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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Predeployment progress of the Canister Deposition Field Demonstration

This report updates the high-level test plan for evaluating surface deposition on three commercial 32PTH2 spent nuclear fuel (SNF) canisters inside NUTECH Horizontal Modular Storage (NUHOMS) Advanced Horizontal Storage Modules (AHSMs) from Orano (formerly Transnuclear Inc.) and provides a summary of the surface sampling activities that have been conducted to date. The details contained in this report represent the best designs and approaches explored for testing as of this publication. Given the rapidly developing nature of this test program, some of these plans may change to accommodate new objectives or requirements. One goal of this testing is to collect defensible and detailed dust deposition measurements from the surface of dry storage canisters in a marine coastal environment to guide chloride-induced stress corrosion cracking (CISCC) research. Another goal is to provide data for the validation of computational fluid dynamics (CFD) based deposition modeling. To facilitate surface sampling, the otherwise highly prototypic dry storage systems will not contain SNF but rather will be electrically heated to mimic the decay heat and thermal hydraulic environment. Test and heater design is supported by detailed CFD modeling. Instrumentation throughout the canister, storage module, and environment will provide extensive information about the thermal-hydraulic behavior of horizontal dry cask storage systems. Manual sampling over a comprehensive portion of the canister surface at regular time intervals will offer detailed quantification and composition of the deposited particulates from a realistic storage environment. Discussions of a potential host site for the Canister Deposition Field Demonstration (CDFD) are ongoing. Until a host site is chosen, testing of key CDFD hardware components including the heater assemblies, power skid, and remote data acquisition system will continue. Functional testing of the finalized heater assemblies and test apparatus started this fiscal year. These initial heater tests have shown the assemblies are performing within design specifications. Staged surface sampling of a mockup of a canister outside the AHSM on a transfer skid was also performed. Refinements to the sampling procedures and techniques were captured from observation of these activities and lessons-learned debriefs. These updated sampling procedures and techniques are planned to be tested again in the field using the mockup in order to assure personnel are using the most accurate and repeatable methods possible prior to deployment for actual CDFD testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Interstitial and substitutional zirconium in SrTiO3

We investigate Zr in SrTiO3 (STO) as a model for nuclear waste forms in which the fission product Sr90 eventually decays to stable Zr through beta emission. The transformation of a divalent into a tetravalent constituent is expected to affect the long-term structural and chemical stability of this solid. Computational methods of electronic structure theory, specifically the density functional theory (DFT) within the supercell model, are used to predict the thermodynamic stability and electronic states of interstitial and Sr- or Ti-substituted Zr atoms in the STO lattice. Native defects such as vacancies and antisites are also considered. When Zr replaces Sr, its most stable configuration is to simply occupy the Sr site (instead of, for example, replacing a Ti and displacing the Ti to the Sr site.) For Zr added to the lattice, its most stable configuration is to replace a Ti, making a ZrTi impurity plus a Ti interstitial (as opposed to the Zr just remaining as an interstitial atom.) ZrSr is predicted to be a double electron donor, ZrTi is electrically inactive and interstitial Zr and Ti are predicted to be quadruple donors, with all donor levels in the conduction band. ZrSr and the tetravalent interstitials are all predicted to increase the crystal volume, and the interstitials also are predicted to lead to a tetragonal distortion of the lattice. Experiments with injection of Zr atoms into STO qualitatively confirm these predictions of crystal structural changes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Constraint of the Astrophysical Al 26 g ( p , γ ) Si 27 Destruction Rate at Stellar Temperatures

The Galactic 1.809-MeV γ-ray signature from the β decay of 26g Al is a dominant target of γ-ray astronomy, of which a significant component is understood to originate from massive stars. The 26g Al(p,γ) 27 Si reaction is a major destruction pathway for 26g Al at stellar temperatures, but the reaction rate is poorly constrained due to uncertainties in the strengths of low-lying resonances in 27 Si. The 26g Al (d,p) 27 Al reaction has been employed in inverse kinematics to determine the spectroscopic factors, and hence resonance strengths, of proton resonances in 27 Si via mirror symmetry. Finally, the strength of the 127-keV resonance is found to be a factor of 4 higher than the previously adopted upper limit, and the upper limit for the 68-keV resonance has been reduced by an order of magnitude, considerably constraining the 26g Al destruction rate at stellar temperatures.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Parameter Uncertainty for Repository Thermal Analysis

This report is one follow-on to a study of reference geologic disposal design concepts (Hardin et al. 2011a). Based on an analysis of maximum temperatures, that study concluded that certain disposal concepts would require extended decay storage prior to emplacement, or the use of small waste packages, or both. The study used nominal values for thermal properties of host geologic media and engineered materials, demonstrating the need for uncertainty analysis to support the conclusions. This report is a first step that identifies the input parameters of the maximum temperature calculation, surveys published data on measured values, uses an analytical approach to determine which parameters are most important, and performs an example sensitivity analysis. Using results from this first step, temperature calculations planned for FY12 can focus on only the important parameters, and can use the uncertainty ranges reported here. The survey of published information on thermal properties of geologic media and engineered materials, is intended to be sufficient for use in generic calculations to evaluate the feasibility of reference disposal concepts. A full compendium of literature data is beyond the scope of this report. The term “uncertainty” is used here to represent both measurement uncertainty and spatial variability, or variability across host geologic units. For the most important parameters (e.g., buffer thermal conductivity) the extent of literature data surveyed samples these different forms of uncertainty and variability. Finally, this report is intended to be one chapter or section of a larger FY12 deliverable summarizing all the work on design concepts and thermal load management for geologic disposal (M3FT-12SN0804032, due 15Aug2012).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Global Solutions Through Simulation For Better Decommissioning - 16457

Each nuclear facility is different due to their history and there is no rule for choosing a decommissioning strategy. There are three major decommissioning strategies. First, 'immediate dismantling', which means the decommissioning begins immediately after the transfer of waste and nuclear material. Second, 'deferred dismantling strategy', meaning the facility is maintained into a containment zone from thirty to one hundred years before being decommissioned. Finally, 'entombment', referring to the placement of the facility into a reinforced containment until the radionuclides decay and reach a level allowing the site release. When a strategy is decided many factors have to be taken into account. Into a major project such as a reactor decommissioning, there are many smaller projects and the decommissioning strategy can be different among these smaller projects. Very often, some entry data are not perfectly known, for example, the dosimetric activity has not been updated through time or after specific events. Because of uncertainties and/or hypothesis existing around projects and their high level of interdependency, global solutions are a good way to choose the best decommissioning strategy. Especially since each entry data has consequences on output results whether it is on costs, cumulated dose, waste or delays. Also the dose, delays or waste management, all have impact on costs. To obtain an optimal scenario into a special environment, it is necessary to deal with all these items together. This global solution can be implemented thanks to simulation in dedicated software which helps define the global strategy, optimize the scenario, and prevent contingencies. For a specific environment, when many scenarios have been developed and simulated, sensitivity studies are really important to choose the optimal decommissioning strategy. Sensitivity studies can be done with simulation tools by comparing results of scenarios directly into the software. This comparison helps to choose the optimal scenario according to which output data is considered as key determinant. (authors)

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

Scoping Thermal Response Calculations of RNS Waste During Transport to and Disposal at the WIPP

Sandia National Laboratories (SNL) was contracted by the United States Department of Energy Environmental Management (DOE-EM), Los Alamos Field Office to perform mechanical and thermal scoping calculations as part of a study seeking to understand the ignitability risk of the Remediated Nitrate Salts (RNS) waste drums during transportation from the Waste Control Specialists (WCS) facility to Waste Isolation Pilot Plant (WIPP) and permanent disposal of the waste at WIPP. The scoping thermal simulations described in this report pertain to thermal calculations performed with a packaging system consisting of one Standard Waste Box (SWB) loaded with drums placed inside a Standard Large Box 2 (SLB2). During transportation, the SLB2 is inside Transuranic Package Transporter Model III (TRUPACT-III), which provides the third layer of the packaging. Once at the WIPP, it is assumed the SLB2 is extracted from the TRUPACT-III and maintained above ground, and then subsequently placed underground for permanent disposal. In these proposed configurations, the space between the SLB2 and the SWB is always filled by a layer of insulation consisting of air-filled glass microbubbles except for the bottom which rests directly on the SLB2. The thermal scoping calculations described in this report specifically address whether the introduction of external heat inputs, combined with the contributions from the internally generated radiolytic decay heat and chemical reactions, lead to an unstable thermal state during the time of its movement and placement in the permanent disposal location. The external heat inputs are of two forms: 1) ambient thermal irradiation (e.g., solar and ambient storage/disposal temperatures) and 2) accident-induced fire. Three scoping calculation scenarios were derived as representative, conservative scenarios: 1A) TRUPACT-III transient transportation, 1B) SLB2 48-hour outdoor storage with solar radiation, and 2) fully-engulfing fire during SLB2 handling or emplacement following a steady-state analysis in a 38 °C environment. All the simulated scenarios are conservative relative to the operational conditions expected for handling the waste package during transportation and placement in the WIPP underground disposal unit. The predictions obtained from simulating the three exposure scenarios revealed that adding the SLB2 and the air-filled glass microbubbles to the transport and storage/disposal configurations provides additional thermal protection of the drums beyond what the SWB provides alone, both during long-term above ground insolation and underground during a fire accident. Under the current transportation/storage/disposal concepts, the degree of protection provided by the packaging concept is sufficient to prevent the waste from being ignitable. The simulation results demonstrate that there is adequate margin to safely transport and place the RNS waste from WCS to the WIPP under the current operational concept.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2012

Khazar Iodine Production Plant Site Remediation in Turkmenistan. NORM Contaminated Waste Repository Establishment - 12398

Radiation safety provisions for NORM contaminated areas are in use in a number of the former Soviet republics. Some of these areas were formed by absorbed radionuclides at the iodine and bromine extraction sites. As a rule, there are not any plant radiation monitoring systems nor appropriate services to ensure personnel, population and environmental radiation safety. The most hazardous sites are those which are situated in the Caspian Sea coastal zone. The bulk of the accumulated waste is represented by a loose mixture of sand and charcoal, which was basically used as the iodine extraction sorbent. The amounts of these wastes were estimated to be approximately 20,000 metric tons. The waste contamination is mainly composed of Ra-226 (U-238 decay series) and Ra-224, Ra-228 (Th-232 decay series). In 2009, the 'ECOMET-S', a Closed Joint-Stock Company from St. Petersburg, Russian Federation, was authorized by the Turkmenistan government to launch the rehabilitation project. The project includes D and D activities, contaminated areas remediation, collected wastes safe transportation to the repository and its disposal following repository closure. The work at the Khazar chemical plant started in September, 2010. Comprehensive radiological surveys to estimate the waste quantities were carried out in advance. In course of the rehabilitation work at the site of the Khazar chemical plant additional waste quantities (5,000 MT, 10,000 m{sup 3}) were discovered after the sludge was dumped and drained. Disposal volumes for this waste was not provided initially. The additional volume of the construction wastes was required in order to accommodate all the waste to be disposed. For the larger disposal volume the project design enterprise VNIPIET, offered to erect a second wall outside the existing one and this solution was adopted. As of May, 2011, 40,575 m{sup 3} of contaminated waste were collected and disposed safely. This volume represents 96.6% of the initial repository volume. Now work is underway to erect the second repository wall, which will allow housing of the additional 16,800 m{sup 3} The Khazar chemical plant territory restoration work is underway as well. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1990

Heating and melting of small icy satellites by the decay of 26Al

We study the effect of radiogenic heating due to 26Al on the thermal evolution of small icy satellites. Our object is to find the extent of internal melting as a function of the satellite radius and of the initial 26Al abundance. The implicit assumption, based on observations of young stars, is that planet and satellite accretion occurred on a time scale of approximately 10(6) yr (comparable with the lifetime of 26Al). The icy satellites are modeled as spheres of initially amorphous ice, with chondritic abundances of 40K, 232Th, 235U, 238U, corresponding to an ice/dust mass ratio of 1. Evolutionary calculations are carried out, spanning 4.5 x 10(9) yr, for different combinations of the two free parameters. Heat transfer by subsolidus convection is neglected for these small satellites. Our main conclusion is that the initial 26Al abundance capable of melting icy bodies of satellite size to a significant extent is more than 10 times lower than that prevailing in the interstellar medium (or that inferred from the Ca-Al rich inclusions of the Allende meteorite, approximately 7 x 10(-7) by mass). We find, for example, that an initial 26Al mass fraction of approximately 4 x 10(-8) is sufficient for melting almost completely icy spheres with radii of 800 km, typical of the larger icy planetary satellites. We also find that for any given 26Al abundance, there is a narrow range of radii below which only marginal melting occurs and above which most of the ice melts (and refreezes later). Since extensive melting may have important consequences, such as differentiation, gas release, and volcanic activity, the effect of 26Al should be included in future studies of satellite interiors.

Non-NASA Center↗
DOE OSTI2014

Waste Management Analyses for Spent Structures at SNS - Paper 106

The Spallation Neutron Source (SNS) in Oak Ridge, Tennessee, is an accelerator driven neutron scattering facility for materials research. Presently, the SNS operates at 1.2 megawatt (MW) proton beam power incident on a mercury target with a proton beam energy of 1 GeV and 60 Hz repetition rate. The SNS consists of an accelerator facility, target facility, and a world-class suite of neutron scattering instruments supporting material, life-science and fundamental physics research. According to the SNS operations plan, some components are replaced when they reach their end-of-life due to radiation induced material damage or burn-up, or because of mechanical failure. During operation, these components, especially those in proximity to the target, are exposed to a radiation environment and build up significant activity during their service lifetime. All these components must be safely removed, placed in a container/package for storage, and ultimately transported offsite for disposal. An accurate estimate of the radionuclide inventory is performed assuming realistic irradiation history and decay time in order to classify a spent component. On the basis of a calculated radionuclide inventory, the spent component is classified, and an appropriate cask/package for transport and storage is suggested. After the choice for the container is made, radiation transport calculations are performed for the facility component placed inside the container, to ensure that the container is compliant with waste management regulations. When necessary, additional shielding is added. For each spent component before it goes out of the SNS site, supporting documentation is produced with a radionuclide inventory prediction, dose rate analyses and, if required, additional shielding. A dose rate analyses for the exposure prediction of personnel during components change out are performed as well. Most of the effort is concentrated on the target facility components such as target, proton beam window (PBW) and core vessel insert (CVI) plugs because of severe radiation conditions. Additionally, analyses are performed for the accelerator facility components, which are removed from the facility. Activation analyses and residual dose rates analyses are performed to support safe handling of SNS spent components at the SNS site and during transport to the waste disposal site. On the basis of the calculated radionuclide inventories, spent components characterized and classified with respect to waste management guidelines, and activity-based determinations are made regarding appropriate packaging for transport. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Hydride Bed Helium-3 Recovery and Partial Regeneration

Savannah River Tritium Enterprise has used LaNi 4.25 Al 0.75 (LANA75) hydride beds to store hydrogen isotopes for over two decades. A benefit of using LANA75 is that the 3He generated from tritium decay is retained in the hydride material, allowing the hydride beds to deliver high-purity product gas. A disadvantage is that the 3He accumulates in the LANA75 material over time, which forms a heel that cannot be removed under normal operating conditions. The heel traps hydrogen in the bed, slowly reducing the operational capacity of the bed as the heel grows. Eventually, the 3 He begins to release from the material, preventing the delivery of high-purity product. The hydride beds are replaced when (1) operational capacity is reduced such that it is impactive to routine operations, and/or (2) product purity is not maintained due to 3 He release.Several beds were operated beyond their design life. One of these beds was selected to undergo heating beyond its normal operating temperature to evaluate the possibility of removing a portion of the hydrogen and helium heel to improve bed function until a replacement could take place. This bake-out removed a portion of the hydrogen and helium heel, and preliminary data indicate that bake-outs may partially regenerate the beds. The bed’s performance will continue to be monitored, and additional bake-outs will likely be performed. Performing bake-outs results in increasing the recovery of 3 He, more efficient end-of-life activities (such as isotopic exchange), and extension of the useful service life of the bed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

International Collaboration Activities on Disposal in Argillite R&D: DECOVALEX19 Task C – Geochemical and Reactive Transport Modeling

International collaborations on nuclear waste disposal R&D are an integral part of the Spent Fuel Waste Science and Technology (SFWST) campaign within the DOE Fuel Cycle and Technology (FCT) program. These partnerships with international repository R&D programs provide key opportunities to participate in experiments developing laboratory/field data (underground research laboratories (URL)) of engineered barrier system (EBS) interactions (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field). The results of these experiments are used in the evaluation of coupled processes and their representation via state-of-the-art simulation approaches to evaluate repository performance. During the thermal heating period, increases in temperature from radionuclide decay in the spent fuel (SF) waste canisters will increase temperature in the surrounding EBS driving chemical and transport processes in the near- and far-field domains of the repository. URL heater-tests for extended periods of times (e.g., years) provide key information and data on thermal effects affecting engineered barriers in response to temperature and water saturation levels. Groundwater interactions with cementitious barriers are also important to in-drift chemistry and EBS performance during post-closure. Descriptions of the various URL experiments for various disposal design concepts according to the host country repository program and relevance to the US program is given elsewhere (Birkholzer et al.,2017;Jové Col& et al., 2016). The DECOVALEX-2019 Task C involves collaboration with the GREET (Groundwater REcovery Experiment in Tunnel) at the Mizunami URL, Japan ,which targets the development of monitoring methodologies of groundwater in granitic rock with applications to THMC simulations (Iwatsuki et al., 2005;Iwatsuki et al.,2015,2017). Some of the goals of GREET is to conduct a facility-scale geochemical characterization study of short- and long-term effects of tunnel excavation activities, impacts on groundwater flow and transport, and influences on groundwater chemistry (Iwatsuki et al.,2015). The data obtained from these URL activities is then used in the development and evaluation of THC models to support post-closure safety and performance assessments of the repository environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Depleted Uranium Discrimination Using Spectral Ratios from Data-Starved Spectra - 18619

The Institute for Clean Energy Technology (ICET) has developed mobile surveying platforms for the retrieval of fired depleted uranium (DU) munitions. ICET surveying platforms are fitted with differential global positioning system (GPS), scintillation detectors, and central processing units (CPUs). The surveying technology has historically been used to assist facilities where DU contamination is a concern to locate, retrieve, and remediate DU-contaminated areas. Residence times over DU sources for mobile systems normally do not produce enough data to use gamma ray spectroscopy to verify the source type. Discussed is a potential method used to discriminate DU signatures in data-starved gamma ray spectral data collected during field surveys. Laboratory tests have shown that data-starved gamma ray spectra for zero-valence, oxide, and uranium ores can be distinguished from each other and from background by spectral ratioing. Zero-valence DU and DU oxide gamma ray signatures differ due to the attenuation of low-energy gammas emitted from U-238 daughters. DU signatures can be differentiated from naturally occurring uranium spectra due to the presence of daughters below U-234, primarily bismuth (Bi)-214, in the uranium decay series. Three primary regions of interest (ROI) within the analyzed test spectra are used to roughly characterize the shape of each test spectrum and compare them to background. The three ROIs are focused around the gamma-emitting daughters of U-238: thorium (Th)-234, protactinium (Pa)-234 metastable (m), and Bi-214. One of the three regions is considered the base region. This region is used to estimate what the counts in the other two regions should be by multiplying the counts in the ROIs by a ratio of background counts from an area previously found to be impacted. Using this spectral ratioing technique, a non-zero difference in spectral ratio values indicates a shift in the test spectra away from background. Calculating ratios between the base region and the other two regions allows for a graphical representation of the data in a two-dimensional spectral ratioing space, which is used to differentiate between the types of uranium sources previously mentioned. Surveying processes that are based on gross counts from scintillation counters can result in a large number of false positives as a result of shifting background signals or the presence of naturally occurring uranium. Application of the spectral ratioing technique can further assist in site characterization by analyzing all of the data-starved survey spectra to identify false positives or the need for additional data collection. Accurate surveys of DU testing and training ranges reduces the costs associated with range clean-up by discriminating clean areas from those requiring remediation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A Case Study Correlating Innovative Gamma Ray Scanning Detection Systems Data to Surface Soil Gamma Spectrometry Results - 13580

HydroGeoLogic (HGL), Inc. completed a United States Environmental Protection Agency (USEPA) study to characterize radiological contamination at a site near Canoga Park, California. The characterized area contained 470 acres including the site of a prototype commercial nuclear reactor and other nuclear design, testing, and support operations from the 1950's until 1988 [1]. The site history included radiological releases during operation followed by D and D activities. The characterization was conducted under an accelerated schedule and the results will support the project remediation. The project has a rigorous cleanup to background agenda and does not allow for comparison to risk-based guidelines. To target soil sample locations, multiple lines of evidence were evaluated including a gamma radiation survey, geophysical surveys, historical site assessment, aerial photographs, and former worker interviews. Due to the time since production and decay, the primary gamma emitting radionuclide remaining is cesium-137 (Cs-137). The gamma ray survey covered diverse, rugged terrain using custom designed sodium iodide thallium-activated (NaI(Tl)) scintillation detection systems. The survey goals included attaining 100% ground surface coverage and detecting gamma radiation as sensitively as possible. The effectiveness of innovative gamma ray detection systems was tested by correlating field Cs-137 static count ratios to Cs-137 laboratory gamma spectrometry results. As a case study, the area encompassing the former location of the first nuclear power station in the U. S. was scanned, and second by second global positioning system (GPS)-linked gamma spectral data were evaluated by examining total count rate and nuclide-specific regions of interest. To compensate for Compton scattering from higher energy naturally occurring radionuclides (U-238, Th-232 and their progeny, and K-40), count rate ratios of anthropogenic nuclide-specific regions of interest to the total count rate were calculated. From the scanning data, locations with observed Cs-137 ratios exceeding six standard deviations above the mean ratio were mapped in high resolution [2]. Field teams returned to those locations to collect static count measurements using the same detection systems. Soil surface samples were collected at 30 locations and analyzed for Cs-137. An exponential correlation was identified between Cs-137 concentrations in surface soil and field-scanned Cs-137 ratios. The data indicate field minimum detectable concentration (MDC) of Cs-137 at 0.02 Bq/g (0.5 pCi/g) or lower depending on contaminant distribution in soil. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Predicting Future Solar Modulation and Implementation in MCNP6

The application of detecting low level radiation in the field requires quantification of the natural radiation background specific to that location. This background is made up two primary components: naturally decaying radionuclides in the soil and atmosphere known as terrestrial background, and the particle cascades caused by high energy cosmic radiation interacting within the earth's atmosphere known as cosmic background. The terrestrial component dominates the background spectrum below approximately 2 MeV, however, the cosmic component comprises the entirety of the high-energy spectrum. Furthermore, the terrestrial component is relatively constant for a given location while the cosmic component is heavily dependent on solar activity, an effect known as solar modulation. Annual values of the solar modulation parameter, φ, have been determined since 1936, and their fluctuations are seen to be roughly sinusoidal over time. If one wished to predict or simulate the outcome of an experiment in the future, a prediction for the solar modulation would be necessary to accurately determine the cosmic background contribution. The previous version 6.1.1 of MCNP contained solar modulation data for 46 years between 1960 and 2005. This data was updated with newer solar modulation data, from I. G. Usoskin et al., which reports solar modulation data for an additional 33 years between 1936 and 2014. In order to use past solar modulation data to predict cosmic background radiation fluctuations for dates beyond 2014 an algorithm was developed to forecast future solar modulation data based on the sinusoidal behavior exhibited in recorded data. This algorithm was implemented into MCNP 6.1.2 to be used for cosmic background simulations and tested against benchmark data to assess its accuracy. (authors)

54 ENVIRONMENTAL SCIENCES↗