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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A brief overview of compartmental modeling for intake of plutonium via wounds

Here, the aim of this study is to present several approaches that have been used to model the behavior of radioactive materials (specifically Pu) in contaminated wounds. We also review some attempts by the health physics community to validate and revise the National Council on Radiation Protection and Measurements (NCRP) 156 biokinetic model for wounds, and present some general recommendations based on the review. Modeling of intake via the wound pathway is complicated because of a large array of wound characteristics (e.g. solubility and chemistry of the material, type and depth of the tissue injury, anatomical location of injury). Moreover, because a majority of the documented wound cases in humans are medically treated (excised or treated with chelation), the data to develop biokinetic models for unperturbed wound exposures are limited. Since the NCRP wound model was largely developed from animal data, it is important to continue to validate and improve the model using human data whenever plausible.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Analysis of urinary excretion data from three plutonium-contaminated wounds at Los Alamos National Laboratory

The NCRP-156 Report proposes seven different biokinetic models for the wound cases depending on the physicochemistry of the contaminant. Because the models were heavily based on experimental animal data, the authors of the report encouraged application and validation of the models using bioassay data from actual human exposures. Each of the wound models was applied to three plutonium-contaminated wounds, and the models resulted in a good agreement to only one of the cases. We then applied a simpler biokinetic model structure to the bioassay data and showed that fitting the transfer rates from this model structure yielded better agreement with the data than does the best-fitting NCRP-156 model. Because the biokinetics of radioactive material in each wound is different, it is impractical to propose a discrete set of model parameters to describe the biokinetics of radionuclides in all wounds, and thus each wound should be treated empirically.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Primary Beam Air Kerma Dependence on Distance from Cargo and People Scanners

The distance dependence of air kerma or dose rate of the primary radiation beam is not obvious for security scanners of cargo and people in which there is relative motion between a collimated source and the person or object being imaged. To study this problem, one fixed line source and three moving-source scan-geometry cases are considered, each characterized by radiation emanating perpendicular to an axis. The cases are 1) a stationary line source of radioactive material, e.g., contaminated solution in a pipe; 2) a moving, uncollimated point source of radiation that is shuttered or off when it is stationary; 3) a moving, collimated point source of radiation that is shuttered or off when it is stationary; and 4) a translating, narrow “pencil” beam emanating in a flying-spot, raster pattern. Each case is considered for short and long distances compared to the line source length or path traversed by a moving source. The short distance model pertains mostly to dose to objects being scanned and personnel associated with the screening operation. The long distance model pertains mostly to potential dose to bystanders. For radionuclide sources, the number of nuclear transitions that occur a) per unit length of a line source, or b) during the traversal of a point source, is a unifying concept. The “universal source strength” of air kerma rate at a meter from the source can be used to describe x-ray machine or radionuclide sources. For many cargo and people scanners with highly collimated fan or pencil beams, dose varies as the inverse of the distance from the source in the near field and with the inverse square of the distance beyond a critical radius. Ignoring the inverse square dependence and using inverse distance dependence is conservative in the sense of tending to overestimate dose.

Beam↗
DOE OSTI2017

Gamma-ray dose from an overhead plume

Standard plume models can underestimate the gamma-ray dose when most of the radioactive material is above the heads of the receptors. Typically, a model is used to calculate the air concentration at the height of the receptor, and the dose is calculated by multiplying the air concentration by a concentration-to-dose conversion factor. Models indicate that if the plume is emitted from a stack during stable atmospheric conditions, the lower edges of the plume may not reach the ground, in which case both the ground-level concentration and the dose are usually reported as zero. However, in such cases, the dose from overhead gamma-emitting radionuclides may be substantial. Such underestimates could impact decision making in emergency situations. The Monte Carlo N-Particle code, MCNP, was used to calculate the overhead shine dose and to compare with standard plume models. At long distances and during unstable atmospheric conditions, the MCNP results agree with the standard models. As a result, at short distances, where many models calculate zero, the true dose (as modeled by MCNP) can be estimated with simple equations.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Interpretation of Urinary Excretion Data From Plutonium Wound Cases Treated With DTPA: Application of Different Models and Approaches

After a chelation treatment, assessment of intake and doses is the primary concern of an internal dosimetrist. Using the urinary excretion data from two actual wound cases encountered at Los Alamos National Laboratory (LANL), this study discusses several methods that can be used to interpret intakes from the urinary data collected after one or multiple chelation treatments. One of the methods uses only the data assumed to be unaffected by chelation (data collected beyond 100 d after the last treatment). This method, used by many facilities for official dose records, was implemented by employing maximum likelihood analysis and Bayesian analysis methods. The impacts of an improper assumption about the physicochemical behavior of a radioactive material and the importance of the use of a facility-specific biokinetic model when available have also been demonstrated. Another method analyzed both the affected and unaffected urinary data using an empirical urinary excretion model. This method, although case-specific, was useful in determining the actual intakes and the doses averted or the reduction in body burdens due to chelation treatments. Finally, this approach was important in determining the enhancement factors, the behavior of the chelate, and other observations that may be pertinent to several DTPA compartmental modeling approaches being conducted by the health physics community.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2018

EVALUATION OF AN UPWARD TREND IN BACKGROUND COUNT RATES FROM A STACK PARTICULATE CONTINUOUS AIR MONITOR

The Pacific Northwest National Laboratory (PNNL) operates the Radiochemical Processing Laboratory (RPL), which is a multi-purpose, non-reactor nuclear research facility. Regulations require both continuous sampling and monitoring of radioactive particulates and tritium gas in the exhaust from the main stack. Releases of other radioactive gases, including planned releases of radon, are tracked separately in a database and reported. During the 2015 calibration of the RPL stack continuous air monitor (CAM), measured alpha and beta background count rates were much higher than expected, especially when compared to count rates from previous calibrations. The source of the higher background count rates was examined by trending of historical CAM measurements and a comparison to the sampler data. The analysis revealed that the sample results showed no increase in emissions, whereas the CAM showed a steady increase in count rates. Ultimately, the CAM filter was analyzed and found to contain higher-than-normal background levels of 220Rn progeny. Assessments were performed to determine the cause of the increased background values, including reviews of building research activities, radioactive material usage and storage, adequacy of procedures, and the potential for internal CAM contamination. Project reviews determined that a research activity involving 228Th was left in a semi-stored state, resulting in 220Rn being released from a hot cell into the exhaust system. The 228Th source material was subsequently repackaged and contained resulting in a decrease of CAM background count rates. An estimate of the 220Rn release was made and the contribution to the annual offsite dose from the facility was calculated. The released activity and reported dose results were well below the permit limits for the facility.

radon, 220Rn, environmental assessment, air monito↗
DOE OSTI2013

Use of CAP88 at Department of Energy Sites

The U.S. Department of Energy is committed to protecting the public and environment against undue risk from radiation associated with radiological activities conducted under its control. Some U.S. Department of Energy Site activities result in emissions of radioactive materials to the air. CAP88 codes are used to model these emissions and the subsequent maximum estimated dose to a member of the public in the vicinity of the U.S. Department of Energy Site. This paper reviews the use of the CAP88 code at the variety of U.S. Department of Energy sites that use it for compliance reporting under 40 CFR Part 61, Subpart H.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Aerial Survey Results for 131I Deposition on the Ground after the Fukushima Daiichi Nuclear Power Plant Accident

In March 2011 the second largest accidental release of radioactivity in history occurred at the Fukushima Daiichi nuclear power plant following a magnitude 9.0 earthquake and subsequent tsunami. Teams from the U.S. Department of Energy, National Nuclear Security Administration Office of Emergency Response performed aerial surveys to provide initial maps of the dispersal of radioactive material in Japan. The initial results from the surveys did not report the concentration of 131I. This work reports on analyses performed on the initial survey data by a joint Japan-US collaboration to determine 131I ground concentration. This information is potentially useful in reconstruction of the inhalation and external exposure doses from this short-lived radionuclide. The deposited concentration of 134Cs is also reported.

Fukushima, I-131, aerial survey↗
DOE OSTI2023

Development of Fast Decay-Energy Spectroscopy With Magnetic Microcalorimeters

Decay energy spectroscopy (DES) is an increasingly important radiometric technique arising from the unique thermal detection physics and the precision of low-temperature microcalorimetry. DES can enable high-precision analysis of very small amounts of radioactive material with simple and rapid sample preparation, making it a key new tool for nuclear safeguards, nuclear forensics, the improvement of nuclear data, measurements of absolute activity, and other applications. For problems in nuclear safeguards and nuclear forensics it is highly desirable to increase the count rates achievable in DES, while maintaining simple and rapid sample preparation. Here, in this report, we describe progress in a new project to achieve high count rates in DES by taking advantage of the physics of magnetic microcalorimetry. We describe the results of initial modeling and experiment to validate the idea, a sample fabrication technique suitable for nuclear materials laboratories, the preparation and initial tests of paramagnetic Au-Er alloy in thin foil form, and the development of a sensing coil geometry that will perform well with the increased-thickness sensor used in the new technique.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

The spectrometric determination of the individual exposure rate for gamma nuclides from an environmental radiation detector

For making the spectrometric determination of the exposure rate from the environment as well as the radioactive material more practical, an accurate calculation method of the individual exposure rate for the detected gamma nuclides from that spectrum should be suggested without the sophisticated calibration procedure. In this study, the calculation method for the individual exposure rate for detected gamma nuclides from a 3'x3' NaI(Tl) detector was suggested by introducing the concept of the dose rate spectroscopy and the peak-to-total ratio in the energy spectrum for the exposure rate, which means just a form of multiplied counts and the value of a G-factor in the spectrum. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

HELIUM LEAK TEST MODELING OF A SPENT NUCLEAR FUEL CANISTER

The U.S. Department of Energy (DOE) is considering the development of one or more federal consolidated interim storage facilities (CISFs) to be used to store commercial spent nuclear fuel (SNF) at locations in the U.S. One of the first technical challenges of a CISF is performing an inspection of SNF canisters upon their receipt to confirm they can be placed into the CISF’s licensed storage configuration. The canister receipt inspection is critical to CISF site operations. The test is conceived as being a helium (He) leak check, intended to confirm that the confinement boundary of a SNF canister is intact. SNF canisters are filled with He when they are sealed, so detection of a He leak indicates that a through-wall flaw has occurred in the canister confinement boundary. Other measurements are planned to occur upon canister receipt in addition to the He leak check such as krypton-85 measurements, which would indicate confinement breaches of one or more fuel rods in addition to a breach of the SNF canister. However, the He leak check has been identified as one of such high importance and has such significant technical challenges that a full-scale demonstration is needed to confirm the He leak test’s viability and to assist in planning relative to its operational requirements. A modeling methodology for simulating the He detection test was developed to help inform the test plan and the design of the test vessels. To develop the modeling methodology a detailed computational fluid dynamics (CFD) benchmark model was constructed to compare against leak rate test data from a transportation package for radioactive material. This report is focused on modeling efforts to simulate the benchmark leak test.

Suffield, Sarah R.↗
DOE OSTI2016

SU-F-T-10: Validation of ELP Dosimetry Using PRESAGE Dosimeter: Feasibility Test and Practical Considerations

Purpose: To validate the use of a PRESAGE dosimeter as a method to quantitatively measure dose distributions of injectable brachytherapy based on elastin-like polypeptide (ELP) nanoparticles. PRESAGE is a solid, transparent polyurethane-based dosimeter whose dose is proportional to a change in optical density, making it useful for visualizing the dose from a radionuclide-tagged-ELP injection. Methods: A PRESAGE dosimeter was designed to simulate an ELP injection. To calibrate, cuvette samples from the batch of PRESAGE were exposed to varying levels of radiation from 0–35.9Gy applied via a linear accelerator, then placed into a spectrophotometer to obtain the optical density change as a function of dose. A pre-optical-CT scan was acquired of the phantom to obtain a baseline tomographic optical density. A 1cc saline solution of I-125 tagged-ELP with and activity concentration of 1mCi/cc was injected into the phantom and left for five days. After five days, the ELP was removed and the cavity cleaned of all remaining radioactive material. Post tomographic optical images were acquired to obtain a differential optical density dataset. Results: Initial results after the 5-day exposure revealed an opaque white film that resembled the volume of the ELP solution injected into the phantom. We think this is possibly due to the saline solution diffusing into the PRESAGE and causing a change in the index of refraction at this shallow depth. Therefore, initially the optical scanner yielded inconclusive results. After several more days, the saline seemed to have evaporated out of the injection site and the ELP dose distribution was visible via color change in the dosimeter. Conclusion: We have created the first experimental design to measure the dose distribution of I-125-tagged-ELP. The PRESAGE formulation proves to be a feasible option for such measurements. Future experimental measurements need to be obtained to further characterize ELP dosimetry.

60 APPLIED LIFE SCIENCES↗
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 OSTI2020

Monitoring the Caustic Dissolution of Aluminum Alloy in a Radiochemical Hot Cell Using Raman Spectroscopy

Chemical processing of highly radioactive materials commonly takes place in heavily shielded hot cells. The remote, real-time monitoring of chemical processing streams via optical spectroscopic techniques in hot cells may be particularly useful. Here in this paper, we describe the implementation of Raman spectroscopy and chemometric analysis to monitor the dissolution of aluminum-clad targets containing irradiated aluminum–neptunium oxide cermet pellets in caustic solutions in a hot cell environment. Partial least squares regression analysis was used to generate calibration models to quantify the concentration of dissolved aluminum, nitrate, and hydroxide in solutions within the radiochemical hot cell. This work explored a systematic approach to optimize a matrix of calibration standards using a D-optimal experimental design. The Design of Experiments-based regression model, in comparison to more traditional analytical approaches, was found to be the more practical method for building calibration models, with fewer samples, to obtain informative analytical data from Raman spectra.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Tank Waste Characterization: History, Challenges, and Success Stories

The preparation and chemical and radiochemical analysis of Hanford tank waste samples can be performed with standard laboratory equipment and instruments as relatively routine processes that are not particularly challenging. Rather, the main challenges of tank waste characterization are associated with radiological dose and sampling limitations. Accurate, representative and effective sampling techniques are difficult with the waste tanks because they were not designed for routine sampling. There are a finite number of sampling locations for each tank based on riser positioning, depth and the operational functionality of the sampling riser. For example, in one recently emptied SST, there was one riser that was found to have had concrete dumped down it, thereby eliminating that sampling port. Additionally, the waste within the tank; especially true for the saltcake and sludge, is not homogenous. The ability to adequately mix a million-gallon double shell tank (DST) is a concern for data reproducibility. Another real challenge that must be addressed for sampling single shell tanks, is how to dissolve the salt cake waste in a compromised (leaking) SST. These physical constraints mean that uncertainty in the representativeness of samples must be considered when applying analytical results to the bulk contents of the tank. The tank waste is highly radioactive and thus can only be handled initially by facilities that can receive samples into concrete-shielded hot cells with remote operation with an example provided in Figure 1. The shielding protects the worker from the radiological dose while mineral oil windows and remotely operated manipulators enables the samples to be handled. At Hanford, analytical laboratories with these hot cell capabilities are limited to the Pacific Northwest National Laboratory and the main Hanford operations support laboratory, 222-S Laboratory. Because of their highly radioactive nature, samples must be sufficiently diluted to facilitate their analysis outside of a shielded cell. In some cases, this means some accuracy must be compromised to complete the analysis beyond that normally encountered for non-radioactive material.

Waste Characterization, BBI, PHOENIX: Tank Farms: ↗
DOE OSTI2013

The Future of Energy from Nuclear Fission

Nuclear energy is an important part of our current global energy system, and contributes to supplying the significant demand for electricity for many nations around the world. There are 433 commercial nuclear power reactors operating in 30 countries with an installed capacity of 367 GWe as of October 2011 (IAEA PRIS, 2011). Nuclear electricity generation totaled 2630 TWh in 2010 representing 14% the world’s electricity generation. The top five countries of total installed nuclear capacity are the US, France, Japan, Russia and South Korea at 102, 63, 45, 24, and 21 GWe, respectively (WNA, 2012a). The nuclear capacity of these five countries represents more than half, 68%, of the total global nuclear capacity. The role of nuclear power in the global energy system today has been motivated by several factors including the growing demand for electric power, the regional availability of fossil resources and energy security concerns, and the relative competitiveness of nuclear power as a source of base-load electricity. There is additional motivation for the use of nuclear power because it does not produce greenhouse gas (GHG) emissions or local air pollutants during its operation and contributes to low levels of emissions throughout the lifecycle of the nuclear energy system (Beerten, J. et. al., 2009). Energy from nuclear fission primarily in the form of electric power and potentially as a source of industrial heat could play a greater role for meeting the long-term growing demand for energy worldwide while addressing the concern for climate change from rising GHG emissions. However, the nature of nuclear fission as a tremendously compact and dense form of energy production with associated high concentrations of radioactive materials has particular and unique challenges as well as benefits. These challenges include not only the safety and cost of nuclear reactors, but proliferation concerns, safeguard and storage of nuclear materials associated with nuclear fuel cycles. In March of 2011, an unprecedented earthquake of 9 magnitude and ensuing tsunami off the east coast of Japan caused a severe nuclear accident in Fukushima, Japan (Prime Minister of Japan and His Cabinet, 2011). The severity of the nuclear accident in Japan has brought about a reinvestigation of nuclear energy policy and deployment activities for many nations around the world, most notably in Japan and Germany (BBC, 2011; Reuter, 2011). The response to the accident has been mixed and its full impact may not be realized for many years to come. The nuclear accident in Fukushima, Japan has not directly affected the significant on-going nuclear deployment activities in many countries. China, Russia, India, and South Korea, as well as others, are continuing with their deployment plans. As of October 2011, China had the most reactors under construction at 27, while Russia, India, and South Korea had 11, 6, and 5 reactors under construction, respectively (IAEA PRIS, 2011). Ten other nations have one or two reactors currently under construction. Many more reactors are planned for future deployment in China, Russia, and India, as well as in the US. Based on the World Nuclear Association’s data, the realization of China’s deployment plan implies that China will surpass the US in total nuclear capacity some time in the future.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Transitioning from Cobalt-60 to X-Ray or E-Beam for Medical Sterilization: Filling Data and Education Gaps

The Office of Radiological Security (ORS) within the U.S. DOE’s National Nuclear Security Administration (NNSA) is charged to “work with government, law enforcement, and businesses across the globe to protect radioactive sources used for medical, research, and commercial purposes; remove and dispose of disused radioactive sources; and reduce the global reliance on high activity radioactive sources through the promotion of viable non-radioisotopic alternative technologies.” In this effort, the NNSA/ORS, along with the Office of Defense Nuclear Nonproliferation R&D, has been working with government and private entities that utilize gamma-ray irradiators in areas that include blood irradiation, radiotherapy, biology research, well logging, and radiation processing to consider alternative technologies that do not depend upon radioactive materials. In the medical device industry, nearly 50% of products are sterilized with ethylene oxide, 40.5% with cobalt-60 gamma-rays, 4.5% with electron-beam (e-beam), and <1% with X-ray.(1,2) The cobalt-60 irradiation facilities use an estimated 150 MCi of isotope to sterilize approximately 200 million cubic feet of product annually.(3) Due to growing safety regulations governing cobalt-60 use, continued advancements in e-beam and X-ray accelerator technology, and the greater irradiation durations and costs for cobalt-60 irradiation compared to e-beam and X-ray, switching to these non-radionuclide modalities is increasingly economically competitive. However, this breakdown of 40.5% cobalt-60 and 5.5% accelerator for product sterilization illustrates the difficulty that market forces and industry players have had in overcoming barriers in transitioning to non-nuclide alternatives. Besides those associated with construction of new e-beam and X-ray facilities, significant financial and regulatory obstacles are also associated with the irradiation and comprehensive testing of product constituent polymers and elastomers. The standards associated with product material performance testing are the Association for the Advancement of Medical Instrumentation (AAMI) Technical Information Report 17 (TIR 17), and its referenced American Society for Testing and Materials (ASTM) polymer testing standards. In the study now underway, and the subject of this article, the NNSA/ORS provided R&D funding to the Pacific Northwest National Laboratory (PNNL) in partnership with university researchers, leading medical product manufacturers, accelerator manufacturers, and contract sterilizers. The resulting Team Nablo (in memory of Sam Nablo) includes individuals representing PNNL, Becton Dickinson (BD), Johnson & Johnson, Stryker, Texas A&M University (TAMU), Steri-Tek, Mevex, Ion Beam Applications (IBA), and Sterigenics. The goals of this study are to 1) Identify specific polymer and/or elastomer-based medical products that are currently sterilized with gamma-rays, and for which there are data gaps and would be of greatest industry impact, 2) Work with relevant ASTM subcommittees for guidance and for potential standard revision, 3) Work with an appropriate Food & Drug Administration (FDA) office for potential collaboration on a guidance document for device manufacturers, 4) Measure the physical effects that product materials exhibit when they are given sterilization-level radiation doses from cobalt-60, e-beam or X-ray, and determine whether these effects would preclude the use of e-beam or X-ray for sterilization, 5) Communicate the resulting scientific datasets and findings via industry conferences, peer-reviewed journals, and relevant trade magazines, and 6) Post the resulting datasets on a publicly accessible website, which can be built on in future years. The end objective is to identify technical barriers preventing the adoption of X-ray and e-beam technologies, and share the results to encourage X-ray and e-beam adoption where feasible.

sterilization, alternate technologies, gamma, X-ra↗
DOE OSTI2014

Nuclear Arms Control, Nonproliferation, and Counterterrorism: Impacts on Public Health

Reducing the risks of nuclear war, limiting the spread of nuclear weapons and reducing global nuclear weapons stockpiles are key national and international security goals. They are pursued through a variety of international arms control, nonproliferation and counter-terrorism treaties and agreements. These legally binding and political commitments, together with the institutional infrastructure that supports them, work to establish global norms of behavior and have limited the spread of weapons of mass destruction. Beyond the primary security objectives, reducing the likelihood of the use of nuclear weapons, preventing environmental releases of radioactive material, increasing the availability of safe and secure nuclear technology for peaceful purposes, and providing scientific data relevant to predicting and managing the consequences of natural or human-caused disasters world-wide provide significant benefits to global public health.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗