prepare FLiBe/U, and presents our characterizations of prepared FLiBe/U.

Ammonium beryllium fluoride thermal decomposition↗
DOE OSTI2021

New capability for neutron transmission measurements at LANSCE: The DICER instrument

A new capability for neutron transmission measurements has been developed at the Los Alamos Neutron Science Center (LANSCE). The DICER (Device for Indirect Capture Experiments on Radionuclides) instrument is designed to tightly constrain neutron-capture cross sections for short-lived radionuclides that are beyond the reach of direct techniques, via measurement and analysis of neutron transmission resonances on these same radionuclides. The innovative DICER apparatus enables simultaneous sample-in and sample-out measurements, thus minimizing the amount of sample material needed and reducing the measurements time by half. A description of the current status of the new transmission station, along with a performance overview and results from known reactions, will be presented here.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Trithiol ligand provides tumor-targeting 191 Pt-complexes with high molar activity and promising in vivo properties

The Auger electron-emitting radionuclide 191 Pt is a promising candidate for radiopharmaceutical therapy. Herein, we explored novel labeling methods for 191 Pt using thiol-containing ligands to improve the in vivo stability and targeting ability of 191 Pt-labeled complexes. We synthesized dithiol-containing N 2 S 2 and NS 2 ligands, and a trithiol ligand, and then compared their radiochemical reactivity with 191 Pt. [ 191 Pt]Pt-trithiol was synthesized and its biodistribution was evaluated in mice and compared with free 191 Pt. Finally, a 191 Pt-trithiol complex targeting prostate-specific membrane antigen (PSMA): [ 191 Pt]Pt-trithiol-PSMA was developed and evaluated in mice bearing tumor xenografts and compared with a 191 Pt-complex labeled via monothiol-containing Cys ([ 191 Pt]Pt-Cys-PSMA). A comparison of N 2 S 2 , NS 2 , and trithiol showed that the trithiol ligand is the best for producing 191 Pt-labeled compounds in high yield and as a single peak in preparative HPLC. Notably, the trithiol ligand made 191 Pt-labeled compounds and precursors separatable, achieving 191 Pt-labeled products with a high molar activity: 200–400 mCi/μmol (7.4–14.8 GBq/μmol) at EOS. Additionally, [ 191 Pt]Pt-trithiol and [ 191 Pt]Pt-trithiol-PSMA were stable in vivo with rapid clearance compared with free 191 Pt and [ 191 Pt]Pt-Cys-PSMA. [ 191 Pt]Pt-trithiol-PSMA resulted in a low uptake in most normal organs and a high uptake in the kidneys and prostate cancer with PSMA expression. Furthermore, this study demonstrated that a labeling method with trithiol for Pt radionuclides achieves 191 Pt-labeled products with high molar activity. 191 Pt-trithiol-PSMA showed promising in vivo stability and tumor-targeting specificity, which should facilitate the pharmaceutical development of Pt radionuclides for radiopharmaceutical therapy, especially Auger electron cancer therapy.

Auger emitters↗
DOE OSTI2023

Bioaccumulation of contaminants in Scarabaeidae and Silphidae beetles at sites polluted by coal combustion residuals and radiocesium

Anthropogenic contamination from coal-fired power plants and nuclear reactors is a pervasive issue impacting ecosystems across the globe. As a result, it is critical that studies continue to assess the accumulation and effects of trace elements and radionuclides in a diversity of biota. In particular, bioindicator species are a powerful tool for risk assessment of chemically contaminated habitats. Using inductively coupled plasma mass spectrometry (ICP-MS) and auto-gamma counting, we analyzed trace element and radiocesium contaminant concentrations in Scarabaeidae and Silphidae beetles (Order: Coleoptera), important taxa in decomposition and nutrient cycling, at contaminated and reference sites on the Savannah River Site, South Carolina, U.S. Our results revealed variability in trace element concentrations between Scarabaeidae and Silphidae beetles at uncontaminated and contaminated sites. Compared to Scarabaeidae, Silphidae had higher levels of chromium (Cr), copper (Cu), iron (Fe), manganese (Mn), nickel (Ni), and zinc (Zn). Unexpectedly, concentrations of Cr, Cu, and Ni were higher in both taxa at the uncontaminated sites. Scarabaeidae and Silphidae beetles at the coal combustion waste site consistently had high concentrations of arsenic (As), and Scarabaeidae had high concentrations of selenium (Se). Of the 50 beetles analyzed for radiocesium levels, two had elevated radioactivity concentrations, both of which were from a site contaminated with radionuclides. Furthermore, our results suggest carrion beetles may be particularly sensitive to bioaccumulation of contaminants due to their trophic position and role in decomposition, and thus are useful sentinels of trace element and radionuclide contamination.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2025

Quantitative Encapsulation and Homogeneity Assessment of Sol–Gel Based Nuclear Explosive Debris Simulants

Nuclear explosive debris simulants are an important material in training and validating aspects of post-detonation nuclear forensic processes. Realistic simulants should replicate several aspects of nuclear explosive debris such as the size, shape, color, density, and chemical and radiological properties. Silica particles produced via sol-gel synthesis have recently been found to successfully reproduce many of these parameters including the controllable incorporation of radionuclide content. However, to be useful as a benchmarking material for validation and verification of laboratory methodologies, radionuclide content from batch-to-batch must be reproducible. Here, in this work, we explore the variance in radionuclide distribution incorporated into sol-gel benchmarking materials with respect to sample subdivision. Results will help inform the sample sizes required to minimize variance between samples.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Harvesting krypton isotopes from the off-gas of an irradiated water target to generate 76Br and 77Br

Abstract A flowing-water target was irradiated with a 150 MeV/nucleon beam of 78 Kr at the National Superconducting Cyclotron Laboratory to produce 77 Kr and 76 Kr. Real-time gamma-imaging measurements revealed the mass transport of the krypton radioisotopes through the target-water processing, or “isotope harvesting”, system. The production rates were determined to be 2.7(1) × 10 –4 nuclei of 76 Kr and 1.18(6) × 10 –2 nuclei of 77 Kr formed per incident 78 Kr ion. Utilizing an off-gas processing line as part of the isotope harvesting system, a total of 7.2(1) MBq of 76 Kr and 19.1(6) MBq of 77 Kr were collected in cold traps. Through the decay, the daughter radionuclides 76 Br and 77 Br were generated and removed from the traps with an average efficiency of 77 ± 12%. Due to the differences in half-lives of 76 Kr and 77 Kr, it was possible to isolate a pure sample of 76 Br with 99.9% radionuclidic purity. The successful collection of krypton radioisotopes to generate 76 Br and 77 Br demonstrates the feasibility of gas-phase isotope harvesting from irradiated accelerator cooling-water. Larger-scale collections are planned for collecting by-product radionuclides from the Facility for Rare Isotope Beams.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Basis for the ICRP’s updated biokinetic model for systemic astatine

The International Commission on Radiological Protection (ICRP) recently updated its biokinetic models for workers in a series of reports called the OIR (occupational intakes of radionuclides) series. A new biokinetic model for astatine (At), the heaviest member of the halogen family, was adopted in OIR Part 5 (ICRP in press). Occupational intakes of radionuclides: Part 5). Furthermore, this paper provides an overview of available biokinetic data for At; describes the basis for the ICRP's updated model for At; and tabulates dose coefficients for intravenous injection of each of the two longest lived and most important At isotopes, 211 At and 210 At. At-211 (T 1/2 = 7.214 h) is a promising radionuclide for use in targeted α-particle therapy due to several favourable properties including its half-life and the absence of progeny that could deliver significant radiation doses outside the region of α-particle therapy. At-210 (T 1/2 = 8.1 h) is an impurity generated in the production of 211 At in a cyclotron and represents a potential radiation hazard via its long-lived progeny 210 Po (T 1/2 = 138 days). Tissue dose coefficients for injected 210 At and 211 At based on the updated model are shown to differ considerably from values based on the ICRP's previous model for At, particularly for the thyroid, stomach wall, salivary glands, lungs, spleen, and kidneys.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

IDAC-Bio, A Software for Internal Dosimetry Based on the New ICRP Biokinetic Models and Specific Absorbed Fractions

Radiation dosimetry is central to virtually all radiation safety applications, optimization, and research. It relates to various individuals and population groups and to miscellaneous exposure situations—including planned, existing, and emergency situations. The International Commission on Radiological Protection (ICRP) has developed a new computational framework for internal dose estimations. Important components are more detailed and improved anatomical models and more realistic biokinetic models than before. The ICRP is currently producing new organ dose and effective dose coefficients for occupational intakes of radionuclides (OIR) and environmental intakes of radionuclides (EIR), which supersede the earlier dose coefficients in Publication 68 and the Publication 72 series, respectively. However, the ICRP only publishes dose coefficients for a single acute intake of a radionuclide and for an integration period of 50 years for intake by adults and to age 70 years for intakes by pre-adults. The new software, IDAC-Bio, performs committed absorbed dose and effective dose calculations for a selectable intake scenario, e.g., for a continuous intake or an intake during x hours per day and y days per week, and for any selected integration time. The software uses the primary data and models of the ICRP biokinetic models and numerically solves the biokinetic model and calculates the absorbed doses to organs and tissues in the ICRP reference human phantoms. The software calculates absorbed dose using the nuclear decay data in ICRP publication 107. IDAC-Bio is a further development and an important addition to the internal dosimetry program IDAC-Dose2.1. Here, the results generated by the software were validated against published ICRP dose coefficients. The potential of the software is illustrated by dose calculations for a nuclear power plant worker who had been exposed to varying levels of 60 Co and who had undergone repeated whole-body measurements, and for a hypothetical member of the public subject to future releases of 148 Gd from neutron spallation in tungsten at the European Spallation Source.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Analcime-wairakite formation during experimental cement-bentonite alteration at 200–300 °C

Engineered barrier system materials in nuclear waste repositories may undergo hydrothermal alteration in response to groundwater saturation and heating events over their long operational timescales. Hydrothermal interactions between engineered materials (e.g., bentonite buffers, cements, waste canister materials) and the host rock environment may drive alteration processes that affect the advantageous properties of some barrier materials. However, such alteration will also promote the formation of zeolites, hydrothermally formed minerals that may themselves act to isolate radionuclides. Understanding the environmental conditions that control radionuclide-sorbing properties of zeolites, such as Si/Al ratio or Na content, will be valuable for assessing the changing properties of engineered barrier materials in the case of an in situ heating event. Here, we present experimental work characterizing the formation of zeolites during hydrothermal interactions between generic clay and cement barrier materials, with a focus on the analcime-wairakite zeolite series, which has known radionuclide sorption and exchange properties. We present the results of hydrothermal experiments combining uncured ordinary Portland cement powder with steel and Wyoming bentonite in the presence of Opalinus clay and a synthesized Opalinus clay groundwater to simulate water-saturated conditions in an argillaceous rock repository. The experiments were conducted isothermally at 200 or 300 °C for 8 to 24 weeks. In contrast to analogous studies that did not include cement reactants, we observed the formation of analcime-group minerals in all experiments. The addition of cement resulted in lower Si/Al ratios in the zeolites compared to similar studies that did not include cement. The presence of cement as a reactant was interpreted to promote analcime formation at 200 °C. At 300 °C, we observed higher calcium and silica concentrations in the aqueous solutions as well as increased wairakite formation and decreased analcime formation compared to the experiments at 200 °C. These results show a fully realized analcime-wairakite solid solution that falls between Si/Al = 2 (ideal) and a trend of analcime minerals that have increasing Si/Al ratios with increasing Na/(Na+Ca). Our results show predictable relationships between the Si/Al ratio and the analcime-wairakite content under hydrothermal conditions and illustrate that repository material interactions may promote the formation of zeolites in the analcime-wairakite solid-solution series during heating events in the subsurface.

58 GEOSCIENCES↗
DOE OSTI2022

Integrated Disposal Facility FY2011 Glass Testing Summary Report [Erratum]

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 x 10 5 m 3 of glass (Certa and Wells 2010). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 8.9 x 10 14 Bq total activity) of long-lived radionuclides, principally 99 Tc (t 1/2 = 2.1 x 10 5 ), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2011 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗