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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Technical/ administrative options for managing tritium MCL exceedances in P-area groundwater and Steel Creek

This white paper was requested by the Core Team (United States Department of Energy [USDOE], United States Environmental Protection Agency [USEPA], and South Carolina Department of Health and Environmental Control [SCDHEC]) at the P-Area Groundwater (PAGW) Operable Unit (OU) Scoping Meeting held in January 2017 to discuss recent data and potential alternatives in support of a focused Corrective Measures Study/Feasibility Study (CMS/FS). This white paper presents an overview of the problem, and a range of technical and administrative options for addressing the tritium contamination in groundwater and Steel Creek. As tritium cannot be treated practicably, alternatives are limited to media transfer, containment and natural attenuation principally relying on radioactive decay. Using other groundwater OU decisions involving tritium as precedent, Savannah River Nuclear Solutions (SRNS) recommends that final tritium alternatives be evaluated in a CMS/FS, understanding that the likely preferred remedy will include natural attenuation with land use controls (LUCs). This is based on the inability to significantly reduce tritium impact to Steel Creek using an engineered solution as compared to natural attenuation. The timing of this evaluation could be conducted concurrently with the final remedy evaluation for volatile organic compounds (VOCs).

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS2002

Potential Operating Orbits for the SAFE-400

Safety must be ensured during all phases of space fission system design, development, fabrication, launch, operation, and shutdown. One potential space fission system application is fission electric propulsion (FEP), in which fission energy is converted into electricity and used to power high efficiency (Isp is greater than 3000s) electric thrusters. For these types of systems it is important to determine which operational scenarios ensure safety while allowing maximum mission performance and flexibility. Space fission systems are essentially non-radioactive at launch, prior to extended operation at high power. Once high power operation begins, system radiological inventory steadily increases as fission products build up. For a given fission product isotope, the maximum radiological inventory is typically achieved once the system has operated for a length of time equivalent to several half-lives. After that time, the isotope decays at the same rate it is produced, and no further inventory builds in. For an FEP mission beginning in Earth orbit, altitude and orbital lifetime increase as the propulsion system operates. Two simultaneous effects of fission propulsion system operation are thus (1) increasing fission product inventory and (2) increasing orbital lifetime. Phrased differently, as fission products build up, more time is required for the fission products to naturally convert back into non-radioactive isotopes. Simultaneously, as fission products build up, orbital lifetime increases, providing more time for the fission products to naturally convert back into non-radioactive isotopes. Operational constraints required to ensure safety can thus be quantified.

Houts, Mike↗
DOE OSTI2016

Implications of Grain-Scale Mineralogical Heterogeneity for Radionuclide Transport in Fractured Media

The geological disposal of nuclear waste is based on the multi-barrier concept, comprising various engineered and natural barriers, to confine the radioactive waste and isolate it from the biosphere. Some of the planned repositories for high-level nuclear waste will be hosted in fractured crystalline rock formations. The potential of these formations to act as natural transport barriers is related to two coupled processes: diffusion into the rock matrix and sorption onto the mineral surfaces available in the rock matrix. Different in situ and laboratory experiments have pointed out the ubiquitous heterogeneous nature of the rock matrix: mineral surfaces and pore space are distributed in complex microstructures and their distribution is far from being homogeneous (as typically assumed by Darcy-scale coarse reactive transport models). We use a synthetically generated fracture–matrix system to assess the implications of grain-scale physical and mineralogical heterogeneity on cesium transport and retention. The resulting grain-scale reactive transport model is solved using high-performance computing technologies, and the results are compared with those derived from two alternative models, denoted as upscaled models, where mineral abundance is averaged over the matrix volume. In the grain-scale model, the penetration of cesium into the matrix is faster and the penetration front is uneven and finger-shaped. The analysis of the cesium breakthrough curves computed at two different points in the fracture shows that the upscaled models provide later first-arrival time estimates compared to the grain-scale model. The breakthrough curves computed with the three models converge at late times. These results suggest that spatially averaged upscaled parameters of sorption site distribution can be used to predict the late-time behavior of breakthrough curves but could be inadequate to simulate the early behavior.

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

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Comparative assessment of status and opportunities for carbon Dioxide Capture and storage and Radioactive Waste Disposal In North America

Aside from the target storage regions being underground, geologic carbon sequestration (GCS) and radioactive waste disposal (RWD) share little in common in North America. The large volume of carbon dioxide (CO{sub 2}) needed to be sequestered along with its relatively benign health effects present a sharp contrast to the limited volumes and hazardous nature of high-level radioactive waste (RW). There is well-documented capacity in North America for 100 years or more of sequestration of CO{sub 2} from coal-fired power plants. Aside from economics, the challenges of GCS include lack of fully established legal and regulatory framework for ownership of injected CO{sub 2}, the need for an expanded pipeline infrastructure, and public acceptance of the technology. As for RW, the USA had proposed the unsaturated tuffs of Yucca Mountain, Nevada, as the region's first high-level RWD site before removing it from consideration in early 2009. The Canadian RW program is currently evolving with options that range from geologic disposal to both decentralized and centralized permanent storage in surface facilities. Both the USA and Canada have established legal and regulatory frameworks for RWD. The most challenging technical issue for RWD is the need to predict repository performance on extremely long time scales (10{sup 4}-10{sup 6} years). While attitudes toward nuclear power are rapidly changing as fossil-fuel costs soar and changes in climate occur, public perception remains the most serious challenge to opening RW repositories. Because of the many significant differences between RWD and GCS, there is little that can be shared between them from regulatory, legal, transportation, or economic perspectives. As for public perception, there is currently an opportunity to engage the public on the benefits and risks of both GCS and RWD as they learn more about the urgent energy-climate crisis created by greenhouse gas emissions from current fossil-fuel combustion practices.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

On-site Destruction of Radioactive Oily Wastes Using Adsorption Coupled with Electrochemical Regeneration - 12221

Arvia{sup R}, working with Magnox Ltd, has developed the technology of adsorption coupled with electrochemical regeneration for the degradation of orphan radioactive oil wastes. The process results in the complete destruction of the organic phase where the radioactivity is transferred to liquid and solid secondary wastes that can then be processed using existing authorised on-site waste-treatment facilities.. Following on from successful laboratory and pilot scale trials, a full scale, site based demonstrator unit was commissioned at the Magnox Trawsfynydd decommissioning site to destroy 10 l of LLW and ILW radioactive oils. Over 99% of the emulsified oil was removed and destroyed with the majority of activity (80 - 90%) being transferred to the aqueous phase. Secondary wastes were disposed of via existing routes with the majority being disposed of via the sites active effluent treatment plant. The regeneration energy required to destroy a litre of oil was 42.5 kWh/l oil. This on-site treatment approach eliminates the risks and cost associated with transporting the active waste oils off site for incineration or other treatment. The Arvia{sup R} process of adsorption coupled with electrochemical regeneration has successfully demonstrated the removal and destruction of LLW and ILW radioactive oils on a nuclear site. Over 99.9% of the emulsified oil was removed, with the majority of the radioactive species transferred to the aqueous, supernate, phase (typically 80 - 90 %). The exception to this is Cs-137 which appears to be more evenly distributed, with 43% associated with the liquid phase and 33 % with the Nyex, the remainder associated with the electrode bed. The situation with Plutonium may be similar, but this requires confirmation, hence further work is underway to understand the full nature of the electrode bed radioactive burden and its distribution within the body of the electrodes. - Tritium gaseous discharges were negligible; hence no off-gas treatment before direct discharge to atmosphere is necessary. All secondary wastes were suitable for disposal using existing disposal routes, with the majority of the activity being successfully discharged as active water via the site active drains. - Oil destruction was achieved at a rate of 28.2 ml/hr using a regeneration energy of 42.5 kWh/l oil. The treatment of different active and non-active oils was achieved using the same operating parameters, providing strong evidence that the process is robust and will treat a wide range of oils, organic wastes and additives. - Currently the design of a plant capable of processing 1000 ml/hr is being established in discussion with Magnox Ltd. The plant will run automatically with little operator attention and so process between 5-8 m{sup 3} of ILW oil per annum. (authors)

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

Sorption of Cationic Radionuclides on Chlorite and Sericite formed by the Hydrothermal Alteration - 18180

Chlorite and sericite formed by the hydrothermal alteration of granite distribute predominantly around the geological repository in deep underground. These minerals often cover the surface of a solid phase and fill the fracture of granite considered as the natural barrier to the migration of radionuclides leaching from radioactive waste. Therefore, to assess the practical performance of the natural barrier, it is necessary to clarify the interaction between radionuclides and these minerals. This study examined the sorption of cationic radionuclides by chlorite and sericite through batch experiments, whereby Cs{sup +}, Sr{sup 2+}, and Eu{sup 3+} were used as tracer ions. In these experiments, powder-form chlorite or sericite was mixed with the solutions of the tracer ions to a liquid/solid ratio of 10 ml/g, and the resulting mixture was allowed to incubate for 2-7 days. The concentration of tracer ions was set to 1 mM and that of NaCl to a specific value in the 0-0.6 M range. The solution pH was 3, 5, or 8. As a result, tracer ions were significantly adsorbed by chlorite and sericite, regardless of pH or NaCl concentration, except when NaCl concentration reached the high value of 0.6 M. The distribution coefficient for europium ions at pH 5 was estimated to be 2-7 mL/g in chlorite and 10{sup 3}-10{sup 4} mL/g in sericite. These values significantly exceeded that measured for feldspar (2-4 mL/g), the predominant component of granite before it undergoes any hydrothermal alteration. These results suggest that cationic radionuclides will be immobilized through sorption in chlorite and sericite filling granite fractures, even if these thermal-fluid-formed minerals cover the surface of granite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Intrinsic Radiation Source Generation with the ISC Package: Data Comparisons and Benchmarking

The characterization of radioactive emissions from unstable isotopes (intrinsic radiation) is necessary for shielding and radiological-dose calculations from radioactive materials. While most radiation transport codes, e.g., MCNP [X-5 Monte Carlo Team, 2003], provide the capability to input user prescribed source definitions, such as radioactive emissions, they do not provide the capability to calculate the correct radioactive-source definition given the material compositions. Special modifications to MCNP have been developed in the past to allow the user to specify an intrinsic source, but these modification have not been implemented into the primary source base [Estes et al., 1988]. To facilitate the description of the intrinsic radiation source from a material with a specific composition, the Intrinsic Source Constructor library (LIBISC) and MCNP Intrinsic Source Constructor (MISC) utility have been written. The combination of LIBISC and MISC will be herein referred to as the ISC package. LIBISC is a statically linkable C++ library that provides the necessary functionality to construct the intrinsic-radiation source generated by a material. Furthermore, LIBISC provides the ability use different particle-emission databases, radioactive-decay databases, and natural-abundance databases allowing the user flexibility in the specification of the source, if one database is preferred over others. LIBISC also provides functionality for aging materials and producing a thick-target bremsstrahlung photon source approximation from the electron emissions. The MISC utility links to LIBISC and facilitates the description of intrinsic-radiation sources into a format directly usable with the MCNP transport code. Through a series of input keywords and arguments the MISC user can specify the material, age the material if desired, and produce a source description of the radioactive emissions from the material in an MCNP readable format. Further details of using the MISC utility can be obtained from the user guide [Solomon, 2012]. The remainder of this report presents a discussion of the databases available to LIBISC and MISC, a discussion of the models employed by LIBISC, a comparison of the thick-target bremsstrahlung model employed, a benchmark comparison to plutonium and depleted-uranium spheres, and a comparison of the available particle-emission databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Treatment of radioactive liquid waste by the forced-air exhaust system - 15190

A Natural Evaporation Facility (NEF) was constructed to treat low level (LL) radioactive liquid waste arising from a Radioactive Waste Treatment Facility (RWTF) at the Korea Atomic Energy Research Institute (KAERI). In order to process LL radioactive liquid waste, we investigated the influence of climate condition, mainly atmospheric humidity and temperature, on the amount of evaporation in a forced-air exhaust system in KAERI. Evaporation media were made of cotton and polyester. Air exhaust in the facility was forced by exhaust fans. The evaporation rate and the decontamination factor were calculated and reported as shown in the results section. The evaporation and air supply flow increased at the same rate. In addition, the liquid waste feed rate and temperature of supplied air increased. Regarding the relative humidity of supplied air, the evaporation rate increased as the relative humidity decreased. As the result of this study, the operating conditions for the NEF were optimized in the following areas: The air temperature above 8 deg. C, the air relative humidity below 70%, the air flow rate 1.14-1.47 m/sec, and the liquid waste feed rate 4.6 l/hr.m{sup 2}. The decontamination factor was 1.1x10{sup 3}. (authors)

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

Health Hazards of Exposures to Radioiodine

Iodine is a chemical element with atomic number 53. Iodine-127 is stable (non-radioactive) and commonly found in nature. Elemental iodine is a purple-colored solid at room temperature and pressure, but spontaneously sublimates (turns into vapor). Iodine is an essential element for life, and is required for proper functioning of the thyroid. Iodine is present in many foods, and is readily absorbed by the body and concentrated in the thyroid gland. A fraction of iodine ingested or inhaled is rapidly removed by the kidneys. The rest of the inhaled or ingested iodine is absorbed the by thyroid and retained for many months. Iodine has a biological half-life of approximately 120 days in health individuals. The biological half-life can be shorter in individuals with hyperthyroidism, and longer in individuals with hypothyroidism. Iodine has a number of radioactive isotopes, most of which have relatively short half-lives (days or weeks). Short half-life iodine isotopes are useful for a variety of medical applications, including imaging and cancer therapy. For example, Iodine-123 (half-life 13 hours) is commonly used for medical imaging of the thyroid, while iodine-131 (half-life 8 days) is used for suppressing thyroid function in individuals with hyperthyroidism or ablating (killing) thyroid cells to treat thyroid cancer. Iodine-125 (half-life 59 days) is produced in nuclear reactors, and has medical uses. Although iodine-125 can be used for thyroid imaging, Iodine-123 is more commonly used for that purpose because of its shorter half-life and higher-energy emissions. Iodine-125 is more commonly used for cancer treatment, and can be processed into small metal pellets (seeds) inserted directly into a tumor. Iodine-125 emits low-energy x-rays which can kill tumor cells and generally cannot escape the tumor, sparing other tissues. Medical iodine for imaging or treatment is typically administered orally in the form a pill or liquid solution. A typical adult thyroid scan using iodine-123 involves having the patient swallow between one and four 0.1 millicuries pills, with the exact dose dependent on the patient’s weight. This results in a whole-body committed effective dose of 80 – 320 mrem, and a thyroid equivalent dose of 1443 – 5772 mrem. Note that the whole-body effective dose relates to the overall cancer risk, while the larger equivalent dose to the thyroid only indicates that most of this risk is the result of exposure to the thyroid. These doses are considered safe, although the procedure is not recommended for pregnant or breastfeeding women. In contrast, the quantities of iodine-131 used for treatment of hyperthyroidism and thyroid cancers are much higher. For treatment of hyperthyroidism, 4 – 10 millicuries are administered, while for thyroid cancer the administration can range from 50 – 150 millicuries of I-131. In addition to medical exposures, large populations were exposed to radioiodine as result of the atomic bombings of Hiroshima and Nagasaki in Japan, and the Chernobyl nuclear accident. These populations have been carefully followed for many years to assess the effect of their radiation exposures on cancer risk. As a result, a great deal is known about the cancer risks associated with radioiodine exposure. Because iodine is concentrated in the thyroid, the principal risk of exposure to radioiodine is thyroid cancer. Children have the highest risk of thyroid cancer after exposure to radioactive iodine. According to a large study of Japanese atomic bomb survivors, an effective dose of one Sievert (100,000 mrem) has been observed to increase the risk of thyroid cancer by a factor of 9.5 in children aged zero to nine years old, by a factor of 3 in children aged 10 to 19 years old, and by barely detectable amounts in adults. Another way of quantifying the risk from radioiodine exposure is from risk coefficients, which provide the risk per unit intake of radionuclides in terms of both morbidity (any cancer) and mortality (death). Both morbidity (risk of cancer) and mortality (death) risks are shown in the table below. Note that because thyroid cancer is almost never fatal, the morbidity coefficients are much larger than the mortality coefficients.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Unveiling the Structure and Dynamics of Water Confined in Colloidal Boehmite Suspensions

Thin fluid layers confined between nanoparticles play an important role in several natural and industrial systems, including radioactive wastes stored in tanks at the U.S. Department of Energy’s Hanford site. Multimodal neutron and computational approaches have been integrated to examine the properties of one, two, and four layers of water (H 2 O or D 2 O) on nanoparticulate, hydrous, or deuterated boehmite (γ-AlOOH or γ-AlOOD). Exposure of deuterated boehmite to H 2 O at 90 °C yielded rapid H/D isotopic exchange likely driven by a Grotthus-like proton-hopping mechanism. The in-plane and out-of-plane vibrations of the structural hydroxyls involved in this exchange were observed for both the hydrated and deuterated boehmite. These observations were confirmed by molecular dynamics simulations that also showed that single water molecules on the (010) surface bond to the structure via four bonds: two from hydrogens (deuteriums) on the water to surrounding oxygens and two from the water’s oxygen to surrounding OD/OH. Bulk water-like properties began to appear once four monolayers of water had been added, but steric crowding limited water diffusion rates once two monolayers had been added. The super-Arrhenius temperature dependence observed at four monolayers indicated glass-like behavior in a well-formed hydrogen-bonding network. Such a network is not sufficiently developed, however, when the surface water coverage is less than four layers. In conclusion, the unique nature of these layers can provide critical information for understanding forces between particles in proximity, and resultant effects on suspension rheology.

Anovitz, Lawrence M. [Oak Ridge National Laborator↗
DOE OSTI2012

Low Pore Connectivity in Natural Rock

As repositories for CO₂ and radioactive waste, as oil and gas reservoirs, and as contaminated sites needing remediation, rock formations play a central role in energy and environmental management. The connectivity of the rock's porespace strongly affects fluid flow and solute transport. This work examines pore connectivity and its implications for fluid flow and chemical transport. Three experimental approaches (imbibition, tracer concentration profiles, and imaging) were used in combination with network modeling. In the imbibition results, three types of imbibition slope [log (cumulative imbibition) vs. log (imbibition time)] were found: the classical 0.5, plus 0.26, and 0.26 transitioning to 0.5. The imbibition slope of 0.26 seen in Indiana sandstone, metagraywacke, and Barnett shale indicates low pore connectivity, in contrast to the slope of 0.5 seen in the well-connected Berea sandstone. In the tracer profile work, rocks exhibited different distances to the plateau porosity, consistent with the pore connectivity from the imbibition tests. Injection of a molten metal into connected pore spaces, followed by 2-D imaging of the solidified alloy in polished thin sections, allowed direct assessment of pore structure and lateral connection in the rock samples. Pore-scale network modeling gave results consistent with measurements, confirming pore connectivity as the underlying cause of both anomalous behaviors: imbibition slope not having the classical value of 0.5, and accessible porosity being a function of distance from the edge. A poorly connected porespace will exhibit anomalous behavior in fluid flow and chemical transport, such as a lower imbibition slope (in air–water system) and diffusion rate than expected from classical behavior.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2018

NORM in Building Materials: A Case Study in Releasing and Disposing of Building Debris within the DOE System - 18123

In 2013, the Office of Management and Budget issued Memorandum 2013-02, clarifying its 'Freeze the Footprint' policy. Memorandum 2013-02 requires that all federal agencies maintain or reduce their total square footage, within each subsequent year, from a FY2012 baseline. Memorandum 2013-02 also requires that total square footage remain constant with respect to the type/intended use of a building. At Los Alamos National Laboratory (LANL), this means that new construction cannot occur without demolition of equivalent (both in type and square footage) building(s). Since 2013, decommissioning and demolition of aging, unused facilities at LANL has increased. In 2016, the TA16-280 building complex was scheduled for demolition. TA16-280 is a former high explosives (HE) processing facility that consists of six buildings connected to each other via weatherproof walkways. It was constructed in the 1950's and has been subsequently used for HE shipping, receiving, and storage. Interviews with radiation protection staff suggest that some of the buildings within the complex may have been used to store depleted uranium metal parts in a non-dispersible form. It is LANL policy to conduct decommissioning surveys for radioactive materials when building history cannot confirm that a structure is radiologically non-impacted. LANL follows the Multi-Agency Radiation Survey and Assessment of Materials and Equipment (MARSAME) process to develop its surveys and clear structures from radiological control. After clearance, a structure may be dispositioned as non-regulated waste, at a significant cost savings compared to regulated disposition pathways such as low-level waste. MARSAME measurements conducted in the 16-280 complex were elevated compared to reference measurements of surrogate materials, and a majority of the survey units failed to pass indistinguishable from background (IFB), which is LANL's unregulated clearance criteria. Further measurements conducted both in situ and in the laboratory indicated the presence of naturally-occurring thorium in several building materials including: poured concrete, the glaze atop ceramic blast tiles, and the porcelain enamel on the exterior of blast-proof ceiling lamps. Other than naturally-occurring thorium, there were no elevated radionuclide concentrations detected in building materials. It should be noted that there was no history of thorium use in the 16-280 complex, and the material in question was in all cases fixed to and within the building materials in question. Except in certain specific cases, such material is exempt from regulation under the Atomic Energy Act of 1954 [3], as amended, and is therefore exempt from regulation under the Nuclear Regulatory Commission, the Department of Energy, and agreement state regulatory frameworks. Despite this regulatory exemption, disposition of material from the 16-280 complex proved to be challenging, principally because local industrial landfills stipulate 'No radioactive material' in their Waste Acceptance Criteria. This disconnect between what is regulated and what is disposable creates a situation where some DOE sites find it easier to disposition property as Low Level Waste in lieu of pursuing a non-regulated disposition pathway. This presentation will explore the specifics of the 16-280 decommissioning project as well as the implications of the release/disposition disconnect for naturally occurring radioactive material and project planning at LANL. (authors)

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

Regulatory framework for NORM residues in Belgium

The Belgian radiation protection authority (Federal Agency for Nuclear Control - FANC) has published in March 2013 a decree regulating the acceptance of NORM residues by nonradioactive waste treatment facilities. This regulation is based on the concept of 'work activities involving natural radiation sources' in the sense of article 40 of the 96/29/EURATOM directive. The disposal or processing facilities which accept NORM residues with an activity concentration above a generic exemption level will be considered as 'work activities' and submitted to declaration according to the Belgian radiation protection regulations. On basis of this declaration, specific acceptance criteria for the different types of processing/ disposal of the residues (disposal on landfill, recycling into building materials, etc.) are imposed. FANC has drafted guidelines for these acceptance criteria. A methodological guide for the operators of the concerned facilities was also published. Moreover, sites where significant quantities of NORM residues are or have been disposed, are subjected to an environmental monitoring in the framework of the national program of radiological surveillance of FANC. FANC also introduced in its regulations the concept of anthropogenic radon-prone areas: e.g. former phosphogypsum stacks have been defined as anthropogenic radon-prone areas, which allows some form of regulatory control of these sites. (authors)

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

Impact of lowering potassium contamination in liquid scintillation cocktails for ultra-sensitive radiation detection

Intrinsic 40 K radioactive backgrounds from impurities of natural K in liquid scintillation cocktails have previously been demonstrated to limit their use in ultra-sensitive applications. Here, this work explores two methodologies in parallel for the reduction of 40 K backgrounds in the cocktails, and lays the groundwork for use in ultra-sensitive applications. In one method, alternative low-K liquid scintillation matrix constituents were identified and in the other, a simple purification method for single components and finished cocktails was developed. Both methods were verified via ICP-MS analysis. Liquid scintillation counting of selected purified cocktails demonstrated background reduction, improved stability, and enhanced performance. The best performing purified cocktail was also counted on a custom-built ultra-low background liquid scintillation counter, with results below the detector background.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2012

The limits of the nuclear landscape

In 2011, 100 new nuclides were discovered1. They joined the approximately 3,000 stable and radioactive nuclides that either occur naturally on Earth or are synthesized in the laboratory2,3. Every atomic nucleus, characterized by a specific number of protons and neutrons, occupies a spot on the chart of nuclides, which is bounded by ‘drip lines’ indicating the values of neutron and proton number at which nuclear binding ends. The placement of the neutron drip line for the heavier elements is based on theoretical predictions using extreme extrapolations, and so is uncertain. However, it is not known how uncertain it is or how many protons and neutrons can be bound in a nucleus. Here we estimate these limits of the nuclear ‘landscape’ and provide statistical and systematic uncertainties for our predictions. We use nuclear density functional theory, several Skyrme interactions and high-performance computing, and find that the number of bound nuclides with between 2 and 120 protons is around 7,000. We find that extrapolations for drip-line positions and selected nuclear properties, including neutron separation energies relevant to astrophysical processes, are very consistent between the models used.

Science & Technology - Other Topics↗
DOE OSTI2012

Argonne National Laboratory Site Environmental Report for Calendar Year 2011

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2011. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP-88 Version 3 computer code, was used in preparing this report.

Davis, Theresa M.↗
DOE OSTI2014

Argonne National Laboratory Site Environmental Report for Calendar Year 2013

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2013. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP-88 Version 3 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗