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At least 253 records · Page 14

Glass-bonded ceramic waste forms for immobilization of radioiodine from caustic scrubber wastes

Glass-bonded sodalite composite waste forms have been developed for the immobilization of liquid radioactive wastes resulting from off-gas treatment during aqueous reprocessing of used nuclear fuel, with a particular focus on 129I. The proposed composite waste form is comprised of aluminosilicate ceramic phases containing volatile radionuclides bonded with a glassy matrix. In this work, a suite of ten candidate low-temperature glass binders (ZnO-Bi2O3-based glasses and a Na2O-B2O3-SiO2 glass) were examined. Six glasses were mixed with caustic scrubber waste simulant previously converted into a sodalite-rich material (to provide glass fractions of 10 and 20 wt.%), uniaxially pressed into pellets, and sintered at 350 °C or 550 °C for 8 h in air. Iodine retention after heat treatment was assessed by neutron activation analysis, showing retention of 67-100 % of expected iodine. The aqueous durabilities of the resulting materials were then determined, following the ASTM C1308 standard test, showing iodine releases of 1 to 23 g m-2 after 4 d. The cumulative iodine release for the best performing system (a zinc-bismuth-borate glass binder) was <1 g m-2, and its iodine retention from processing was 67 %. The iodine releases compared favorably with other waste forms. In parallel, this best-performing composition was also consolidated via hot isostatic pressing (HIP) in a stainless-steel canister at 550 °C for 2 h under 100 MPa pressure. The HIPed sample was produced at the ~20 g scale and showed improved densification and minimal reaction with the canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Monitoring Potential Transport of Radioactive Contaminants in Shallow Ephemeral Channels: FY2019

Desert Research Institute (DRI) conducted a field assessment of the potential for contaminated soil to be transported from the Smoky Site Contamination Area (CA) because of storm runoff. This activity supported U.S. Department of Energy (DOE) Environmental Management Nevada Program (EM NV) efforts to establish post-closure monitoring plans for the Smoky Site Soils Corrective Action Unit (CAU) 550. The work was intended to confirm the likely mechanism of transport and determine the meteorological conditions that might cause the movement of contaminated soils, as well as determine the particle size fraction most closely associated with transported radionuclide-contaminated soils. These data will facilitate the design of the appropriate post-closure monitoring program. In 2011, DRI installed a meteorological monitoring station on the west side of the Smoky Site CA and a hydrologic (runoff) monitoring station within the CA, near the east side. The meteorological station collected air temperature, wind speed, wind direction, relative humidity, precipitation, solar radiation, barometric pressure, soil temperature, and soil water content data. The maximum, minimum, and average or total values (as appropriate) for each of these parameters were recorded for each 10-minute interval. The maximum, minimum, and average water depth in the flume installed at the hydrologic station were also recorded for each 10-minute interval. This report presents the data collected from these stations during fiscal year (FY) 2019. During the FY2019 reporting period, the warmest months were June, July, and August and the coldest were December and February. The highest monthly solar radiation (i.e., sunshine) values were recorded in June, July, and August, whereas December through February had the lowest solar radiation values. Monthly mean wind speeds were highest in the spring (April) and early summer (June). Winds were predominantly from the north, except in February, July, August, and September when winds were mostly from the west. The monthly average relative humidity ranged from 17 percent to approximately 60 percent. Humidity was lowest in the summer and highest in the winter. Monthly total precipitation ranged from 0.0 during August and September to 2.22 inches (in) (56.39 millimeters [mm]) in February. Total precipitation for FY2019 was 7.83 in (198.88 mm). During the reporting period, a single, nonzero, flow event was recorded at the flume. This occurred over two days between March 6 and 7, 2019, in response to 1.16 in (29.46 mm) of precipitation falling within 9.3 hours on March 6, 2019. This event produced a peak water depth of 0.39 in (0.99 centimeters [cm]) in the flume, and runoff was measured for approximately 23.5 hours. However, the maximum recorded flow depth was less than the limit for flow discharge and velocity that can be calculated for the six-inch (15 cm) throat of a Parshall flume. The runoff event produced insufficient flow velocities through the natural channel to cause local erosion and transport of bedload materials. No bedload samples were collected for radiological analysis. Observing meteorological and environmental conditions that lead to storm runoff will help identify the parameters and threshold conditions that should be incorporated into post-closure monitoring plans for this and similar Soils CAU sites. Routine monitoring of meteorological and hydrologic parameters at the Smoky Site CA has been discontinued at the request of EM NV, effective at the conclusion of FY2019. Environmental data acquisition was completed and the meteorological and runoff monitoring stations were removed from the Smoky Site on October 15, 2019.

54 ENVIRONMENTAL SCIENCES↗

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY22 Status Update)

High-level purpose of this work: This report summarizes work carried out by Sandia National Laboratories (SNL) in the fiscal year 2022 (FY22) to evaluate the potential occurrence of stress corrosion cracking (SCC) on spent nuclear fuel (SNF) dry storage canisters. The U.S. currently lacks a repository for permanent disposal of SNF; thus, dry storage systems will be in use for much longer time periods than originally intended. Gap analyses by the US Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), the Nuclear Waste Technical Review Board (NWTRB), and the Electric Power Research Institute (EPRI) have all determined that an improved understanding of the occurrence and risk of canister SCC is critical to demonstrating the safety of long-term dry storage. Should canister penetration by SCC occur, the containment boundary represented by the canister would be breached. A loss of the inert environment (helium) within the canister could occur and intrusion of air and moisture could react with and damage the fuel within the canister. For this reason, the DOE is funding an effort to evaluate the potential occurrence and consequences of dry storage canister SCC and to develop prevention, mitigation, and repair technologies for this degradation mechanism.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Tank 14H Preliminary Cease Waste Removal Sample Analyses

Savannah River Mission Completion Waste Removal and Tank Closure requested that Savannah River National Laboratory perform characterization analyses of Tank 14H in support of closure activities, specifically preliminary cease waste removal. This report provides important characterization of the prepared filtrate and air-dried solids in the January 2026 Tank 14H sample, HTF-14-26-1,2,3,4,5,6, for I-129, Tc-99, Cs-137, and Sr-90. The results reported here were utilized by SRMC to declare preliminary cease waste removal for Tank 14H.

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Protocol for Determining Actual Flow Rate in FTWC Duct Systems

This paper is intended to document the theory and methods to determine the actual flow rate in the exhaust system used to ventilate the Flanged Tritium Waste Containers based on realtime instrument readings. Section 1 provides introductory and background material; Section 2 details the various correction factors used to adjust flow readings for ambient air pressure and temperature and other operational parameters. Section 3 provides guidance on implementing this process during venting operations, and Section 4 provides supporting documentation and also a field checklist that can be used by personnel during venting. For Revision 1, a summary of editorial changes made to the paper was added as Section 4.4, as well as a section documenting peer review (Section 4.5).

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West Valley Demonstration Project (WVDP) Annual Site Environmental Report (ASER) for Calendar Year 2024

The report, prepared for the U.S. Department of Energy West Valley Demonstration Project office (DOE-WVDP), summarizes the environmental protection program at the WVDP for calendar year (CY) 2024. Monitoring and surveillance of the facilities used by the DOE are conducted to verify protection of public health and safety and the environment. The report is a key component of DOE’s effort to keep the public informed of environmental conditions at the WVDP. The quality assurance protocols applied to the environmental monitoring program ensure the validity and accuracy of the monitoring data. In addition to demonstrating compliance with environmental laws, regulations, and directives, evaluation of data collected in 2024 continued to indicate that WVDP activities pose no threat to public health or safety, or to the environment.

54 ENVIRONMENTAL SCIENCES↗

The Argonne SuperGel for CBRN Decontamination - 20160

The Argonne SuperGel was developed between 2003 and 2015 to fill a gap in our nation's capability to quickly decontaminate important structures following a radiological or nuclear release event. Specifically, the decontamination technology was developed to minimize damage to monuments, high valued structures, and critical infrastructure while reducing environmental and health impacts. An important criterion during its development, common reagents were employed that could be easily acquired in order to minimize the timeline for its deployment. Its current formulation uses off-the-shelf super-absorbing hydrogels common to the food and agricultural industry and common salts. Over the years, two formulations of the Argonne SuperGel have been developed to specifically target radioactive cesium contaminations and then, more generically, actinide and fission product contaminations. A biodegradable derivative of phosphoric acid is used in small amounts to promote the removal of insoluble actinide species. The original order of unit operations for the SuperGel technology was 1) applying the wash solution to promote mobility of the contaminants on the surface, 2) applying the hydrogel to absorb the wash solution and contaminants, and 3) removing the hydrogel for disposal. During the course of our studies, we tested a new formulation of the gel technology that eliminated the need for a separate step to apply the wash solution without a compromise in decontamination factors. Greater than 70% and >95% of Cs-137 were removed from concrete and tile coupons, respectively, after two decontaminations, with the best results using wash solutions of 1.0 M KCl and 1.0 M NH{sub 4}Cl formulations. We found no statistical difference between results at 30 deg. C and 40 deg. C or at 70% and 90% relative humidity. We tested the gel under strong UV irradiation to simulate extreme environments. While a tropical noontime UV flux did not adversely affect the gel system decontamination of Cs-137 from concrete or tile coupons, the noontime flux resulted in significant dehydration of the gel after two hours of constant exposure. The rehydration of the gel by the humid air (at 90% humidity and 40 deg. C) was slow and may not be sufficient to offset dehydration by the sun. Americium decontamination from concrete was 70% after optimizing the phosphoric acid derivative and carbonate salt concentration in the wash solution. Since its development, we have had the opportunity to test the SuperGel in the removal of legacy contaminations in hot cell facilities and former glovebox facilities at Argonne. This has provided a unique opportunity to evaluate the SuperGel on a range of contaminants outside the original specifications for its use. We will report on the origin of the gel formulation, some highlighted experimental data including independent testing by the US EPA which has never been reported, and its efficacy for removing legacy alpha contaminations and its potential use for removal of chemical and biological hazardous agents. (authors)

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Iodine Speciation Effects in LAW Feeds

Tests were conducted on a DM10 vitrification system with prototypical off-gas treatment system, which includes an SBS, WESP, and evaporator based on the design of the evaporator to be installed at the WTP for concentration of the liquid effluents. This allowed determination of the partitioning and speciation of iodine between the melter and the off gas system components, including liquid effluents from the SBS and WESP as well as in the evaporator condensate and concentrate. Evaporator testing was decoupled from melter testing to permit filtration and pH adjustments of process solutions prior to evaporation. In addition, a slip stream was pulled from the exhaust downstream of the WESP and High-Efficiency Particulate Air (HEPA) filters and passed through a caustic scrubber simulating the WTP caustic scrubber to assess iodine partitioning to the caustic scrubber liquid effluent. The tests were conducted using simulated Hanford LAW from tank AP-107, which is the planned first feed for Direct Feed LAW (DFLAW) operations, at a nominal feed concentration of 5.6 molar Na over about 120 hours of testing.

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Krypton Concentration using HZ-PAN

Idaho National Laboratory (INL) has developed and tested engineered sorbents to separate and capture volatile fission products such iodine, xenon (Xe), and krypton (Kr) from off-gas streams. As noble gases, Xe and Kr can be difficult to capture and separate. Historically, cryogenic distillation has been used to execute the separation. However, performing this separation from bulk air streams is expensive and can pose significant hazards. INL has successfully developed two different sorbents for Xe and Kr capture, silver mordenite polyacrylonitrile and hydrogen mordenite polyacrylonitrile (HZ-PAN). Adsorption studies to date successfully separated and captured Kr from carrier gas streams, but those studies focused primarily on initial separation and capture. Successful utilization in a used nuclear fuel (UNF) reprocessing facility, however, requires further concentration of the Kr to minimize long term storage volumes. This study focuses on the concentration of Kr utilizing HZ PAN as the concentrating media. Concentrating Kr during desorption is ideal for two reasons: one, to produce Kr that can be used for commercial and research applications, or two, to minimize the volume of the radioactive gaseous waste stream for disposal. This report investigates the Kr concentrating potential of HZ-PAN by running multiple Kr adsorption-desorption cycles, collecting the desorbed effluent from a saturated HZ-PAN column, and loading it onto a fresh column. This study demonstrates that four adsorption-desorption cycles can transform a 150 ppmv Kr stream into a 37.5% Kr stream, an overall concentration factor of 2497. Each concentration step results in successively smaller volumes desorbed, reducing volume by a factor of approximately 40,000. Depending on the operational goals, during desorption one could collect smaller volume fractions of high concentration Kr (~ 67%). This demonstration should be considered proof-of-concept. Further refinement is necessary to develop optimum operating schemes to integrate into UNF off-gas treatment.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

Request for Exemption and Permit Modification for S-6008: Sewer Diversion Complex

Lawrence Livermore National Laboratory (LLNL) operates its Sewer Diversion Complex (SDC) under two relevant permits: discharges to air are controlled by Condition 5533 of LLNL’s BAAQMD permit for Source 6008, and discharges to the municipal sewer system are controlled by LLNL’s Wastewater Discharge Permit (Permit #1250). LLNL intends to modify the SDC by changing the currently permitted tanks to new models. Generally, this change will increase the volume of the tanks, update the abatement carbon system to suit the larger volume, and switch to a set of taller but thinner tanks.

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Simple model for the reconstruction of radionuclide concentrations and radiation exposures along the Techa River

The Techa River (Southern Urals, Russia) was contaminated in 1949-1956 by liquid radioactive wastes from the Mayak complex, the first Russian facility for the production of plutonium. The measurements of environmental contamination were started in 1951. A simple model describing radionuclide transport along the free-flowing river and the accumulation of radionuclides by bottom sediments is presented. This model successfully correlates the rates of radionuclide releases as reconstructed by the Mayak experts, hydrological data, and available environmental monitoring data for the early period of contamination (1949-1951). The model was developed to reconstruct doses for people who lived in the riverside communities during the period of the releases and who were chronically exposed to external and internal irradiation. The model fills the data gaps and permits reconstruction of external gamma-exposure rates in air on the river bank and radionuclide concentrations in river water used for drinking and other household needs in 1949-1951.

Non-NASA Center↗

FY2022 Filter Test for TRU Waste Drum Prototype Unit

Continuing last year’s (fiscal year 2021) work, a prototype instrument and an NPI-6 Integrated Work Document (IWD) were developed to assess the functionality of filters on transuranic waste containers (commonly called 55 Gallon ring-top drums) without requiring the removal of the drum lid. The purpose of this work is to determine the air flow and pressure characteristics associated with filter clogging, filter pressure drop at a fixed flow rate, and leakage around the lid seal. The FY22 objectives included: (A) Development of prototype device, (B) Drafting an IWD document, (C) Defining a filter clogging parameter, (D) Assessing filter damage due to over-pressure events, and (E) A path forward to obtain approval of the system safety filters in compliance with P101-16 Industrial Ventilation – non HVACR (LANL 2022).

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Remote Handled Hot Cell Facility and Other Research Transuranic Waste (Debris)

This acceptable knowledge (AK) Summary Report has been prepared for the Central Characterization Program (CCP) for remote-handled (RH) transuranic (TRU) waste generated and managed by Sandia National Laboratories/New Mexico (SNL/NM) in Albuquerque, New Mexico. The waste described in this report was predominately generated in the SNL/NM Hot Cell Facility (HCF) during the removal and packaging of experimental material and decontamination operations in Building 6580 at Technical Area (TA)-V. The waste stream also includes a very small amount of waste (estimated at less than 1 gram of fuel similar to that used for research at the HCF) originating from classified research at TA-I. In addition, the waste also includes secondary waste termed by SNL/NM as process generated waste (PGW), created during repackaging operations for this waste at the Auxiliary Hot Cell Facility (AHCF). The waste has been stored at the Sandia Pulsed Reactor (SPR) Dense Pack Storage Facility within TA-V or the Manzano Bunkers located at Manzano Base within Kirtland Air Force Base. All of the waste is being repackaged at the AHCF. This report was prepared in accordance with CCP-TP-005, CCP Acceptable Knowledge Documentation (Reference 1), to implement the AK requirements of DOE/WIPP-02-3214, Remote-Handled TRU Waste Characterization Program Implementation Plan (WCPIP) (Reference 2); Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan (WIPP-WAP) (Reference 3); and DOE/WIPP-02-3122, Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC) (Reference 4).

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Mechanistic Understanding of Silver Sorbent Aging Processes in Off-Gas Treatment

The effect of aging in various off-gas streams including dry air, humid air, 1% NO/N 2 , 2% NO 2 /air, and a gas mixture consisting of 99.25% humid air (dew point: -15 °C), 0.25% NO/N 2 , and 0.5% NO 2 /air in terms of concentration on reduced silver exchanged mordenite (Ag 0 Z) and silver functionalized silica aerogel (Ag 0 -aerogel) were studied in this project. Utilizing simple and sophisticated experimental information combined with molecular and macroscopic mechanistic modeling, this project has significantly advanced the current understanding of aging for silver bearing adsorbents in nuclear-fuel-reprocessing off-gas streams. In addition, this project has produced models that can be utilized to predict the influence of adsorbent aging on the uptake kinetics and capacity of aged adsorbents for iodine and iodine-containing organic compounds. The results of the study have been published in the open literature, and can be used to help DOE better predict the removal of radioactive gases from nuclear-fuel-reprocessing off-gas streams.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Radon Survey of Los Alamos National Laboratory Buildings and Los Alamos County Residential Buildings

A radon survey of office spaces was of interest to LANL for several reasons. First, exposure to natural radon in LANL office spaces has been considered to be outside the regulatory reach of the federal regulations and the U.S. Department of Energy (U.S. DOE) orders, and thus has not been studied as extensively as occupational exposure to other radionuclides and is not subject to the same rigorous radiological controls. Second, a general survey of radon concentrations in LANL workspaces was of interest to management to ensure safe environments for workers and is in keeping with the intent of OSHA requirements. Though radon in LANL offices would not be derived from enhanced radioactive sources, radon exposure in general office spaces has the potential to significantly impact worker risk. The purpose of the 2009 radon survey was to measure and document indoor radon levels across a broad spectrum of office-type workspaces and neighboring homes. The measured concentrations were compared against those measured across the United States and across the world. The results were also compared against a wide variety of radiation protection thresholds such as (1) the action levels of 148 Bq m -3 (the U.S. EPA action level for public housing), (2) ≈1,200 Bq m -3 OSHA threshold for office spaces, (3) effective dose threshold of 1 mSv threshold for defining a radiological worker, (4) 50 mSv occupational effective dose limit, and (5) 0.1 mSv and 1 mSv limits for public exposures from the air pathway and all pathways, respectively.

07 ISOTOPE AND RADIATION SOURCES↗

Impact of carbonation on leaching of constituents from a cementitious waste form for treatment of low activity waste at the DOE Hanford site

Carbonation can be a major aging process during disposal of alkaline cementitious waste forms and can impact constituent leaching by changes in material alkalinity, pore structure, and controlling mineral phases. The effect of carbonation on the leaching of major and trace constituents from Cast Stone, a cementitious waste form developed to treat high salt content low activity waste, was studied through a combination of leaching experiments and reactive transport simulations. Diffusive transport of constituents in the waste form was evaluated using reactive transport modeling of diffusion-controlled leaching test results and a geochemical speciation model derived from pH-dependent leaching. Comparisons between Cast Stone materials aged under nitrogen, air, and 2% carbon dioxide in nitrogen showed that carbonation impacts solubility, physical retention and observed diffusivity of major and trace constituents. Carbonation under 2% CO 2 decreased the diffusion-controlled leaching of chromium by two orders of magnitude. Finally, modeling results suggest that carbonation may also decrease solubility of technetium while changes to microstructure by carbonation increases effective diffusivity of constituents in Cast Stone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nuclear Waste Tank Emission Contributions to Particle Size Distribution

Pollutants from anthropogenic activities including industrial processes are ubiquitous to the environment. To understand the impact from industrial aerosol on climate and human health, industrial aerosol needs to be better characterized. Here, in this study, particle number concentrations were used as a proxy for atmospheric pollutants, which include both particles and gases. Particle concentration and size distribution were measured using a scanning mobility particle sizer (SMPS) approximately 4.5 km from primary industrial areas at the Savannah River Site in Aiken, SC. Industrial areas include numerous nuclear waste storage and processing tanks. The SMPS data were divided into two groups depending on the wind direction measured onsite to categorize transport from the industrial area or from elsewhere. Industrial contributions were found to have a higher concentration of particles with sizes less than 200 nm, 859 ± 564 cm -3 , in comparison to non-industrial attributed particles, 733 ± 495 cm -3 on average from March-July 2021. For sizes larger than 200 nm, industrial and non-industrial particles have a similar concentration, 89 ± 59 cm -3 and 99 ± 61 cm -3 , with non-industrial concentrations being slightly larger. To confirm that industrial particles could travel to the sampling location, air dispersion modeling was completed for specific case studies during the sampling period. The atmospheric dispersion modeling results confirmed that particles released at the industrial areas reached the sampling location when the wind direction was favorable for transport from the industrial areas. The greater concentration of smaller-sized particles in industrial emissions has implications for typical particulate measurements (PM2.5), heath impacts, and climatological influences.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Respirator Cartridge Performance Testing on Hanford Tank AX-101

Washington River Protection Solutions (WRPS) conducted tests using two types of chemical cartridges for use in air purifying respirators to determine the period of time that the cartridges would provide adequate protection to worker when exposed to a mixture of Chemicals of Potential Concern (COPCs) from vapors emanating from the headspace of tank AX-101 on the Hanford Site. Pacific Northwest National Laboratory (PNNL) was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and change schedules.

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