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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Subject-specific modeling framework for particle deposition using computational fluid dynamics

Quantifying particle deposition and dose in the respiratory tract requires a physiologically realistic representation and reproducible computational workflows. However, existing modeling frameworks, such as the International Commission on Radiological Protection (ICRP) compartmental models and the Multiple Path Particle Dosimetry (MPPD) tool, lack detailed deposition profiles and subject-specific capabilities. The combination of advances in computer vision algorithms applied to the respiratory tract and Computational Fluid and Particle Dynamics (CFPD) allows high-fidelity simulations of particle behavior in anatomically accurate geometries derived from individual CT scans. The segmentation, preprocessing, and file preparation task for a CFPD simulation was often time-consuming, and no prior studies to-date have yet presented a fully automated framework. This work presents a fully automated workflow to obtain individualized particle deposition profiles in the human respiratory tract. The pipeline starts with segmenting upper and lower airway geometries using morphological and deep learning-based methods, generating three-dimensional (3D) models from CT imaging data. Next, a series of algorithms are presented to quality check and prepare the 3D geometry for a CFD or CFPD simulation. The preprocessing step includes correcting geometric artifacts, enforcing a physically consistent mesh, and automatically identifying and capping multiple outlets, which is required for CFD/CFPD simulations. These processed models are then input into open-source (OpenFOAM) or commercial (StarCCM+) CFD solvers, where flow and transient particle transport equations — including turbulence and particle–wall interactions are solved under realistic breathing conditions. Finally, the resulting particle deposition profiles can be integrated with Monte Carlo radiation transport codes and state-of-the-art computational phantoms to assess organ-specific absorbed doses in scenarios of radioactive aerosol inhalation. The presented work streamlines respiratory tract segmentation, preprocessing for CFD/CFPD simulations, and integration with dose assessment workflows, reducing manual intervention and improving access to high-fidelity, subject-specific modeling. The high precision in predicted particle deposition and dose distributions can improve personalized treatment strategies in respiratory medicine and refine dose estimates for radiation protection.

AI↗
DOE OSTI2022

Megahertz pulse trains enable multi-hit serial femtosecond crystallography experiments at X-ray free electron lasers

The European X-ray Free Electron Laser (XFEL) and Linac Coherent Light Source (LCLS) II are extremely intense sources of X-rays capable of generating Serial Femtosecond Crystallography (SFX) data at megahertz (MHz) repetition rates. Previous work has shown that it is possible to use consecutive X-ray pulses to collect diffraction patterns from individual crystals. Here, we exploit the MHz pulse structure of the European XFEL to obtain two complete datasets from the same lysozyme crystal, first hit and the second hit, before it exits the beam. The two datasets, separated by <1 µs, yield up to 2.1 Å resolution structures. Comparisons between the two structures reveal no indications of radiation damage or significant changes within the active site, consistent with the calculated dose estimates. This demonstrates MHz SFX can be used as a tool for tracking sub-microsecond structural changes in individual single crystals, a technique we refer to as multi-hit SFX.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Probing soft X-ray induced photoreduction of a model Mn-complex at cryogenic conditions

Soft X-ray absorption spectroscopy of first row transition elements at their respective L -edges provides important information about the oxidation and spin states of the metal centers. However, the associated sample damage in radiation-sensitive samples substantially alters the electronic and chemical structures of redox-active metal centers. Here, we measure the soft X-ray spectrum of the model Mn III (acac) 3 complex containing a redox-active Mn III metal center in an octahedral environment with a superconducting transition-edge sensor detector. To reduce the secondary damage resulting primarily from the diffusion of radicals and electrons, the spectra are collected at 30 K and 80 K on solid samples. Starting from the first scan, we detect the contribution of X-ray induced sample damage leading to a change in the Mn II intensity. However, at low temperatures, particularly at 30 K, we do not observe a gradual increase in the radiation damage with successive scans with the X-ray beam at the same spot. At our estimated dose of 90 kGy, we find 62% of Mn III (acac) 3 is still intact at 30 K. However, at room temperature, we see a gradual increase in radiation damage with increasing numbers of scans at the same spot, which is consistent with the possibility of increased diffusion rates of secondary radicals and electrons as noted in other studies.

L-edge spectroscopy↗
DOE OSTI2021

238 PuO 2 In Vitro Lung Dissolution Rate and Particle Size Determination for Material Involved in the June 8, 2020 Incident at the Los Alamos National Laboratory PF-4 Facility

As part of the LANL response to the June 8, 2020 238 PuO 2 inhalation exposure incident in PF-4, the Nuclear and Radiochemistry Group (C-NR) was asked to assist in determining the lung dissolution rate and particle size distribution of the airborne material. These material characteristics, along with urine and fecal bioassay data, are used by the Radiation Protection Services (RP-SVS) internal dosimetrists to refine dose estimates for exposed workers. Similar to previous studies, an in vitro lung dissolution rate study was performed to assess how fast 238 Pu is cleared from the lungs. Given that PuO 2 is a relatively insoluble material, it dissolves very slowly and requires in vitro studies of 100 or more days to estimate dissolution rates. This report details the results of a 100 day in vitro dissolution rate study and the particle size characteristics of 238 PuO 2 collected on an air filters during the incident. Results show that the 238 PuO 2 involved in this incident has behavior comparable to PuO 2 from previous in vitro studies, with approximately 1% of material dissolving quickly, and 99% of the material dissolving very slowly, with a half-time of approximately 120 years. This is somewhat slower than what was determined in a previous study of LANL 238 PuO 2 , which had an experimental dissolution half-time of 3 years, but significantly faster than 3000 years measured for 239 PuO 2 . Particle size distribution measurements indicate that particle sizes range from 0.10 µm – 6 µm, with most < 2 µm, which is reasonable based on our understanding of the process history of the 238 PuO 2 . Particle size, specific surface area, calcining temperature, specific activity, and age all are known factors that can impact the dissolution rate of refractory materials, however, there are too few studies to quantify the exact influence of these parameters on a measured dissolution rate. The results of this study add to that knowledge base, and provide new insights into the dissolution rates for this LANL process-specific 238 PuO 2 material.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2022

Hanford Site Regional Population – 2020 Census

The U.S. Department of Energy conducts radiological operations in south-central Washington State. Population dose estimates must be performed to provide a measure of the impact from site radiological releases. Results of the U.S. 2020 Census were used to determine counts and distributions for the residential population located within 50 miles (80 kilometers) of several operating areas of the Hanford Site. Year 2010 was the first census year that a 50-mile population of a Hanford Site operational area exceeded the half-million mark. All five locations evaluated in this report for the year 2020 census exceeded the half-million mark for 50-mile populations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Pacific Northwest National Laboratory Regional Populations – 2020 Census: Richland Campus and Sequim Campus

The U.S. Department of Energy conducts radiological operations in south-central (Richland) and northwestern (Sequim) Washington State. Collective dose estimates to regional populations must be performed to provide a measure of the impact from site radiological releases. Results of the U.S. 2020 Census and, for the Sequim location only, the Canada 2021 Census, were used to determine counts and distributions for the residential population located within 50 miles (80 kilometers) of the Pacific Northwest National Laboratory Richland Campus and the Sequim Campus. Since the 2010 Census, total regional populations increased 25–40 percent. This revision updates and expands the data presented in the prior report by the addition of PNNL-Richland Campus population distribution information.

2020 census↗
DOE OSTI2024

Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Summary Report - FY 23)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. All the trenches in these burial grounds except two have remained inactive since 2005. The only active trenches that continue to receive low-level and mixed low-level waste are Trenches 31 and 34 that are located in the southern portion of the 218-W-5 Burial Ground. During this reporting period (fiscal year 2023, from October 1, 2022, through September 30, 2023), waste was disposed in these two trenches.

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

Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds (Annual Summary Report - FY 23)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. All trenches in these burial grounds except one have remained inactive since 2005. Trench 94 located in 218-E-12B Burial Ground is the only active trench that is dedicated for disposal of U.S. Navy reactor compartments. During this reporting period (fiscal year 2023, from October 1, 2022, through September 30, 2023), no reactor compartments were disposed in Trench 94.

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

X-ray Security Imaging on Personal Dosimetry [Slides]

Federal regulations mandate that personal dosimetry devices — like optically stimulated luminescence (OSLs) — must be worn by all US Department of Energy (DOE) and associated radiation workers to track their occupational dose. Unfortunately, the inadvertent passage of OSLs through x-ray security scanners can compromise their validity. With the advent of high energy, advanced resolution security technology used in airports, this once insignificant issue now requires that Radiological Control (RadCon) be able to accurately discern non-occupational dose to effected OSLs. This presentation will discuss the principles, methods, and, rather surprising, models for establishing the corrective dose estimates at the Idaho National Laboratory (INL), and its implication across the DOE.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Annual Summary Report (FY 2025) Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. All the trenches in these burial grounds except two have remained inactive since July 2004. The only active trenches that continue to receive low-level and mixed low-level waste are Trenches 31 and 34 that are located in the southern portion of the 218-W-5 Burial Ground. During this reporting period (fiscal year 2025, from October 1, 2024, through September 30, 2025), waste was disposed in Trench 31 only.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Annual Summary Report (FY 2025) Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low Level Burial Grounds (LLBGs) since September 26, 1988. All trenches in these burial grounds except one have remained inactive since July 2004. Trench 94 located in 218-E-12B Burial Ground is the only active trench, which is dedicated for disposal of U.S. Navy reactor compartments. During this reporting period (fiscal year 2025, from October 1, 2024, through September 30, 2025), two reactor compartments were disposed in Trench 94.

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

Examination Of IASCC In 304 Stainless Steel Core Shroud From A Commercial Boiling Water Reactor (BWR)

The 304 stainless steel core shroud and the accompanying weld material from a commercial boiling-water reactor (BWR) experienced an estimated dose of ~3.5 dpa neutron irradiation leading to irradiation-assisted stress corrosion cracking (IASCC). The resulting IASCC crack tips are examined with high-resolution analysis techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT), with select examples of directly correlated STEM/APT observations. Analytical observations reveal the crack morphology, composition, microstructure, and any associated corrosion-induced or radiation-induced segregation in the local metal grain boundaries ahead of the propagating SCC cracks. The observed oxide composition and structure were in-line with expectations, consisting of a Cr-rich inner oxide and a Fe-rich outer oxide, with the leading crack tip oxide being Cr-rich. In addition, the leading grain boundary exhibited typical radiation-induced segregation signatures that were largely unperturbed by the accompanying corrosion. Finally, the quantitative capabilities and limitations of STEM and APT in the context of corrosion fronts and grain boundary chemistry from the directly correlated measurements are discussed.

Olszta, Matthew J.↗
DOE OSTI2025

NUTRON NESHAPs Dashboard

My project aims to enhance the Nuclear Material Tracking Application (NUTRON) software by integrating a tool for the National Emission Standards for Hazardous Air Pollutants (NESHAP) emissions calculations and presenting this data on a dashboard. Inefficiencies in the current NESHAP calculation process were addressed, which will result in more timely regulatory compliance efforts. Key improvements include revising the transfer request system, incorporating effective dose calculations, and developing a data visualization dashboard into NUTRON. The project involved creating a wireframe, preparing an Engineering Calculations and Analysis Report (ECAR), stakeholder meetings, and providing supplemental documentation. Key findings indicate that the proposed modifications will streamline the NESHAP calculation process, reduce human error, and provide REC personnel with accurate and timely data for material use determinations and dose estimations. Future work focuses on completing the NUTRON modifications and fully integrating the new features, ensuring a more efficient and reliable system for tracking and reporting nuclear material transfers.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Preliminary Criticality and Radiation Shielding Analysis for the Storage and Transfer of MARVEL Reactor Spent Nuclear Fuel

This report documents the results of preliminary nuclear criticality and radiation shielding assessments during transfer and dry storage of irradiated Microreactor Applications Research Validation and Evaluation (MARVEL) reactor fuel at Idaho National Laboratory (INL). The assessments focus on transfer casks and storage canisters that are currently in use at INL, which may be compatible with the irradiated MARVEL reactor fuel. The criticality assessments were performed using the radiation transport code MCNP6 with 37 MARVEL reactor fuel elements in various configurations and scenarios. All transfer and storage configurations under dry conditions were below the assumed criticality safety limit of 0.93. Some storage and transfer configurations under wet conditions exceeded the criticality safety limit. This suggests that the appropriate administrative and engineering controls, in addition to reducing the number of MARVEL reactor fuel elements per transfer cask or storage canister, can be expected to ensure criticality safety under all scenarios. The radiation shielding assessments were performed by generating conservative neutron and photon source terms using the ORIGEN module in the SCALE suite of codes. These source spectra were used to estimate the dose equivalent rates using the radiation transport code MCNP6, both on contact and 1 m away from the fuel and transfer casks. The maximum estimated dose equivalent rate of 37 unshielded MARVEL reactor fuel elements on contact is approximately 42000 R/hr. The maximum estimated dose equivalent rates on contact to the ATR transfer cask, HFEF-5 transfer cask, and high load charger were approximately 233 mR/hr, 171 mR/hr, and 201 mR/hr, respectively. This suggests that with the appropriate administrative and engineering controls, all three transfer casks analyzed can be expected to provide sufficient radiation shielding to workers during transfer of irradiated MARVEL reactor fuel. These calculations are performed to support the planning and strategy for the MARVEL project and will demonstrate the technical viability of the different configurations discussed and help identify where engineered or administrative controls may be necessary. A complete criticality safety analysis and radiation shielding analysis, including validation and contingency and accident analysis must be completed by licensed and authorized personnel before any transfer or storage of MARVEL reactor nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Estimate of Gamma Dose Rates from Arrays of Fermi-1 Blanket Elements During the MEDE Process

The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.

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

Mortality among Tennessee Eastman Corporation (TEC) uranium processing workers, 1943–2019

Background: There are few occupational studies of women exposed to ionizing radiation. During World War II, the Tennessee Eastman Corporation (TEC) operated an electromagnetic field separation facility of 1152 calutrons to obtain enriched uranium (235U) used for the Hiroshima atomic bomb. Thousands of women were involved in these operations. Materials and methods: We report a new study was conducted of 13,951 women and 12,699 men employed at TEC between 1943 and 1947 for at least 90 days. Comprehensive dose reconstruction techniques were used to estimate lung doses from the inhalation of uranium dust based on airborne measurements. Vital status through 2018/2019 was obtained from the National Death Index, Social Security Death Index, Tennessee death records and online public record databases. Analyses included standardized mortality ratios (SMRs) and Cox proportional hazards models. Results: Most workers were hourly (77.7%), white (95.6%), born before 1920 (58.3%), worked in dusty environments (57.0%), and had died (94.9%). Vital status was confirmed for 97.4% of the workers. Women were younger than men when first employed: mean ages 25.0 years and 33.0 years, respectively. The estimated mean absorbed dose to the lung was 32.7 mGy (max 1048 mGy) for women and 18.9 mGy (max 501 mGy) for men. The mean dose to thoracic lymph nodes (TLNs) was 127 mGy. Statistically significant SMRs were observed for lung cancer (SMR 1.25; 95% CI 1.19, 1.31; n = 1654), nonmalignant respiratory diseases (NMRDs) (1.23; 95% CI 1.19, 1.28; n = 2585), and cerebrovascular disease (CeVD) (1.13; 95% CI 1.08, 1.18; n = 1945). For lung cancer, the excess relative rate (ERR) at 100 mGy (95% CI) was 0.01 (–0.10, 0.12; n = 652) among women, and –0.15 (–0.38, 0.07; n = 1002) among men based on a preferred model for men with lung doses <300 mGy. NMRD and non-Hodgkin lymphoma were not associated with estimated absorbed dose to the lung or TLN. Conclusions: There was little evidence that radiation increased the risk of lung cancer, suggesting that inhalation of uranium dust and the associated high-LET alpha particle exposure to lung tissue experienced over a few years is less effective in causing lung cancer than other types of exposures. There was no statistically significant difference in the lung cancer risk estimates between men and women. The elevation of certain causes of death such as CeVD is unexplained and will require additional scrutiny of workplace or lifestyle factors given that radiation is an unlikely contributor since only the lung and lymph nodes received appreciable dose.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2022

Dose Coefficients for Internal Dose Assessments for Exposure to Radioactive Fallout

This paper presents values as well as the bases for calculating internal dose coefficients suitable for estimating organ doses from the exposure to radioactive fallout that could result from the detonation of a nuclear fission device. The 34 radionuclides discussed are the same as those given in a priority list of radionuclides for fallout dose assessments presented in a companion overview paper. The radionuclides discussed are those that are believed to account for a preponderance of the organ doses that might be received by intake by persons of all ages (including in utero and via breast feeding for infants) following exposure to radioactive fallout. The presented dose coefficients for ingestion account for age and include modifications for variations in solubility with distance as discussed previously in the literature, and those for inhalation similarly account for age, solubility, and particle sizes that would be relevant at various distances of exposure as discussed in a companion paper on ingestion dose methods. The proposed modifications peculiar to radioactive fallout account for systematic changes in solubility and particle sizes with distance from the site of detonation, termed here as the region of “local fallout” and the region “beyond local fallout.” Brief definitions of these regions are provided here with more detailed discussion in a companion paper on estimating deposition of fallout radionuclides. This paper provides the dose coefficients for ingestion and inhalation (for particle sizes of 1 μm, 5 μm, 10 μm, and 20 μm) for the region “local fallout.” These dose coefficients for “local fallout” are specific for particles formed in a nuclear explosion that can be large and have radionuclides, particularly the more refractory ones, distributed throughout the volume where the radionuclide has reduced solubility. The dose coefficients for the region “beyond local fallout” are assumed to be the ones published by the International Commission on Radiological Protection (ICRP) in 1995. Comparisons of the presented dose coefficients are made with values published by the ICRP.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗