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DOE OSTI2024

FIU Project 2: Environmental Remediation Science & Technology [Slides]

FIU’s research under this project involves conducting basic and applied science to fill knowledge gaps and validate potential remediation technologies for contaminated soil and groundwater and the assessment of the fate and transport of contaminants in the environment. The aim of FIU’s research is to reduce the potential for contaminant mobility or toxicity in the surface and subsurface through the development and application of state-of-the-art scientific and environmental remediation technologies at the Hanford Site, Savannah River Site (SRS), and the Waste Isolation Pilot Plant (WIPP), which is the Nation’s only mined geologic repository for permanent disposal of transuranic waste. FIU collaborates with scientists from Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), Savannah River Ecology Laboratory (SREL), Los Alamos National Laboratory (LANL) and the DOE Carlsbad Field Office (CBFO) in order to plan and execute research that is synergistic with the work being conducted at the sites, and that supports the resolution of critical science and engineering needs which leads to a better understanding of the long-term behavior of subsurface contaminants. The knowledge gained through this research will be used to transform experimental and modeling innovations into practical applications deployed at the sites to support EM’s primary goal of expediting the closure of major contaminated soil and groundwater sites and waste units. Collaborative relationships between FIU and the national laboratories have provided large benefits over the years to FIU, the national laboratories, the DOE complex, and the DOE EM mission. By working closely with the national laboratories, FIU’s research is not only closely aligned with the cleanup mission priorities at the DOE sites, but complements and supports ongoing work at the national laboratories for screening of new remedial technologies. This coordination and leveraging of research efforts results in time- and cost-savings, and will accelerate progress of the DOE EM environmental restoration mission.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Improbability of Nuclear Criticality in Compacted Criticality Control Overpacks after Room Closure by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, nuclear criticality is improbable after salt creep causes compaction of criticality control overpacks (CCOs) disposed at the Waste Isolation Pilot Plant, an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality potential has been updated here because of the introduction of CCOs, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCOs during two representative conditions: (1) large salt block fall, and (2) gradual salt compaction (without brine seepage and subsequent gas generation to permit maximum room closure). Geomechanical models of rock fall demonstrate three tiers of CCOs are not greatly disrupted. Geomechanical models of gradual room closure from salt creep predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Criticality models of spheres and cylinders of 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular spacing demonstrate that an array of CCOs is not critical when surrounded by salt and magnesium oxide, provided the amount of hydrogenous material shipped in the CCO (usually water and plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

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DOE OSTI2022

Discrete-event Simulation Process Model for the Pyrochemical Processing of Plutonium at Los Alamos National Laboratory

The pyrochemical metal production operations that occur in the Plutonium Facility at Los Alamos National Laboratory perform plutonium purification with the aim to provide plutonium metal for a variety of defense- and non-defense missions within the National Nuclear Security Administration. The demands and constraints associated with the pyrochemical processing of plutonium are complex, making decision analyses challenging for program managers who require plutonium production for their mission applications. The construction of a discrete-event simulation process model is proposed to measure and report the process capacity, material throughput, equipment requirements, and dose accumulation for operators of the pyrochemical metal production operations. The process model, constructed in the ExtendSim™ software, will represent the cause-and-effect relationships between the pyrochemical processing environment and the process constraints, including criticality limitations, material control and accountability measures, chemical analysis requirements, and equipment availability. An accurate representation of the pyrochemical metal production process capacity through simulation modeling will be helpful to program managers in their efforts to forecast plutonium availability for mission applications. Furthermore, the proposed process model will be vital for future analyses that will measure the interactions between the pyrochemical metal production operations and the aqueous reprocessing operations and their ability to minimize transuranic waste disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Modeling 233 Pa Generation in Thorium-fueled Reactors for Safeguards

Thorium has been considered as a possible alternative to uranium for nuclear fuel for many decades. It is three to four times more abundant in the earth than uranium and produces significantly less long-lived transuranic nuclear waste. Some claim thorium poses fewer proliferation concerns than other fuel types largely due to 232 U buildup (and associated high energy gamma-emitting decay products) in the irradiated thorium fuel. However, to fully explore potential proliferation concerns, generation and subsequent decay of 233 Pa produced in the reactor core still must be studied. With its half-life of 27 days, 233 Pa decays to 233 U, which is an International Atomic Energy Agency (IAEA) defined special fissionable material that can be used for nuclear weapons production. With more research being dedicated to thorium-fueled reactors, and several of these reactor designs possessing online fuel processing (allowing for on-site protactinium separation), it is important to understand this potential proliferation pathway. In particular, it is theoretically possible to extract protactinium from the irradiated fuel salt before it decays into 233 U. This hypothetical potential diversion can become an even greater proliferation concern if the extracted protactinium is purified through a second separation of protactinium approximately ten days later to remove the short half-life decay products of 232 Pa and 234 Pa, thus resulting in a higher concentration of the 233 Pa isotope, which decays into weapons usable 233 U with hardly any 232 U or 234 U in it. To estimate the concern of this potential proliferation challenge of thorium, different nuclear material accountancy techniques were reviewed for their viability to quantify 233 Pa if extracted from used thorium fuel. Characteristics of interest included technology maturity, cost, precision, and time taken to acquire results. Some technologies, like hybrid K-edge densitometry and passive gamma spectroscopy, appear to be viable techniques based on current literature. Due to the limited scope of this project, only passive gamma spectroscopy was further investigated. Three different reactor types (PWR, CANDU, MSR) were modeled with mixed thorium-uranium oxide fuels that were burned until the fuel was spent. The protactinium in the used fuel was extracted at the time of shutdown and the change in isotopic content of the protactinium quantified. Gamma spectroscopy simulations were performed for the protactinium isotopes and their decay products at various decay times. Given the simplicity of the models and large assumptions made (e.g. no background, no shielding, no self-attenuation), the initial results indicate that though 233 Pa is detectible for all the reactor types modeled at all decay times (0 to 300 days), more work should be done with higher fidelity models.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

OPP-B332-032 Rev AD

The purpose of this procedure is to provide the approved method for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

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DOE OSTI2022

OPP-B332-035 Rev AD

This procedure provides instructions for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

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DOE OSTI2022

Analysis of a Bulk 237 Np Oxide Sample for Trace Actinides

A neptunium oxide sample was dissolved and trace actinide constituents were measured by isotope dilution ICP-MS (for 239–244 Pu, 233–238 U, 237 Np, 241–243 Am) and alpha spectrometry ( 238 Pu, 242 Cm, 244 Cm). This material contained significant quantities of transuranic elements, notably americium and curium. The isotopic composition of the material will continue to evolve over time due to radioactive decay; the measurements reported here include all isotopes decay corrected to a reference date of 1/1/2022. These factors may be considerations in criticality studies of the related 237 Np sphere.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

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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 OSTI2023

Exploratory Synthesis for Reactive f-Element Separations without Solvent

Solvent extraction is commonly used to separate different f-metals found in spent nuclear fuel, but this process results in large volumes of highly contaminated liquid waste that must be remediated or stored at significant expense. The project funded under this award was aimed at addressing solvent-related issues encountered in conventional f-metal separations. Our strategy focused on using solvent-free mechanochemical reactions to prepare lanthanide and actinide borohydride complexes so that they could be separated based on differences in their volatility. Key deliverables described in this report include proof-of-principle results demonstrating that borohydrides called aminodiboranates can be used for volatile lanthanide/lanthanide and lanthanide/uranium separations, as we proposed. Moreover, exploration of a closely related class of borohydrides called phosphinodiboranates allowed us to identify underlying chemical factors that control the volatility of trivalent lanthanide and uranium borohydride complexes with identical structures. Details of published and pending research products are provided, and these include comprehensive synthesis and characterization efforts required to support fundamental studies related to the proposed separations. These technical efforts supported the training of nine graduate students and three undergraduate researchers, and they established air-sensitive transuranic capabilities at the University of Iowa to enhance radiochemical workforce development.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Measurement Plan for Uncertainty Contributions to 252 Cf Waste Measurements

This measurement plan establishes the methodology for performing and analyzing mock-up measurements of Building 7930 Cell G Cf waste generated from 252 Cf product preparation to better determine the measurement uncertainty contributors. Understanding the uncertainty contributors will establish the total measurement uncertainty. This is critical because it will impact the administrative threshold at which personnel can discriminate between low-level waste (LLW) and transuranic (TRU) waste.

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DOE OSTI2023

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

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DOE OSTI2023

Report on the Analysis of Plutonium-238 Oxide, Campaign #P5PO

Oak Ridge National Laboratory (ORNL) is currently producing Plutonium-238 oxide for NASA space programs. Through the irradiation of NpO 2 pellets in the High Flux Isotope Reactor (HFIR), the production of Pu-238 oxide will be used for electrical power and heat on NASA spacecraft such as the Curiosity Mars Rover and the Voyager 1 and 2. The Transuranium Analytical Laboratory (TAL) leads the analytical effort for this project, which includes the dissolution of plutonium product oxide (PUP) and the analysis of the dissolved product solution using both radiological and inductively coupled plasma mass spectrometry (ICP-MS) techniques that are required to meet product specifications. The Transuranic Analytical Laboratory received three samples of Pu-238 oxide for analysis in January of 2022. This report documents the analytical protocols performed by the TAL along with the results produced from said analytical protocols. This report also includes a comparison with the general-purpose heat source (GPHS) specifications.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2024

Exploratory Synthesis for Reactive f-Element Separations without Solvent

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Scoping Thermal Response Calculations of RNS Waste During Transport to and Disposal at the WIPP

Sandia National Laboratories (SNL) was contracted by the United States Department of Energy Environmental Management (DOE-EM), Los Alamos Field Office to perform mechanical and thermal scoping calculations as part of a study seeking to understand the ignitability risk of the Remediated Nitrate Salts (RNS) waste drums during transportation from the Waste Control Specialists (WCS) facility to Waste Isolation Pilot Plant (WIPP) and permanent disposal of the waste at WIPP. The scoping thermal simulations described in this report pertain to thermal calculations performed with a packaging system consisting of one Standard Waste Box (SWB) loaded with drums placed inside a Standard Large Box 2 (SLB2). During transportation, the SLB2 is inside Transuranic Package Transporter Model III (TRUPACT-III), which provides the third layer of the packaging. Once at the WIPP, it is assumed the SLB2 is extracted from the TRUPACT-III and maintained above ground, and then subsequently placed underground for permanent disposal. In these proposed configurations, the space between the SLB2 and the SWB is always filled by a layer of insulation consisting of air-filled glass microbubbles except for the bottom which rests directly on the SLB2. The thermal scoping calculations described in this report specifically address whether the introduction of external heat inputs, combined with the contributions from the internally generated radiolytic decay heat and chemical reactions, lead to an unstable thermal state during the time of its movement and placement in the permanent disposal location. The external heat inputs are of two forms: 1) ambient thermal irradiation (e.g., solar and ambient storage/disposal temperatures) and 2) accident-induced fire. Three scoping calculation scenarios were derived as representative, conservative scenarios: 1A) TRUPACT-III transient transportation, 1B) SLB2 48-hour outdoor storage with solar radiation, and 2) fully-engulfing fire during SLB2 handling or emplacement following a steady-state analysis in a 38 °C environment. All the simulated scenarios are conservative relative to the operational conditions expected for handling the waste package during transportation and placement in the WIPP underground disposal unit. The predictions obtained from simulating the three exposure scenarios revealed that adding the SLB2 and the air-filled glass microbubbles to the transport and storage/disposal configurations provides additional thermal protection of the drums beyond what the SWB provides alone, both during long-term above ground insolation and underground during a fire accident. Under the current transportation/storage/disposal concepts, the degree of protection provided by the packaging concept is sufficient to prevent the waste from being ignitable. The simulation results demonstrate that there is adequate margin to safely transport and place the RNS waste from WCS to the WIPP under the current operational concept.

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DOE OSTI2024

SCALE Demonstration for Sodium-Cooled Fast Reactor Fuel Cycle Analysis

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, SCALE 6.3.1 capabilities for radionuclide characterization, criticality, and shielding were demonstrated for scenarios in the sodium-cooled fast reactor (SFR) nuclear fuel cycle. Three postulated accident scenarios were selected for analysis in this work. As a basis for all scenarios, irradiated fuel inventories were generated using SCALE/ORIGAMI. To cover multiple SFR design choices, two different types of SFR fuel, uranium/transuranic-loaded and U-based fuels, were considered. For the first scenario, SCALE/MAVRIC was used to calculate the radiation shielding and dose rates inside and outside of the containment building due to a drop of a spent fuel assembly from the fuel-handling system during unloading inside the containment building. For the second scenario, potential critical configurations in an electrofiner were investigated through criticality calculations with SCALE/CSAS. For the third scenario, the activity of the waste salt from an electrorefiner was evaluated using SCALE/ORIGEN. The dose rate produced by the analyzed SFR assemblies is similar to that produced by a typical pressurized water reactor (PWR) fuel assembly with a discharge burnup of 50 GWd/MTU, with the same cooling time of 10 days. The criticality analyses suggested that the different electrorefiner configurations have a large margin to criticality. The activity analysis of the electrofiner waste revealed that shielding and cooling may be required for the waste salt that contains transuranics and fission products produced by the electrorefiner because of the high activity of the waste salt. In general, the application of various capabilities in the SCALE code system for SFR fuel inventory generation, criticality, and shielding was successfully demonstrated for the selected scenarios in the SFR nuclear fuel cycle. Additional analyses can be performed to provide more accurate results when more details of the SFR nuclear fuel cycles are available, for example, the dimension of the electrorefiner and the salt compositions during reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Prototype X-ray and Gamma Detection with Cyclotron Radiation Emission Spectroscopy

Cyclotron radiation emission spectroscopy, or CRES, is a novel approach to measuring the energy of an electron. By trapping a free electron in a high magnetic field, it undergoes cyclotron motion and emits radiofrequency (RF) waves. The frequency of this RF radiation is directly related to the energy of the electron. Because many cycles of the RF emission are recorded, the energy resolution of the CRES system is on the order of a single electron volt. To make a CRES system sensitive to photons, a target gas is used to induce a photoelectric effect, producing the electron that is subsequently trapped. By adding the binding energy of the target atom, the energy of the incident photon may be reconstructed. Using a xenon target gas, photoelectric interactions dominate up to approximately 300 keV, covering not only all atomic shell X-rays of the elements, but many low-lying nuclear states as well, including key transuranic elements related to nuclear security. CRES holds the potential of maintaining single-eV resolution up to this 300-keV range, thereby surpassing current state-of-the-art detectors by a factor of 10-100. The instrumental resolution of the system is limited by the uniformity of the applied magnetic field.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2024

Segmented Gamma Scanner for Radioactive Waste Assessment: A User Guide

Radioactive waste assessment is important for ensuring nuclear material security at various types of facilities, such as enrichment, fuel fabrication, and reprocessing plants. The waste generated at such nuclear facilities is stored in standard containers and is required to be characterized for material-accounting purposes. The segmented gamma scanner system is a popular, nondestructive analysis measurement system used for characterizing nuclear material, including radioactive waste. This document provides guidance on how to achieve effective performance from a segmented gamma scanner system for accurately quantifying fission products, activation products, and transuranic wastes.

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