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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Development and Characterization of Cementitious Waste Forms for Immobilization of Granular Activated Carbon, Silver Mordenite, and HEPA Filter Media Solid Secondary Waste

At the Department of Energy’s Hanford site, over 53 million gallons of chemically complex and radioactive wastes have been stored in 177 underground tanks. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is under construction and is designed to treat and immobilize these wastes. During operations of WTP, solid secondary wastes (SSWs) will be generated as a result of waste treatment, vitrification, off-gas management, and supporting process activities. SSW treatment processes and resulting disposal pathways for the final disposition form of the SSW are needed to support direct feed low activity waste (DFLAW) operations and facilitate continued operation of WTP. The SSWs produced through WTP operations are expected to include used process equipment, contaminated tools and instruments, decontamination wastes, high-efficiency particulate air (HEPA) filters, carbon absorption beds (granular activated carbon, GAC), silver mordenite (AgM) and spent ion-exchange resins. These waste streams are planned to be immobilized in a cementitious waste form and disposed of either as stabilized/blended (non-debris) or encapsulated (debris) in a cementitious waste form. Accordingly, cementitious waste forms from these streams were included in the 2017 Integrated Disposal Facility (IDF) Performance Assessment (PA). The input data used to represent these SSW forms in the 2017 IDF PA involved many assumptions and associated uncertainties. This data limitation was due to the lack of material- and site-specific data available for representative SSW materials in cementitious matrices. To verify the assumed values used in the IDF PA and fill this limitation in available data, Washington River Protection Solutions, LLC (WRPS), has initiated a program targeted toward gathering site specific data relevant to Hanford SSW disposal. The work within this report is a continuation of this ongoing program.

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DOE OSTI2026

Quantification of trace iodine using laser-induced breakdown spectroscopy for real-time monitoring of nuclear off-gas streams

This study evaluated the potential of laser-induced breakdown spectroscopy (LIBS) for real-time monitoring of trace gas-phase iodine, which is an element of high significance in nuclear applications due to its long radioactive half-life (as iodine-129), volatility, and biological impact. In anticipation of iodine evolving into off-gas systems in molten salt reactor and nuclear fuel recycling applications, this research aimed to assess LIBS performance in flowing argon and helium matrices; optimize measurement parameters using a multichannel spectrometer; and perform calibrations to assess predictive capabilities and limits of detection (LODs). Experimental results successfully measured gas-phase iodine in flowing argon and helium; however, trace iodine was not detected in air. Optimal delay times were determined to be 10 µs for argon and 1 µs for helium, which are consistent with the expected shorter plasma lifetime in helium relative to argon. An emission line survey was provided with the 206.16, 804.37, 902.24, and 905.83 nm peaks, which were identified as the strongest emission peaks. Calibration models were successfully built in both helium and argon, achieving LODs down to 3 ppm in helium and 5 ppm in argon. The iodine emission at 905.83 nm emerged as the most robust for calibration and was subsequently applied to a time series dataset in argon. The predictive trace confirmed the feasibility of employing LIBS for continuous, online quantification of trace iodine in flowing gas systems.

Andrews, Hunter B. [Oak Ridge National Laboratory ↗
DOE OSTI2026

Evaluation of Iron‐Phosphate Glass–Ceramic Waste Form for Electrorefiner Salt Waste Simulant Dechlorinated With Phosphoric Acid

The importance of glass and glass–ceramic nuclear waste forms has been reaffirmed in recent years by the growing interest in nuclear power as a reliable energy source. Determination of processing methods for the disposal of halide-containing wastes will be essential for the advancement of nuclear technologies such as non-aqueous fuel reprocessing. Phosphate-based dechlorination and subsequent vitrification of radioactive salt waste into an iron-phosphate waste form have been identified as a potential processing scheme for electrochemical processing waste. The impact of H 3 PO 4 -based dechlorination of complex salt mixtures on the vitrification process and structure of the final iron-phosphate waste form has not yet been investigated. In this work, iron-phosphate glass–ceramics were made from simulant salt waste (48LiCl–33KCl–19NaCl mol%) dechlorinated with the H3PO4-based method. The glass-forming region was compared to that of traditionally prepared Na 2 O–Fe 2 O 3 –P 2 O 5 systems. For a candidate glass-forming composition, the processing scheme presented here was determined to favor Fe 3+ species. The O/P molar ratio was consistent for the candidate composition when dechlorinated at 400°C and 600°C in air and argon environments, indicating glass network connectivity was maintained despite variations in processing parameters. The results presented here validate processing schemes requiring iron-phosphate waste form synthesis following H 3 PO 4 -based dechlorination.

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DOE OSTI2021

Tritiated Sulfur Hexafluoride Disposition Strategies

Tritium (T 2 ) is a radioactive isotope of hydrogen that is produced in nuclear fission reactions and is often used in nuclear fusion reactions and accelerator-based applications for medical isotope production. As a hydrogen isotope, tritium can readily bind to hydroxyl radicals (OH), forming tritiated water (HTO or T 2 0), and to carbon atoms. Tritium decays to helium-3 ( 3 He) via beta-decay with max decay energy of 18.6 keV. While it is not an external radiation hazard, it can be an internal radiation hazard if tritium is inhaled, ingested, or absorbed through the skin. In applications where tritium is handled, tritium confinement is performed using different barriers to minimize releases to the environment. For gaseous (elemental), liquid (oxide), and metal (hydride) tritium, process piping and components provide the primary confinement function. Secondary tritium confinement is typically provided by inert (i.e. non-flammable gases such as nitrogen, argon, or helium) gloveboxes which are connected to a tritium stripper system. Primary tritium confinement barriers typically produce low volumes of high tritium concentrations of tritiated methane from carbon in steels or organic materials along with tritiated water/tritium oxide (e.g. HTO) and tritiated ammonia from reactions with oxygen and nitrogen. Tritium escaping primary confinement into secondary confinement atmospheres (e.g. gloveboxes) produce higher volumes of lower activity contamination than found in process piping. Tritium contamination also occurs by leaks or tritium permeation/diffusion through confinement materials. Tritium from inside primary confinement barriers will diffuse or leak out of the primary confinement barrier and usually into the air, if the system is inside an air hood or ventilated hot cell, or into the secondary confinement (e.g. glovebox) atmosphere which is either exhausted or stripped based on the function of the secondary confinement (glovebox) system. Accelerator based processes for medical isotope production represent an atypical tritium contamination challenge. In medical isotope production, deuterium supply gas is ionized and accelerated to a tritium gas target to produce neutrons that are then used to produce the medical isotopes through additional nuclear fission reactions. To create large voltage differentials for accelerator operations, an electrical insulation medium is needed to prevent or rapidly quench electric discharges. A common electrical insulation medium utilized in accelerator applications is sulfur hexafluoride (SF 6 ) gas. SF 6 has a high dielectric strength and allows for the construction of smaller accelerator systems compared to other electrical insulation mediums such as air or dry nitrogen. Due to tritium permeation/diffusion through accelerator process and confinement materials, there is the possibility that tritium can contaminate the electrical insulation medium of the accelerator. Tritium contaminated SF 6 creates a material without any obvious processes for managing the contamination, reuse, or disposal of the used SF 6 . This document will discuss possible management strategies for tritium contaminated SF 6 for accelerator-based processes for Molybdenum-99 (Mo-99) production.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Sandia National Laboratories Annual Site Environmental Report, 2019: Tonopah Test Range, Nevada and Kaua'i Test Facility, Hawai'i

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the United States Department of Energy (DOE) National Nuclear Security Administration. The National Nuclear Security Administration’s Sandia Field Office administers the contract and oversees contractor operations at Sandia National Laboratories, Tonopah Test Range (SNL/TTR) in Nevada and Sandia National Laboratories, Kaua‘i Test Facility (SNL/KTF) in Hawai‘i. Activities at SNL/TTR are conducted in support of DOE weapons programs and have operated at the site since 1957. SNL/KTF has operated as a rocket preparation launching and tracking facility since 1962. DOE and its management and operating contractor are committed to safeguarding the environment, assessing sustainability practices, and ensuring the validity and accuracy of the monitoring data presented in this Annual Site Environmental Report. This report summarizes the environmental protection, restoration, and monitoring programs in place at SNL/TTR and SNL/KTF during calendar year 2019. Environmental topics include air quality, ecology, environmental restoration, oil storage, site sustainability, terrestrial surveillance, waste management, water quality, and implementation of the National Environmental Policy Act. This report is prepared in accordance with and as required by DOE O 231.1B, Admin Change 1, Environment, Safety, and Health Reporting, and has been approved for public distribution.

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DOE OSTI2020

Drying Spent Magnox Fuel for Interim Dry Storage

Spent Magnox fuel is usually stored under water for shielding and cooling following use in-reactor, but the storage water can react with Magnox clad fuel and cause corrosion and forms a sludge which generates a secondary waste product. One option to prevent unwanted corrosion and avoid expensive waste management is to dry the wet fuel. This project aims to test a drying process which will convert wet and corroded Magnox fuel to a stable form to demonstrate viability for conversion from wet to dry storage. Samples of corroded Magnox supplied by National Nuclear Laboratory. Corrosion product retrieved by various methods including dry scraping, ultrasonic treatment to form sludge and sludge drying to isolate suspension. TGA undertaken on Sample 1 and Sample 2, with brucite and hydromagnesite for comparison. Both samples showed ∼10% mass loss up to 250 deg. C attributable to water held in sample. Mass loss from dehydroxylation which liberates further water in both samples. TGA data plotted alongside chemically pure brucite and hydromagnesite shows expected similarities/differences between Sample 1(mostly brucite) and Sample 2 (brucite/hydromagnesite mixture). Following treatment, dried samples observed to slowly gain mass in ambient conditions. Presumed to be hygroscopic absorption of air water vapour as effect is reversible, with similar effect observed for hydromagnesite and brucite. Etched Magnox vacuum dried for 3.5 h at 90 deg. C to test for reaction with residual water - some discoloration but no noticeable corrosion observed. Drying etched Magnox metal to observe surface effects/corrosion during drying process.

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DOE OSTI2026

A Guide to Nonaqueous Electrochemistry of f-Element Complexes

Electrochemistry is a powerful tool for assessing and understanding the redox chemistry of molecular complexes. Cyclic voltammetry enables the f-element community to study molecules in unusually high or low oxidation states, which potentially pose important broad-scope questions of electronic structure. In the pursuit of boundary-pushing compounds, reactive air- and moisture-sensitive species are often encountered, which can be challenging to characterize, especially when they are chemically incompatible with certain solvents, electrolytes, or electrodes or when their potentials lie outside of common electrochemical windows. Nonaqueous solvents and pseudo reference electrodes complicate many of the standard practices in acquiring high-quality and reproducible electrochemical data. This guide presents a detailed discussion of selecting appropriate cell conditions and referencing and addresses metrics for evaluating electrochemical and chemical reversibility. These methodological approaches have been extended to best practices for the electrochemical analysis of radioactive transuranic complexes.

Electrodes↗
DOE OSTI2020

Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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DOE OSTI2020

Geology and Design of Major Spent Fuel Repositories

This report summarizes the geology and design concepts of four major spent fuel repositories from around the world. The crystalline rock with bentonite backfill is currently implemented in Finland with another repository being constructed in Sweden. The salt formation design focuses on the Waste Isolation Pilot Plant (WIPP) repository in the United States. The volcanic tuff open air design is based on the Yucca Mountain location. The volcanic tuff open air design at Yucca Mountain is only a concept as this repository has not been approved for operations. The last spent fuel repository design, clay/shale formation with backfill, focuses on the Cigeo Project in France, which is also pending.

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DOE OSTI2020

Comprehensive Baseline Environmental Monitoring of an Interim Legacy Waste Store in Arid Zone Australia - 20062

Radionuclide containing soil and rubble from the clean-up of a former uranium, thorium and rare earth minerals processing research facility carried out in the early 50's and 60's is currently stored at the Woomera Test Range in South Australia awaiting final disposal. Prior to developing a final disposal pathway commensurate with waste acceptance criteria, a comprehensive baseline environmental monitoring has been carried out to establish the background radiation levels. The survey included in situ gamma scanning of the surface soil ({sup 238}U, {sup 232}Th, {sup 40}K), a laboratory analysis of 86 soil samples (from 0-20 cm depth), which involved XRF measurements (U, Th, and rare earth metals) and spectroscopic analysis ({sup 238}U, {sup 232}Th, {sup 226}Ra and daughter nuclides). A time series of radon concentration measurements in air was established, while soil radon ({sup 222}Rn) and thoron ({sup 220}Rn) flux measurements were undertaken at 18 different locations. The 2018 comprehensive baseline survey demonstrated radiation levels in the immediate vicinity of the Woomera interim waste store are typical background values of soil and rock. No contamination was detected. (authors)

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DOE OSTI2021

A novel defoamer for processing nuclear waste: Testing and performance

Abstract Legacy radioactive waste from nuclear weapons material processing is the biggest environmental challenge in the state of South Carolina, and one of the biggest in the United States. Although substantial progress has been made in processing and vitrifying high‐level radioactive waste at the Savannah River site, approximately 35 million gallons remains to be treated and dispositioned. In this article, we show the development of a new defoamer for use in the processing of high‐level radioactive waste. The application of the new defoamer is not only more effective at controlling foam but will shorten the batch processing time leading to safer and more efficient processing. This has the potential to help shorten the overall site mission lifetime, saving the federal government hundreds of millions of dollars. The novel aspect of the new defoamer is its effectiveness and stability in the harsh conditions needed for processing high‐level radioactive waste: high temperatures, pH 4–13, strong oxidizing and reducing agents, and numerous metal catalysts. The defoamer will replace an existing antifoam agent that decomposes to form multiple flammable gases. Because the existing antifoam agent decomposes quickly, processing changes increased processing time. Statement of novelty We demonstrate that a new defoamer has been developed for use in processing high‐level radioactive waste. This novel defoamer is effective and stable in the harsh processing conditions with no detectable flammable by‐products. Because the existing antifoam agent decomposes quickly, processing changes, including lowering acid addition rates and evaporation rates, increased processing time. The new defoamer is not only more effective at controlling foam but will shorten the batch processing time and lead to safer and more efficient processing by eliminating the production of flammable gases. Statement of industrial relevance Foaming is of great concern in many industrial processes involving three‐phase gas/liquid/fine‐solids systems, like in water evaporation, mineral floatation, air sparging in situ remediation techniques, and in those found in the paper industry.

Lambert, Dan P.↗
DOE OSTI2023

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2022)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2022, extending from October 1, 2021, through September 30, 2022), approximately 8.52E+04 metric tons (9.39E+04 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2022, approximately 17.0 million metric tons (18.7 million U.S. tons) of waste has been disposed at ERDF, which equates to consumption of approximately 89.1% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the groundwater pathways and the air pathway inventory limits. Computed values were 5.50E-04 and 3.30E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

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DOE OSTI2022

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2021)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2021, extending from October 1, 2020, through September 30, 2021), approximately 9.14E+04 metric tons (1.01E+05 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2021, approximately 16.9 million metric tons (18.9 U.S. tons) of waste has been disposed of at ERDF, which equates to consumption of approximately 88.7% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operation within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the all pathways and the air pathway inventory limits. Computed values were 7.63E-03 and 1.49E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

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DOE OSTI2022

CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.

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DOE OSTI2020

Fueling Nuclear Research at MFC: The Environmental Backbone of Nuclear Operations

The National Environmental Policy Act (NEPA) review guidelines are an essential part of the approval process of any construction or modification being done at Idaho National Laboratory (INL). The Material and Fuels Complex (MFC) Environmental team is responsible for writing and reviewing Environmental Impact Statements (EISs), Environmental Assessments (EAs), Environmental Compliance Permits (ECPs), Air Permitting Applicability Determinations (APADs), and other NEPA compliance documentation. This documentation allows for innovation in MFC research by permitting the introduction of new laboratory equipment, research facilities, and materials for researchers to use in their experiments. Without the work of the Environmental team, nuclear research and operations at MFC would not be possible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Measuring aerosol collection efficiency for the Bladewerx “New Speclon TM 5” and the older “Speclon TM 5” filter

Bladewerx TM LLC (Rio Rancho, NM) manufactures instrumentation, neutron shielding and activation foils for the radiation protection industry. Specializing in portable alpha/beta air monitors and sample counters, Bladewerx is the source of Speclon TM PTFE filter media that they recommend for high-resolution alpha spectroscopy. Los Alamos National Laboratory (LANL) utilizes Speclon TM filter material in CAM (Continuous Air Monitor) samplers for workplace air monitoring. The LANL Aerosol Engineering Facility received air filter material from Bladewerx, referred to as “New Speclon 5” in this document, in order to distinguish from filter material that was previously received (referred to as “Speclon 5” in this document). In this document, the aerosol collection efficiency and airflow resistance (pressure drop) were measured for the New Speclon 5 filter material.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Survey Report on Commercial Off-The-Shelf (COTS) Tracking Tools

Radioactive/nuclear (R/N) materials are integral parts of many business sectors such as medicine, industry, and research. In the United States alone, approximately 3 million packages containing R/N materials are shipped annually via highways, rail, air, and water. These materials are utilized in medical diagnostics and treatments, industrial radiography, and scientific research. The secure and efficient transportation of R/N materials is paramount to public safety and national security. Recent incidents involving the loss of R/N materials during transportation highlight significant security and safety concerns. For instance, in December 2024, a package containing four encapsulated Selenium-75 sources was reported lost in transit from the Czech Republic to Madrid, raising alarms about the security of such shipments. Similarly, a Germanium-68 pin source was reported lost in transit in New Jersey in December 2024, underscoring the risks associated with transporting radioactive materials.

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

Hanford Waste Treatment Plant Low Activity Waste Facility Stack Effluent Monitoring - Sampling Probe Location Qualification Evaluation

The Hanford Tank Waste Treatment and Immobilization Plant low activity waste (LAW) facility stack monitor locations were qualified using scale model stacks to mitigate the risk of identifying that sampling locations do not meet the qualification criteria on the full-scale stack. As required by the American National Standards Institute/Health Physics Society (ANSI/HPS) N13.1-1999 standard, the scale model and its sampling location were geometrically similar to the actual stack and the Reynolds numbers for both the actual and model stacks were >10,000. An additional criterion is that the product of the hydraulic diameter and mean velocity (DV) of the full-scale stack must be between 1/6 DV and 6 DV of the scale model stack tests. Verification tests of the LAW stacks were performed at normal operating conditions. The minimum 1/6 DV value, along with the maximum 6 DV value from the scale model testing determines the range of conditions for which the full-scale stack may be operated and remain in compliance with the stack verification criterion. Based on these DV values, the corresponding stack flow rates for each of the LAW stacks are 815–55,758 scfm for LV-S1, 980–112,078 scfm for LV-S2, 264–22,901 scfm for LV-S3, and 981–79,832 scfm for LV-C2. The remaining criteria for the stack verification to be considered valid involve the flow angle and velocity uniformity results. First, the flow angle at the full-scale stack must be =20°. Second, the velocity uniformity at the full-scale stack must be =20% coefficient of variance (COV). Finally, the velocity uniformity results for the actual and scale model stack tests must agree within 5% COV. These criteria were met through the full-scale stack tests at the LAW facility. Flow angle results were primarily less than 10°, except for one LV-S2 Fan A result, which was 13.2°; all flow angle results were within the =20° criterion. The velocity uniformity results for each test condition ranged between 1.5 COV and 9.2% COV, which were all within the range of the target % COV values from the scale model tests. Based on these stack verification test results, the four LAW filtered exhaust stack sampling locations meet the qualification criteria provided in the ANSI/HPS N13.1-1999 standard for all fan operating configurations. This includes single-fan operating conditions for LV-S1 and LV-S2, dual-fan operations for LV-S3 at both the continuous air monitor and record sampler locations, and both the single-fan as well as the dual-fan operations for LV-C2. Further changes to the system configuration or operating conditions that are outside the bounds described in this report may require additional tests or analyses to determine compliance with the standard.

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