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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Uncertainty Analysis of Inhalation Dose Coefficients for Nuclear Incident Response

This study addressed the need to characterize variability in inhalation dose coefficients due to uncertainties in respiratory tract deposition, systemic biokinetics, and physiological parameters. A Python-based implementation of the International Commission on Radiological Protection Publication 66 Human Respiratory Tract Model was developed called the Radiological Exposure Dose Calculator (REDCAL) to propagate parameter uncertainty.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRSDate not supplied

Actively Controlled Louver for Human Spacecraft Radiator Ultraviolet (UV), Dust, and Freeze Protection

This paper examines the use of actively controlled louvers to attenuate UV and dust, as well as mitigate freezing concerns for human spacecraft radiators during Artemis missions. Artemis missions to the lunar orbit or surface will expose the radiators to high energy UV radiation and dust, which will degrade the radiator’s coating emissivity and consequently reduce heat rejection performance. In addition, subfreezing environmental temperatures during transit to lunar orbit and nighttime on lunar south pole can rupture coolant tubes, reduce heat rejection performance, and worst-case scenario result in a Loss of Mission (LOM). Louver technology would be a promising solution to maintaining radiator performance and integrity for Artemis missions, but heritage louvers are passively controlled. This technology needs maturing to active control, or motor actuation, to achieve faster thermal response times. Actively controlled louver design considerations are discussed in this paper. The analysis that follows shows actively controlled louvers can attenuate high energy UV radiation and dust, as well as protect the coolant from freezing.

Louver↗
DOE OSTI2023

Cadmium Magnesium Telluride for Next-Generation X-Ray Free Electron Laser, Synchrotron, and Many Other Applications

We developed a picosecond photodetector based on our Bridgman-grown and specially engineered cadmium magnesium telluride (Cd 1–x MgxTe) single crystal that is sensitive to both optical and X-ray pulses for coarse timing in free-electron laser applications. Cd 1–x MgxTe is a widebandgap semiconductor with potential applications, not only in optoelectronics, but also in particle physics as an intense pulse radiation detector for bremsstrahlung, X-ray/gamma-ray radiation, thermal neutrons, and medical imaging. For femtosecond optical and X-ray crosscorrelation, the material must have a very short lifetime, a condition that is opposite to that required for nuclear spectroscopy applications. At the same time, the material also needs to have a very low bulk leakage current, in the 10–90 nA range for voltages to even 1000 V. Hence, the ability to tailor or engineer the material is very crucial. Picosecond response and the crystal growth of this specially engineered Cd 1–x MgxTe material are presented. Furthermore, other characterization and transient measurements are discussed along with room-temperature semiconductor detector performance for other nuclear radiation detection applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Preliminary Experiences with the Rexon UL-320-FDR: An Automated Thermoluminescent Dosimeter Reader with Removable Contact Heating Planchets and an Infrared Temperature Feedback System

The Rexon UL-320 FDR is a novel resistive-heating thermoluminescent dosimeter reader with a unique temperature measurement system and an automated dosimeter processing mechanism. The removable contact heating planchets have black-body adhesives on the back for capturing temperature information with infrared sensors. A heating cycle feedback loop ensures accurate, precise, and reproducible heating sequences. Heating rates between 0.8 and 40°C s -1 for up to 1,000 s are possible. Photomultiplier tube sensitivity and drift, dark current counts, and planchet glow were measured experimentally. Additionally, 25 LiF:Mg,Ti dosimeters were calibrated to demonstrate reader performance. Sensitivity was optimized at 1,200 V, which produced the highest reference light count to dark current count ratio while extending photomultiplier tube life. Dark current counts measured with typical time-temperature profiles for LiF:Mg,Ti were below 10 counts per channel but increased by up to 2.5% for more extreme heating cycles. Reader sensitivity drifts of up to 10% were observed during extended automated operations with typical time-temperature profiles. Total counts resulting from planchet glow decreased with faster heating rates. Calibrations performed with LiF:Mg,Ti dosimeters yielded results comparable to more established reader designs. Spikes were observed in ~3% of the glow curves from planchet dust and oil burning off at elevated temperatures. The use of N 2 gas and sensitivity drift corrections are recommended to improve dosimetry performance for the UL-320 FDR reader.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Mobile Hot Cell Digital Twin: End-of-life Management of Disused High Activity Radioactive Sources – 23598

Sealed radioactive sources are utilized for a wide range of applications across nuclear facilities, universities, hospitals, and industry. When these sources reach the end of serviceable life, they become waste. As waste, this radioactive material then goes through a process of recapture and then transfer to long term storage. With the advancement of technology in conjunction with better accessibility of technology, industries are exploring the use of digital automation to enhance productivity, efficiency, and safety while minimizing operation and maintenance costs, health and environmental risks, and uncertainty in the project life cycles. One area for exploration is the use of a digital twin to help design a robust, versatile, and safe solution for recapturing spent sources. We believe this avenue can also provide further advantages in the operations aspect of end-of-life management of spent radioactive sources by reducing the deployment time, increasing operator safety, and reducing operation & maintenance cost. We present a novel digital twin framework for end-of-life management of disused high activity radioactive sources. We have designed and developed a framework that houses a digital twin for visualizing and monitoring the recapture process to inform the engineering and design of a new Mobile Hot Cell. Furthermore, we demonstrate the feasibility of the proposed framework by providing a prototypical implementation, supporting the Mobile Hot Cell and human-machine interface's virtual replication.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Mobile Hot Cell Digital Twin: End-of-life Management of Disused High Activity Radioactive Sources

Sealed radioactive sources are used in various settings such as nuclear facilities, universities, hospitals, and industry. When these sources reach the end of their lifespan, they become waste and are recaptured and stored long-term. A digital twin can be used in the design of a solution for recapturing spent sources, offering benefits in deployment time, operator safety, and cost reduction. We propose a digital twin framework for managing the end-of-life process of disused high activity radioactive sources and have created a prototype implementation to demonstrate its feasibility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

User’s Manual for RESRAD-BUILD Code V.4: Vol. 1 – Methodology and Models Used in RESRAD-BUILD Code

The RESRAD-BUILD computer code models radionuclide release and transport in indoor environments and performs pathway analyses to evaluate the potential radiological dose and risk incurred by an individual who works or lives in a building contaminated with radioactive material or housing radioactively contaminated furniture or equipment. The code provides four geometries to characterize a radiation source: point, line, area, and volume, in which radionuclides are homogeneously distributed. Radionuclides contained in a source are considered to be released to the indoor air due to various processes including erosion (mechanically or weathering), diffusion (for tritium and radon in a volume source), or emanation (radon in a point, line, or area source). The release can proceed through different time phases with different rates. In RESRAD-BUILD Version 4.0, a dynamic ventilation model is implemented to simulate the fate and transport of source material particles and radionuclides after their releases. This dynamic ventilation model considers (1) air exchange between rooms in the building and between the rooms and the outdoor environment, (2) deposition from air to floor, (3) resuspension from the floor to the air, and (4) periodical vacuuming that reduces the floor deposition. The fate and transport modeling provides estimates of radionuclide concentrations in the source, in the air, and on the floor at different times, which are then integrated over the exposure duration for the calculation of radiation doses and cancer risks. A single run of the RESRAD-BUILD code can model a building with up to 9 rooms, 10 sources, and 10 receptors. The potential radiation dose and cancer risk incurred by each receptor are calculated for seven exposure pathways: (1) external radiation directly from the sources (accounting for shielding), (2) external radiation from radioactive particles deposited on the floors, (3) external radiation from airborne radionuclides, (4) inhalation of airborne radionuclides, (5) inhalation of radon and radon progenies, (6) inadvertent ingestion of radioactive particles directly from the source, and (7) ingestion of radioactive particles deposited on the floors. Various exposure scenarios can be modeled with RESRAD-BUILD, including but are not limited to, office worker, renovation worker, decontamination worker, building visitor, and resident. Both deterministic and probabilistic analyses can be performed to obtain results in both text reports and graphic displays.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Verification of RESRAD-BUILD Code Version 4

This report documents the verification of the RESRAD-BUILD code, Version 4.0, which was released on December 22, 2022. Two earlier reports verifying Versions 3.0 and 3.1, respectively, were published in 2001 (Kamboj, et al. 2001) and 2003 (Tetra Tech NUS 2003). Version 4.0 of the RESRAD-BUILD code has many new features and modeling enhancements over the earlier versions, including the previously released Version 3.5. Chapter 2 of this report focuses on verifying the external dose and risk modeling for point, line, area, and volume sources, as well as for floor deposition. Besides verification, the external radiation doses calculated by RESRAD-BUILD were also benchmarked with those calculated by the MCNP code (Briemeister 1993). Section J.3 of the RESRAD-BUILD User’s Manual Vol. 1 (Yu et al. 2022) documents the results of that benchmarking effort. Chapter 3 of this report focuses on verifying the ventilation modeling, from checking the remaining source inventory, releases of radionuclides to the air, air concentrations and deposited floor concentrations over time, to the radiation dose and risk associated with inhalation, ingestion, and air submersion, with and without vacuuming. The verification efforts involve designing spreadsheets to perform calculations the same as or like those performed by the RESRAD-BUILD code and then comparing the spreadsheet results with those produced by the code. When the results agree or the differences are within acceptable range, the accuracy of model implementation in the code is verified. In addition to model implementation, the implementation of key functions and features that facilitate the modeling or the use of the code were also verified during the release testing of the code. Appendix A presents the test cases developed for these verification testing, and Appendix B presents the testing results that verify proper implementation of key functions and features.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS2000

Analysis of MIR-18 results for physical and biological dosimetry: radiation shielding effectiveness in LEO

We compare models of radiation transport and biological response to physical and biological dosimetry results from astronauts on the Mir space station. Transport models are shown to be in good agreement with physical measurements and indicate that the ratio of equivalent dose from the Galactic Cosmic Rays (GCR) to protons is about 3/2:1 and that this ratio will increase for exposures to internal organs. Two biological response models are used to compare to the Mir biodosimetry for chromosome aberration in lymphocyte cells; a track-structure model and the linear-quadratic model with linear energy transfer (LET) dependent weighting coefficients. These models are fit to in vitro data for aberration formation in human lymphocytes by photons and charged particles. Both models are found to be in reasonable agreement with data for aberrations in lymphocytes of Mir crew members: however there are differences between the use of LET dependent weighting factors and track structure models for assigning radiation quality factors. The major difference in the models is the increased effectiveness predicted by the track model for low charge and energy ions with LET near 10 keV/micrometers. The results of our calculations indicate that aluminum shielding, although providing important mitigation of the effects of trapped radiation, provides no protective effect from the galactic cosmic rays (GCR) in low-earth orbit (LEO) using either equivalent dose or the number of chromosome aberrations as a measure until about 100 g/cm 2 of material is used.

Flight Experiment↗
NASA NTRS2002

Chromosome aberrations in human lymphocytes induced by 250 MeV protons: effects of dose, dose rate and shielding

Although the space radiation environment consists predominantly of energetic protons, astronauts inside a spacecraft are chronically exposed to both primary particles as well as secondary particles that are generated when the primary particles penetrate the spacecraft shielding. Secondary neutrons and secondary charged particles can have an LET value that is greater than the primary protons and, therefore, produce a higher relative biological effectiveness (RBE). Using the accelerator facility at Loma Linda University, we exposed human lymphocytes in vitro to 250 MeV protons with doses ranging from 0 to 60 cGy at three different dose rates: a low dose rate of 7.5 cGy/h, an intermediate dose rate of 30 cGy/h and a high dose rate of 70 cGy/min. The effect of 15 g/cm2 aluminum shielding on the induction of chromosome aberrations was investigated for each dose rate. After exposure, lymphocytes were incubated in growth medium containing phytohemagglutinin (PHA) and chromosome spreads were collected using a chemical-induced premature chromosome condensation (PCC) technique. Aberrations were analyzed using the fluorescence in situ hybridization (FISH) technique with three different colored chromosome-painting probes. The frequency of reciprocal and complex-type chromosome exchanges were compared in shielded and unshielded samples. c2002 COSPAR. Published by Elsevier Science Ltd. All rights reserved.

Non-NASA Center↗
NASA NTRS2001

Predictions of secondary neutrons and their importance to radiation effects inside the International Space Station

As part of a study funded by NASA MSFC to assess thecontribution of secondary particles in producing radiation damage to optoelectronics devices located on the International Space Station (IS), Monte Carlo calculations have been made to predict secondary spectra vs. shielding inside ISS modules and in electronics boxes attached on the truss (Armstrong and Colborn, 1998). The calculations take into account secondary neutron, proton, and charged pion production from the ambient galactic cosmic-ray (GCR) proton, trapped proton, and neutron albedo environments. Comparisons of the predicted neutron spectra with measurments made on the Mir space station and other spacecraft have also been made (Armstrong and Colborn, 1998). In this paper, some initial results from folding the predicted neutron spectrum inside ISS modules from Armstrong and Colborn (1998) with several types of radiation effects response functions related to electronics damage and astronaut-dose are given. These results provide an estimate of the practical importance of neutrons compared to protons in assessing radiation effects for the ISS. Also, the important neutron energy ranges for producing these effects have been estimated, which provides guidance for onboard neutron measurement requirements.

Neutrons↗
DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Optically stimulated luminescence dosimetry on mobile phone back protective glass

Protective glasses from the back of modern smartphones were tested as potential emergency dosimeters using optically stimulated luminescence (OSL). The glasses were shown to exhibit several favorable dosimetric properties including a strong sensitivity to radiation, a linear dose-response relationship from 0.1 to 8 Gy, a 3% reproducibility, and a minimum detectable dose in the low tens of milligray, even after significant fading. The OSL signal was observed to lose ~78% and ~90% of the original intensity over 1 h and 1 week, respectively; however, a fading curve was developed to correct for any such signal loss. A dose-recovery test demonstrated that reconstructed doses were capable of accuracies within 20%.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

User's Manual for RESRAD-BUILD Code Version 4: Vol. 2 - User's Guide for RESRAD-BUILD Code Version 4

The RESRAD-BUILD computer code is designed to assess radiological doses to individuals who live or work in a building contaminated with radioactive material. The code is equipped with a user-friendly interface that has many features to facilitate using the computer code and understanding the results. The design of the interface provides various options, from entering data and performing calculations to displaying calculation results. General and context-specific help are available, providing information on editing and viewing the radionuclide database, the definitions of input parameters and their use in the calculations, and the selection of the calculation results for placement into other applications. Two types of sensitivity analysis are supported by the code, i.e., deterministic and probabilistic, that can be used to study the influence of input parameters on the calculation results. This user’s guide provides instructions to help users on how to install the RESRAD-BUILD code, navigate the interface, and use the various features to set up a dose/risk analysis and to view/print the results in text and graphical outputs.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Deep Multitask Learning Models for Radiation Estimation at High Energy Accelerator Facility

Controlling the dose of radiation exposure in potential radioactive facilities is critical for ensuring the safety of staff and the public. Here, in this paper, we developed machine learning models to estimate radiation exposure efficiently at the Thomas Jefferson National Accelerator Facility (JLab), aiming to enhance safety in both accelerator facilities and public areas. Multiple sensors were deployed around the three experimental halls at JLab. Data on single-beam currents, energy levels, and radiation values at the sensor locations were collected during accelerator operation. We proposed a multi-task learning model for radiation estimation, utilizing either one-dimensional convolutional neural networks (1-D CNNs) or long short-term memory networks (LSTMs) as the backbone. The proposed model was trained to simultaneously estimate radiation levels at the sensor locations. Experimental results demonstrated that the proposed model with LSTM backbone achieved the best estimation performance, with an average R 2 score of 0.7557 for estimation within the same year and 0.7157 for estimation across different years. These results significantly surpassed those of competing models.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 3, July-September 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS 181 Attitudes and Practices Regarding Disposal of Liquid Nuclear Waste at Clinton Laboratories in the Very Early Years: A Historical Analysis, S. H. Stow; 202 Off-Site Nuclear Emergency Management in Germany Under the Auspices of the Federal Structure, A. Bayer, S. Bittner, and H. Korn; DESIGN FEATURES: 211 Twenty-Fourth DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, R. D. Porco, and M. W. First; OPERATING EXPERIENCES: 222 Explosion in the Tomsk-7 Reprocessing Plant on April 6, 1993, M. L. Hyder, W. G. Lussie, and F. E. Witmer; 234 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 237 Assessment of Spent Fuel Cooling, J. G. Ibarra, W. R. Jones, G. F. Lanik, H. L. Ornstein, and S. V. Pullani; RECENT DEVELOPMENTS: 256 Reports, Standards, and Safety Guides, D. S. Queener; 260 Proposed Rule Changes as of June 30, 1996; ANNOUNCEMENTS: 210 Tennessee Industries Week; 221 21st International Symposium on The Scientific Basis for Nuclear Waste Management; 259 IAEA Programme of Scientific Meetings for 1997; 270 PSAM 4 (International Conference on Probabilistic Safety Assessment and Management); 265 The Authors; 268 Errata.

05 NUCLEAR FUELS↗
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↗