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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Calculation of Groundwater Pathway Radiological Dose for the Hanford Site Composite Analysis Base Case

The purpose of this environmental calculation file (ECF) is to present the results of the exposure route-specific and total radiological dose assessments for the groundwater exposure pathway as a part of the updated Hanford Site Composite Analysis (CA). The purpose of these radiological dose assessments is to provide an estimate of the cumulative radiological impacts from all screened sources of ionizing radiation and exposure routes that could potentially contribute to the projected dose to a hypothetical member of the public from both existing or future disposal facilities and other sources including past-practice discharge sites.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Calculation of Dangerous Values for Radionuclides Considered by the IAEA Code of Conduct

The D-value or dangerous quantity system was designed by the International Commission for Radiological Protection for the determination of source protection categories that can be used to reduce the likelihood of accidents, the consequences of which could result in harm to individuals or costly or expensive cleanup. The process includes multiple scenarios for exposure and two different approaches to the evaluation of detriment. This document provides an example calculation using 137 Cs to walk through the complex process of determining its D-value in the hopes of making the process easily understandable.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

FY24 accomplishments in preparation for startup of Activated Materials Laboratory (AML)

The Activated Materials Laboratory (AML) will be a new radiological facility at the Advanced Photon Source (APS) in Argonne National Laboratory (ANL), adjacent to the high-energy x-ray microscopy (HEXM) beamline in the long beamline building (LBB) constructed under the APS-upgrade (APS-U) project. The AML is a centralized facility to facilitate the safe conduct of experiments on activated materials at the APS. This report provides an overview of the status of the AML as a Nuclear Science User Facilities (NSUF) partner user facility in preparation for the general user access in 2025 upon the commissioning of the APS-U beamlines. The AML's scope, functionality and components are detailed. The NSUF partner beamlines’ commissioning status in the post-APS-U era is provided, along with the AML’s operational updates and operational plan.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing iron-chromium-aluminum (FeCrAl) alloys in place of zirconium-based alloys in major core structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement

To extend NUREG-1465 and high burnup fuel source term (SAND2023-01313) recommendations, representative radiological releases to containment – patterned after NUREG-1465 – have been evaluated for LWRs utilizing the chromium-coating on major zircaloy structures (cladding and fuel canisters) and high burnup fuel with enrichments of 8% and 10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated for this accident tolerant fuel concept by applying non-parametric bootstrap methods to MELCOR simulation results. Accident scenarios considered in this analysis include principle contributors to historical core damage frequency estimates for a range of nuclear reactor technologies representative of the operating U.S.A. fleet of nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development and testing of a fission-like neutron field for dosimetry and instrument calibrations based on deuterium-tritium generator

For nearly five decades 252 Cf sources have been used for testing and calibration over a wide range of neutron detection devices used for nuclear safety and radiation protection. However, the vastly increased cost of 252 Cf sources, its short half-life, and concerns about future shortage have prompted nuclear facilities and radiological calibration laboratories that rely on high-intensity neutron sources to seek alternatives. There appear to be no other radionuclide neutron sources that would match “like-to-like” 252 Cf performance characteristics. Alternatives of producing a fission-like neutron spectrum based on commercially available deuterium-tritium (D-T) generators, and without any fissionable materials or beryllium, have been explored. This paper describes efforts of designing the spectrum shaping assembly, its construction and installation. The produced neutron field was evaluated both from the perspective of the neutron fluence spectrum and the resulting dosimetric outcome. Finally, it was tested for practical application by examining response of various health physics instruments and personal monitoring devices in comparison to the response of those devices when exposed to 252 Cf. In conclusion, the response of survey-type instruments tested within the prototype, surrogate fission, reference field are acceptably close to their response in the targeted 252 Cf field.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

A Fission-like Neutron Spectrum Shaping Assembly based on D-T Generator

For nearly five decades, 252 Cf sources have been used for testing and calibration over a wide range of neutron detection devices used for nuclear safety and radiation protection. However, the vastly increased cost of 252 Cf sources, its short half-life, and concerns about future shortage have prompted nuclear facilities and radiological calibration laboratories that rely on high-intensity neutron sources to seek alternatives. While there appear to be no other radionuclide neutron sources that would match the 252 Cf performance characteristics, all of them, including 252 Cf, present similar concerns in terms of hazards, safety, and security. As a possible solution to this issue, the Pacific Northwest National Laboratory (PNNL) developed a low-risk, cost-efficient alternative to produce fission-like spectra based on commercially available deuterium-tritium (D-T) neutron generators. The alternative of producing a fission-like neutron spectrum without any fissionable materials or beryllium would enable the nuclear facilities across the entire complex to eliminate their reliance on high intensity 252 Cf sources for nuclear safety and radiation protection applications, substantially reducing risks and associated costs. This report describes efforts of designing the spectrum shaping assembly, its construction and characterization of the neutron field produced. Performance of this newly developed D-T generator based neutron field is demonstrated via calibration of a criticality detector, neutron survey instruments, and personal dosimeters verified against current standardized fields produced by 252 Cf spontaneous fission sources.

42 ENGINEERING↗
DOE OSTI2023

Python Urban Deployment Model (PyUDM) v1.0.0

The Python Urban Deployment Model (PyUDM) is a simulation tool used to investigate networks of static radiation detectors in urban environments. PyUDM simulates traffic, stationary NaI gamma-ray detectors, and moving radioactive sources on simulated vehicles. The analyzed output of the simulations contain valuable insights into the performance of different configurations of urban radiological detector configurations such as their ability to detect sources moving through the environment. To accurately simulate radioactive material moving through urban environments, PyUDM combines Monte Carlo simulation tools, publicly available map and traffic data, and measured gamma-ray spectra from urban environments.

Rofors, Emil↗
DOE OSTI2023

Review of particle deposition to and removal from clothing, skin, and hair after a radioactive airborne dispersal event

Explosive Radiological Dispersal Devices (RDD) – aka dirty bombs – are seen as a credible method to carry out a radiological terror attack. After exploding a radioactive source, the radionuclide-laden plume will be blown downwind of ground zero, with particles falling out and potentially depositing on people caught in and under the cloud. Some of these people may not show any sign of radiation sickness and therefore not realize they have been contaminated and may take the radioactive particulate with them on their daily activities, thus spreading the radioactive particulate outside the initially contaminated area. This paper reviews the scientific literature to better understand the rate at which particulate deposits on and is removed from the different “surfaces” of a person, i.e., hair, skin, and clothing. Prior research indicates that: 1) particle deposition is usually higher on skin than on hair and clothing; 2) particle deposition is greater for a person with higher skin moisture, 3) stronger wind increases the deposition flux onto a person, and 4) the fraction of particulate deposited on the hair, skin, and clothing respectively depends on the length of the hair, assuming all the hair surface is available for deposition. The studies taken into consideration show that the largest uncertainty in particulate deposition onto a person is due to clothing type because of the different possible weave arrangements and tightness which translate into differences in actual surface area and surface roughness. A factor of 2-to-20 variation in deposition rate was found. Removal of the particulate from the contaminated person may be due to wind, a person's movement, and/or contact transfer, i.e., by touching a different clean surface. Experiments show that the majority of the particulate is resuspended within 2–6 h mostly depending on the intensity of physical activity. The largest uncertainty in particulate removal from skin depends on the skin moisture, transfer rate of single-contact, and how many objects/people a person touches per hour. No data for hair were found for particle removal and resuspension. The studies considered did not utilize radionuclides directly; however, data on adhesion of radioactive vs. their non-radioactive counterpart have shown that the uncertainty due to the radioactivity of the particles is lower than that due to other factors. In conclusion, an idealized scenario involving a single building in the path of the cloud showed the impact of building-influenced flow on the cloud transport path and mixing, which affects the radiological dose the downwind population is exposed to and consequently the health effects.

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

Perspectives on Response to an Abnormal Radiological Event [Slides]

On May 2, 2019, a sealed radioactive source containing cesium-137 was breached at the University of Washington Harborview Research and Training (HRT) Building in downtown Seattle while attempting to recover the source for the National Nuclear Security Administration’s (NNSA’s) Off-site Source Recovery Program (OSRP). The breach of the source resulted in the contamination of personnel, large sections of the building, and a release of material to the local environment.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Threat Sources for Creating Synthetic Urban Search Data

Equivalent point source energy emission distributions were computed for various threat sources for use in simulating the detector responses for urban search scenarios. The sources include standard isotopic sources used in detector testing, medical and industrial sources occasionally encountered in urban searches, and several types of special nuclear materials. Most of the equivalent point source distributions represent small sources inside some amount of shielding, but the special nuclear material sources represent volumetrically distributed sources in spheres of metal. Text-based inputs for emission distributions are available for the Monte Carlo transport codes Monte Carlo N-Particle, SCALE/MAVRIC, and Omnibus/Shift, any of which can easily be converted to other formats. These sources were developed for use in the Radiological Anomaly Detection and Identification (RADAI) project and the follow-on project, the RADAI-Extended project, sponsored by the National Nuclear Security Administration Office of Defense Nuclear Nonproliferation Research and Development.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Evaluating and Countering the Insider Threat to the Radioactive Source Supply Chain

The modern supply chain is a global enterprise and little drove this home more than the global COVID-19 pandemic which sent economic shockwaves throughout the world. Many goods became scarce, as products were delayed, in limited supply, or simply not available. The suddenly diminished supply collided with still high demand and led to greatly increased costs. This was particularly true for the radioactive source supply chain. The pandemic introduced extensive delays for construction projects, slowed the transport of radiological materials to facilities, interrupted treatment deliveries, and impaired the mobility of contractors across the industry. All of these concerns not only adversely affected the economy, but also impacted the safety and security of radiological material, potentially raising national security concerns. The vulnerability of the supply chain, a critical element in an increasingly interconnected world, was exposed. One example that challenged the adaptive capacity of the overall supply chain is the Ever Given container ship, which became stuck in the Suez Canal in 2020. This accident immediately shut down shipments that accounted for 12% of global trade, with long-term impacts estimated to be much larger. Developing the ability to anticipate and react in real-time to sudden changes has quickly become a necessity, particularly in industries that deal with the transport of hazardous material. The reaction to these dramatic incidents was to largely focus attention and resources on protecting the supply chain from external threats. However, the threat to the radioactive material supply chain from insiders intimately involved in the process may be even greater and remains a blind spot that requires increased attention. Recent events revealed the blueprint for targeting and disrupting that supply chain, so the potential for a malicious insider—or a manipulated, unwitting insider—to take advantage of this vulnerability is elevated, creating security concerns for radiological industries. This paper examines and analyzes the potential insider threat to the radioactive source supply chain and recommends steps to take to counter this possibility.

Kinney, Justin↗
DOE OSTI2025

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. Here, we propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Tritium Safety: Self-Study Course 11952

Course Overview: Tritium is a radioactive isotope of hydrogen which has characteristics that require workers understand what should be done to work safely and responsibly around it. Tritium is used for defense applications, experimental purposes for general research, and is a byproduct of some of the work that is done at LANL. Practical guidance offered in this training will inform safe handling to minimize negative impacts on personnel, the workspace, and the surrounding environment. This course, Tritium Safety Self-Study (Course 11952), provides information on the physical, chemical, and radioactive properties of tritium; sources of tritium; uses of tritium; biological effects of tritium; radiological control methods for tritium; monitoring methods of tritium; tritium waste minimization and handling; and alarms and proper response to abnormal conditions in a tritium facility. Course Objectives: When you have completed this course, you will be able to: Identify the physical, chemical, and radioactive properties of tritium; Identify the sources of tritium; Identify the uses of tritium; Identify the modes of tritium exposure; Identify the biological effects of tritium; Identify the radiological control methods of tritium; Identify methods used to monitor for tritium; Identify tritium waste minimization and handling techniques; Identify proper response to abnormal conditions in tritium facilities.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Radionuclide-specific Parameters Dataset

The radionuclide-specific parameters dataset is searchable for radiological information for multiple isotopes simultaneously. After selecting radionuclides of interest and the desired parameters, the RAIS will generate a table containing the values, chosen according to an established hierarchy. Results can be downloaded in Excel format. 50 parameters are available, including atomic number, soil to animal transfer coefficients, plant uptake coefficients, half-life, specific activity, and water solubility. Seven primary sources are used to populate the dataset of radiological-specific parameters. These values should be used in cancer risk assessments for the calculation of preliminary remediation goals (PRGs), hazard characterization, and transport modeling. Users can select up to 1000 radionuclides per query. The dataset supports environmental risk assessments, regulatory decision-making, and environmental planning with tools for benchmarking against risk-based standards. This structured approach ensures a robust evaluation of environmental risks tailored to regulatory needs.

Manning, Karessa [Oak Ridge National Laboratory (O↗
DOE OSTI2026

A Mobile Hot Cell for Conditioning Disused Sealed Radioactive Sources for Storage or Transportation

An innovative Mobile Hot Cell (MHC) has been developed for conditioning Disused Sealed Radioactive Sources (DSRS) category 1 and 2 for storage or transportation. The MHC is designed to provide both Radiological and Biological containment with a maximum capacity of 1000 Ci Co60 or 5000Ci Cs137 source and can be transported via standard cargo containers. This project has been supported through the National Nuclear Safety Administration (NNSA) Offsite Source Recovery Program (OSRP). The project is intended for the international community rather than domestic although domestic use is a possibility. Many countries have significant stockpiles of these devices that are often stored in less-than-optimal circumstances. This necessitates that these devices be addressed expeditiously, and the sources secured. The MHC utilizes robotics, automation, and other non-traditional methods for disassembling, characterizing, and packaging these sources that have reached end of life or are otherwise not needed. These innovative approaches are necessary to facilitate an expedited timeline to efficiently and safely secure these sources in a non-proliferation effort. Conditioning efforts include disassembling the device such as a teletherapy head used for cancer treatment, or blood/research irradiators such that the radioactive sources may be removed safely. The sources are then characterized. Leak checks are performed, dimensions are verified, and serial numbers are confirmed. Upon completion, the sources are typically placed into a Standard Forms Capsule which is seal welded closed. It is leak tested and placed into a Long-Term Storage Shield (LTSS) which can either be secured for storage directly or loaded into an appropriate cask for transportation. Further innovations include multiple deployment scenarios that include a full deployment of MHC components, deployment of the MHC automation internal components to an existing hot cell, deployment of minimally required MHC components and incorporation of sand for shielding, and integration of the MHC for Silo Storage, or Bore Hole Storage efforts. The MHC has evolved from a very specific use case to a “Swiss Army Knife” type of a tool in that it can be readily adapted to a large variety of situations. Innovative approaches such as the use of robotics, Computer Numeric Control (CNC) machining centers, automated welding equipment, HDMI Cameras, and LED lighting are some of the developed technologies incorporated into the MHC design. Shielding is accomplished with a steel walled Base Box which is surrounded by four nesting doll shield shells which when combined limits the external dose rate to 5mr/hr when a 1000 Ci Co60 source is exposed inside.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2026

Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses

Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗