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At least 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Characterization Results for the October 2024 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the October 2024 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 (137Cs), strontium-90 (90Sr), and iodine-129 (129I) with the radionuclide concentration limits specified by the WAC. These analyses were performed in duplicate, and a summary of the analytical results for this 3H evaporator overhead sample includes the following: The measured cesium-137 activity in the 3H evaporator overhead sample averaged 6.72E+01 dpm/mL, (3.90E+00 %RSD), which is below the ETP WAC limit of 1.30E+03 dpm/mL. The strontium-90 activity in the 3H evaporator overhead sample averaged <2.72E+00 dpm/mL, which is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample averaged <1.63E-01 dpm/mL, which is below the ETP WAC limit of 1.00E+00 dpm/mL.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Technology-Neutral, Accident Containment-Based Path to Reduce Nuclear Power Costs

Cost remains a substantial impediment to nuclear power plant (NPP) deployment. The origin of the multifold cost difference between NPPs and other industrial facilities is in their unique potential for societally disruptive accidents. Consequently, one means to substantially reduce NPP costs would be to provide a transparent means of demonstrating that a plant’s potential for land-contamination accidents is so remote to amount to a practical certainty that no significant quantity of radionuclides would ever reach the public. The necessary confidence can be developed through employing a containment capable of withstanding both a complete internal energy release accident along with credible external events with ample margin. This paper provides an overview of a path forward to develop a technology-neutral, containment-based safety-adequacy demonstration method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

2024 Results for Avian Monitoring at the Technical Area 36 Minie Site, Technical Area 39 Point 6, Technical Area 16 Burn Ground, and DARHT at Los Alamos National Laboratory

Los Alamos National Laboratory (LANL) biological subject matter experts in the Environmental Protection and Compliance Division initiated a multi-year program in 2013 to monitor avifauna (birds) at two open detonation sites and one open burn site on LANL property. Additional monitoring began in 2017 at a third firing site, the Dual-Axis Radiographic Hydrodynamic Test (DARHT) Facility. In this annual report, we compare monitoring results from these efforts among years to identify and evaluate firing and open burn site impacts on the local bird community. The objectives of this study are • to determine whether LANL operations impact bird abundance, species richness, or diversity; • to examine occupancy and nest success of secondary-cavity nesting birds that use nest boxes; and • to examine chemical concentrations (such as radionuclides, inorganic elements, and/or organic compounds) in nonviable eggs and deceased nestlings that are collected opportunistically with the upper-level bounds of background concentrations, when available. During May through July 2024, LANL biologists completed multiple avian point count surveys at each of the following treatment sites: • Technical Area (TA) 36 Minie Site, • TA-39 Point 6, • TA-16 Burn Ground, and • DARHT. We recorded a total of 1,088 birds that represented 65 species at the four treatment sites and compared these results with data from their associated control sites. In 2024, abundance and species richness at treatment and control sites continued to trend similarly from year to year, with minor random deviations expected from bird communities. Species richness at firing sites differed little from the previous year’s values. Two new bird species were observed at the firing sites—cedar waxwing (Bombycilla cedrorum) and pinyon jay (Gymnorhinus cyanocephalus). Shannon diversity values at TA-36 Minie Site, TA-39, and DARHT were statistically higher than one or more of their associated controls. Annual species diversity at treatment sites was high in 2024 across all firing sites relative to similar habitat control sites. We also monitored avian nest boxes to compare occupancy and nest success data from nest boxes at treatment sites with the overall avian nest box monitoring network and against a subset of relevant control sites. Nest box success has decreased at both treatment and control sites since monitoring began, suggesting that overlapping climatic factors are responsible for patterns of declining nest success. In 2024, nonviable avian eggs and one nestling were opportunistically collected at Bandelier National Monument, TA-16 Burn Ground, TA-36 Minie, TA-39 Point 6, and DARHT. All egg samples and the one nestling sample were evaluated for per- and polyfluoroalkyl substances, which were detected from all locations, including the control site at Bandelier National Monument. Overall results from 2024 continue to suggest that operations at the four treatment sites are not negatively impacting bird populations. This long-term project will continue to monitor for any changes over time.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3

Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Argonne National Laboratory Site Environmental Report for Calendar Year 2024

This report discusses the status and accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2024. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, efficiency efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups and on-site members of the public was estimated. Results are interpreted with respect to the origin of radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Hanford Composite Analysis Special Studies FY2025 - Groundwater Flow Simulation for use in a Comparison of Simulated Concentrations in the Saturated Zone Estimated by the Plateau-to-River Model Versions 8.3 and 9.1

In order to meet the requirements of U.S. Department of Energy (DOE) in DOE O 435.1, Radioactive Waste Management, a composite analysis (CA) must be completed for the Central Plateau at the Hanford Site. The CA requires estimates of fate and transport of radionuclides in the groundwater from multiple sources within the modeling domain. This Environmental Calculation File (ECF) details the application of the Plateau-to-River (P2R) Model (CP-57037, Rev. 3, Model Package Report for the P2R Model: Version 9.1) to predict the flow of groundwater on the Central Plateau for the 10,000-year simulation to support a special analysis associated with the Hanford Site CA. The simulated flow field will support the simulation of fate and transport of contaminants for comparison to the results obtained as part of the CA to evaluate impact of the updated version of the P2R Model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Reassessment of Dose Contributions from the Paducah Cask Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater and inadvertent human intruder screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in the addition of U-235 being brought back into the list of radionuclides requiring inventory limits within the Paducah Cask special waste form (SWF). In addition, while reviewing the earlier vadose zone transport analysis efforts, an inconsistency was observed in the initial U-235P inventory that was supplied to the Consolidated Waste Tracking System (CWTS) back in 2009.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Results for the April 2025 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the April 2025 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained. The April 2025 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the First Quarter Fiscal Year 2025 (1QFY2025) and the Second Quarter Fiscal Year 2025 (2QFY2025).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Bioaccumulation Factor for Radium in Rio Grande Biota

The Los Alamos National Laboratory (LANL or Laboratory) Environmental Protection and Compliance Division’s Environmental Stewardship Group (EPC-ES) performs the radiological analysis portion of biota dose assessments using the computer program RESRAD-Biota (DOE 2004, DOE 2020). RESRAD-Biota uses a default bioaccumulation factor, B iv , for radium in aquatic animals. This bioaccumulation factor, also listed in Department of Energy (DOE) Standard, DOE-STD-1153-2019, is the ratio of a radionuclide concentration in fresh weight biota to the environment in which it resides, which for aquatic animals is water. DOE-STD-1153-2019 offers conservative values for these ratios that are specific to the radiological contaminant, the biota, and the environment. The B iv in RESRAD-Biota can be used for general screening (Level 1 screening). If the factor proves to be too conservative, then a site-specific B iv may be used. The biota dose assessment of fish from the northern New Mexico region of the Rio Grande fail the general screen, which indicates that further analysis is necessary. Level 2 analysis is site-specific but still has conservative limitations. Level 3 analysis is a more realistic and representative, sitespecific analysis. The purpose of this paper is to use data specific to the Rio Grande to calculate a new aquatic animal radium B iv . This B iv can then be used for Level 3 analysis of aquatic animals specifically in the northern New Mexico region of the Rio Grande. The calculated radium B iv is a useful value for performing detailed, accurate biota dose assessments and calculating dose to fish using RESRAD-Biota when only sediment data are available.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Reassessment of Dose Contributions from the M-Area Glass Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater (GW) and inadvertent human intruder (IHI) screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in U-235 being brought back into the list of radionuclides requiring inventory limits within the M-Area Glass SWF. In addition, while reviewing the earlier vadose zone (VZ) transport analysis efforts, an input deck error was observed in the U-234G VZ transport simulation.

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

Vision 0.3.2 Users Guide: Revision 0

Vision is a research software application developed by Pacific Northwest National Laboratory (PNNL) that incorporates the scientific and operational expertise for reviewing data from treaty monitoring radionuclide stations. These stations are part of a worldwide network to monitor for nuclear explosions, and the data they produce are critical to make the determination of whether a sample is from a nuclear explosion or some other source (i.e., nuclear reactor or medical isotope production facility). Stations deliver their measurements and system status to the International Monitoring System (IMS), which forwards it via email to all subscribers. Vision provides tools for viewing and analyzing these stations’ state of health data.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Updates to the Hanford Soil Inventory Model (SIM Version 2) for FY 2025

Update SIM to newest version (SIMv2.3). SIM is being updated to reflect the updates to the tank leak inventories performed in ECF-HANFORD-24-0061. This update is also being done to publish the site inventories in both 2001 decayed and back-decayed formats. This is particularly needed for modeling that requires inventory data from before 2001 that includes decay information for radionuclides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Accident Progression and Source Term Demonstration Calculations for the Advanced Burner Test Reactor (ABTR)

As advanced reactor designers approach broad deployment of new reactor technologies and the associated applications, a critical milestone on this path includes licensing or authorization of these designs in the current regulatory environment. Commercial licensing and DOE authorization both consist of numerous aspects to not only certify the designed plant will operate as intended, but also related to confirmation of the safe operation of these facilities that comply with specific standards and requirements. To that end, domestic industry and regulators require toolsets and methods which can be used to confirm that advanced reactor designs not only align with expected performance during normal operation, but also a series of transient scenarios. These codes require a satisfactory validation basis, and in the absence of such a basis, code-to-code comparisons are often utilized as a surrogate to confirm model adequacy. To support this need for fast reactor technologies, a series of analyses using SAS4A/SASSYS- 1 and the Simplified Radionuclide Transport (SRT) code have been conducted for the Advanced Burner Test Reactor (ABTR). These analyses are expected to not only provide a baseline for transient and source term analyses of ABTR, but to also provide context for the key physics that determine evolution of specific transient scenarios in a sodium-cooled fast reactor (SFR). These codes are utilized for this work given their adoption in the SFR industry and by key SFR stakeholders.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Units 101 and 102: Central and Western Pahute Mesa Nevada National Security Site, Nevada, Revision 2

This revision was prepared to document changes to the data-collection activities. Specifically, the drilling and completion of the wells have been revised, and enhancements to the surface geophysics measurements have been included. Corrective action is needed because underground radioactive contamination exists in these areas resulting from historic underground nuclear testing. The PM CAUs are located in the northwestern portion of the Nevada National Security Site and comprise 82 corrective action sites. A total of 85 underground nuclear detonations took place within the area covered by the PM CAUs between 1965 and 1992 and resulted in the release of radionuclides in the subsurface in the vicinity of the test cavities.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

TA-54 Flanged Tritium Waste Container Operations, Radioactive Air Emissions Summary Volume 1: Stack Emissions & Off-Site Dose Consequence

In September 2025, pressure mitigation operations began on the four Flanged Tritium Waste Containers (FTWCs) that had been in storage at Los Alamos National Laboratory (LANL), Technical Area (TA-) 54, Area G, Building 1028. Each FTWC was opened at TA-54 and put into a stable configuration, then containers were moved to LANL’s Weapons Engineering Tritium Facility (WETF) for further processing in preparation for shipment off-site. Over the course of 13 operational days, the project had air emissions of 123 curies of tritium (hydrogen-3). The radiological dose to the hypothetical maximally exposed individual (MEI) member of the public from these emissions was 0.012 millirem. This is well within the limits established by the Environmental Protection Agency (EPA), which allows 10 millirem per year to any member of the public. Naturally occurring background radiation in Northern New Mexico averages about 300-400 millirem per year to residents. This report is intended to briefly summarize operations that took place in September and October 2025, provide documentation of the radionuclide air emissions from these operations, and summarize the off-site dose consequence to the public that resulted from these air emissions. Informal notifications addressing these areas were made via email to the EPA and NMED throughout the project, and summaries of these data were also posted on the LANL external web site 1 for FTWC operations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Iodine Capture Studies of Copper- and Bismuth-Based Sorbents

The release of radioiodine, one of several radionuclides of concern when recycling used nuclear fuel (UNF), is an important consideration in the fuel cycle. In this study, two sorbent materials, Cu 0 -polyacrylonitrile (PAN) and Bi 0 -PAN, were tested as solid sorbent candidates for iodine capture. Experiments using a thin bed of sorbent material were exposed to vaporized iodine for over 300 hours (~2 weeks) under varied conditions in a dynamic flow environment. The overall performance was monitored in real-time using thermogravimetric analysis, and the materials were subsequently characterized for surface and bulk analysis using scanning electron microscopy – energy-dispersive spectroscopy (SEM-EDS) and powder x-ray diffraction (pXRD), respectively. Iodine (in the form of I 2 ) is expected to be released primarily in the dissolver off-gas (DOG) stream; therefore, this study demonstrates the effects of elemental iodine (I 2 ), water vapor (H 2 O), and nitrogen dioxide (NO 2 ). When exposed to ‘ideal’ conditions (in which I 2 is carried by dry air), Cu 0 -PAN and Bi 0 -PAN behave differently, with TGA analysis indicating that Cu 0 sorption performance is higher than that of Bi 0 , as evidenced by a larger mass change. Under ‘harsh’ conditions—such as a gaseous feed containing I 2 , H 2 O, and NO 2 vapors carried by air,—iodine capture performance for both Cu 0 - and Bi 0 -PAN are affected. SEM-EDS and pXRD analysis of these materials is discussed herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Integrated Off-gas System: A Preconceptual Design of an Integrated Off-Gas Treatment System

The U.S. has a vested interest in the advancement of nuclear energy to achieve aggressive net-zero goals, with reprocessing and recycling of used nuclear fuels (UNF) playing a vital role. It will not be possible to meet U.S. regulatory requirements without robust off-gas treatment, so it is crucial to advance treatment technologies to facilitate the design of future reprocessing facilities. For many years, teams of researchers across the U.S. Department of Energy (DOE) National Laboratory complex have been investigating off-gas treatment technologies for the capture and removal of volatile radionuclides (i.e., 85 Kr, Xe, 14 C, and 129 I) and oxides of nitrogen (NO X ) that are produced from reprocessing. These investigations have been focused on developing individual technologies for the capture of Kr, Xe, iodine, and CO 2 . Capture technologies for each constituent were tested independently from one another by utilizing nonradioactive surrogates to simulate simplified off-gas streams. The tests have been relatively small, laboratory-scale experiments of up to approximately 1 L/minute total gas flow rate. To increase the readiness of these technologies for deployment, an integrated test system with a larger-scale capacity is needed to bridge the gap between promising bench scale and fully scalable UNF reprocessing off-gas treatment. This document contains the goals, design basis, functional requirements, preconceptual design, and cost estimates for an integrated off-gas demonstration system for the capture and removal of NO x , Kr, Xe, CO 2 , and iodine at 10× higher throughput than earlier laboratory studies. The order-of-magnitude cost estimate for this system is approximately $\$$886,000. Next phases include conceptual design, detailed design, fabrication, and commissioning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗