Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE DECAY”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 181 records · Page 10

Thermal release of {sup 3}He from tritium aged LaNi{sub 4.25}Al{sub 0.75} hydride

The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 (LaNi{sub 4.25}Al{sub 0.75})in the tritium process to store hydrogen isotopes. The vast majority of {sup 3}He born from the radioactive decay of tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah River National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490 C. degrees. The first peak consisted of both {sup 3}He and residual hydrogen isotopes, the second was primarily {sup 3}He. The bulk of the gas was released by 600 Celsius degrees. (author)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Direct Mass Measurements to Inform the Behavior of 128m $\mathrm{Sb}$ in Nucleosynthetic Environments

Nuclear isomer effects are pivotal in understanding nuclear astrophysics, particularly in the rapid neutron-capture process where the population of metastable isomers can alter the radioactive decay paths of nuclei produced during astrophysical events. The β-decaying isomer 128m Sb was identified as potentially impactful since the β-decay pathway along the A = 128 isobar funnels into this state bypassing the ground state. Here we report the first direct mass measurements of the 128 Sb isomer and ground state using the Canadian Penning Trap mass spectrometer at Argonne National Laboratory. We find mass excesses of -84564.8(25) keV and -84608.8(21) keV, respectively, resulting in an excitation energy for the isomer of 43.9(33) keV. These results provide the first key nuclear data input for understanding the role of 128m Sb in nucleosynthesis, and we show that it will influence the flow of the rapid neutron-capture process.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Alpha-Voltaic Sources Using Liquid Ga as Conversion Medium

A family of proposed miniature sources of power would exploit the direct conversion of the kinetic energy of alpha particles into electricity. In addition to having long operational lives, these sources are expected to operate with energy-conversion efficiencies from 70 to 90 percent. A power source as proposed (see figure) would be an electrolytic cell in which liquid gallium would serve as both an electrolyte and an energy-conversion medium. The cell would contain an iridium cathode and a zirconium anode. The alpha particles, each with a kinetic energy approx.5.8 MeV, would be emitted by radioactive decay of Cm-244, which has a half-life of 18 years. The Cm-244 source would be positioned so that the a particles would enter the liquid gallium, where their kinetic energy would be dissipated mostly through ionization of Ga atoms, creating Ga(+) ions and free electrons. The electrons would be collected by iridium cathode, and the Ga(+) ions would be neutralized at the zirconium cathode by electrons returning after flowing through an external circuit. Gallium is a candidate for use as the electrolyte and the energy-conversion medium because in the liquid state it is a semimetal: its electrical conductivity is greater than that of a typical semiconductor but small in comparison with the conductivities of metals. Consequently, in liquid gallium, electrons and Ga(+) can exist without immediate recombination and can be moved by electric fields. It is expected that electric fields, resulting at least partly from the difference between the work functions of the electrode metals, would move the electrons and ions to their respective electrodes. The open-circuit potential of the cell is expected to be 1.62 V - equal to the difference between the work functions of iridium and zirconium. Unlike in a solid-state energy conversion medium, the impingement of energetic a particles would not give rise to displacement damage in the liquid gallium. Hence, the cell should have a long life, limited only by the half-life of Cm-244. A cell having a volume less than 25 cu mm, containing 1 curie of Cm-244 (the curie is a unit of radioactivity equal to 3.7 10(exp 10) disintegrations per second) is expected to deliver a current between 7 and 12 mA, which, at the expected open-circuit potential, would provide a power in the approximate range of 11 to 20 mW.

Patel, Jagdish U.↗

Effects of bentonite heating on U(VI) adsorption

Engineered barrier systems designed to isolate high-level radioactive waste utilize bentonite, a montmorillonite rich material, to restrict contaminant transport due to its low hydraulic conductivity and high adsorption capacity. High temperatures (100–200 °C) near waste canisters resulting from radioactive decay may alter the clay's ability to adsorb contaminants. In this study, we examine U(VI) adsorption onto two different bentonite samples subjected to (1) 18 years of in situ heating during an underground experiment and (2) short-term (7-week), high temperature (300 °C) heating in the laboratory. Results show that U(VI) adsorption was lower for field-heated bentonite located closest to the heater, which experienced temperatures of approximately 95 °C, compared to a control sample, which experienced temperatures of approximately 20 °C, over a range of aqueous chemical conditions. On average, K d values for U(VI) adsorption were 31% lower for 95 °C heated samples. By contrast, U(VI) adsorption onto intermediate-heated bentonite (50 °C) was indistinguishable from the adsorption onto the cold-zone (20 °C) sample. U(VI) adsorption onto lab-heated bentonite was also lower than onto control bentonite over the pH range 4.5–8.0, with an average decrease in K d values of 50% after heating. Lower U(VI) adsorption to field-heated bentonite persisted after bentonite was purified to isolate the clay fraction. This allows us to rule out changes in pore-water chemistry or accessory mineral composition as causes of the lower adsorption. No evidence of montmorillonite illitization was observed in the heated samples. While some of the lower U(VI) adsorption in the lab-heated bentonite can be explained by changes in aqueous U(VI) speciation, we propose that lower U(VI) adsorption to field-heated bentonite may be primarily due to changes in the montmorillonite edge structure. The observed changes in U(VI) adsorption to bentonite after heating have implications for U(VI) diffusive transport through engineered barriers and must be considered when designing radioactive waste disposal repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The DISTINCTIVE University Consortium: Theme 2: PuO{sub 2} and Fuel Residues - 18132

The safe and secure management of Pu is a matter of serious international concern, with ∼250 t of separated Pu currently stockpiled worldwide. Nearly 50% of this material is in long-term storage in the UK whilst the British government develops its options for final treatment and disposition. Options for this are: (i) reuse as a fuel in modern reactors (e.g. as MOX fuel or within a fast reactor) prior to disposal; (ii) prompt immobilisation for disposal as soon as is practicable; or, (iii) continued long term storage (prior to disposition). The focus of the work presented here is options (ii) and (iii). The R and D needs of both are now pressing: in option (iii) due to it being the current default; in option (ii) because of a comparative lack of R and D on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on characterisation methods for Pu bearing materials. In response to these needs, the DISTINCTIVE Consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories, a multi-disciplinary collaboration of 11 universities and 3 key industry partners from across the UK's civil nuclear sector) has been seeking to: - Understand how the structure and properties of PuO{sub 2} change with time in the presence of H{sub 2}O. - Understand the roles these processes play in gaseous product evolution from PuO{sub 2} in storage. - Understand radiation-induced amorphization and dissolution kinetics of Pu waste forms. - Develop novel, fast neutron based radiometric methods for the quantification, isotopic composition assessment and remote imaging of Pu-bearing materials. This paper describes recent progress made towards achieving these objectives. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Backfilling of Geological Disposal Facilities - Development of Optimized Backfill Material - 18146

In Belgium, low- and intermediate-level waste (B-waste) will be conditioned in concrete monoliths and high-level waste (C-waste) in so-called Super-containers (SCs). The Belgian radioactive waste management program has focused on the disposal of these waste types in two emplacement fields of a geological disposal facility (GDF). When the disposal waste packages (DWP) have been emplaced, backfilling of the galleries is supposed to be carried out in order to stabilize the galleries, to reduce voids, to guarantee a favourable chemical milieu with regard to the retention of the radionuclides, to minimize a possible gas pressure build-up, to dissipate radioactive decay heat and to enable good conditions for a potential retrieval of the DWP. The current reference solution is to mix the backfill at the surface and then to pump the mixture through pipeline distribution systems into the underground structures. The catalogue of backfill material requirements comprises common, but also different requirements due to the characteristics of the B and C waste packages and differences in the backfilling systems. It was found that two backfill material recipes have to be developed. Different requirements, such as high porosity (low strength and low gas threshold pressure) and high thermal conductivity have contrary consequences for the material. In addition, during backfill production quality variations of the raw materials that influence the backfill material properties can occur. Different lengths of the delivery pipeline may require adjustments of the material composition. In order to consider the practical experiences of backfill material production, one focus of the development programs was to identify compositional ranges that guarantee the respective material property. A combination of the individual 'conformity fields' results in optimal solutions for the standard operating conditions as well as a compositional range that guarantees compliance with the required material specifications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Long-term stewardship - surveillance and maintenance at Hanford - 15155

In FY 2015, Hanford's Long-Term Stewardship (LTS) Program is tasked to complete the required entries into five cocooned plutonium production reactors (105-C, 105-D, 105-F, 105-H and 105-N), which sit adjacent to the Columbia River. The first reactor entered in October 2014, the cocooned plutonium production reactor, 105-F, signaled the LTS Program's entry into the Surveillance and Maintenance (S and M) era. Reactor entries at Hanford are conducted at 5-year intervals to assess the condition of the structures and evaluate potential deterioration of the reactor core, shield walls and roof. The cocooned structures are designed to protect the reactor core for up to 75 years while radioactive decay continues, ultimately making the structures safe for demolition and removal. The 105-F entry demonstrates that the LTS program at Hanford is shifting from a program focused on transitioning cleaned up land and waste sites to a program focused primarily on S and M activities. The entry process is carefully planned, choreographed and monitored to ensure safety of the workers and minimize the time the reactor structure is opened. The process includes controlled opening of the reactor, air quality monitoring, structural safety inspections, and radiological clearance. Once the structure is cleared for entry, assessments of the structural integrity and radiologic conditions are completed. Evaluation of the interior temperature and flood monitoring equipment is conducted. Facility visits by regulators and DOE personnel close out the process prior to welding the door closed until the next assessment, five years later. This paper will document execution of the entry work package, identify challenges overcome, and provide video documentation as reference material for future workers and the public. The current LTS program has responsibility for approximately 100,000 acres, over 200 waste sites and six cocooned reactors. Just over 5,600 documents have been identified, indexed and tagged for storage in the LTS document library. The program has successfully completed seven consecutive transitions since July 2011, with six more transitions planned in the next 18 months to support the U.S. Department of Energy Richland Operations Office's (DOE-RL) near-term cleanup objectives to significantly reduce the footprint of active cleanup operations for the River Corridor. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Variation of lead isotopic composition and atomic weight in terrestrial materials (IUPAC Technical Report)

The isotopic composition and atomic weight of lead are variable in terrestrial materials because its three heaviest stable isotopes are stable end-products of the radioactive decay of uranium ( 238 U to 206 Pb; 235 U to 207 Pb) and thorium ( 232 Th to 208 Pb). The lightest stable isotope, 204 Pb, is primordial. These variations in isotope ratios and atomic weights provide useful information in many areas of science, including geochronology, archaeology, environmental studies, and forensic science. While elemental lead can serve as an abundant and homogeneous isotopic reference, deviations from the isotope ratios in other lead occurrences limit the accuracy with which a standard atomic weight can be given for lead. In a comprehensive review of several hundred publications and analyses of more than 8000 samples, published isotope data indicate that the lowest reported lead atomic weight of a normal terrestrial materials is 206.1462 ± 0.0028 (k = 2), determined for a growth of the phosphate mineral monazite around a garnet relic from an Archean high-grade metamorphic terrain in north-western Scotland, which contains mostly 206 Pb and almost no 204 Pb. The highest published lead atomic weight is 207.9351 ± 0.0005 (k = 2) for monazite from a micro-inclusion in a garnet relic, also from a high-grade metamorphic terrain in north-western Scotland, which contains almost pure radiogenic 208 Pb. When expressed as an interval, the lead atomic weight is [206.14, 207.94]. It is proposed that a value of 207.2 be adopted for the single lead atomic-weight value for education, commerce, and industry, corresponding to previously published conventional atomic-weight values.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

25 Years Of Environmental Remediation In The General Separations Area Of The Savannah River Site: Lessons Learned About What Worked And What Did Not Work In Soil And Groundwater Cleanup

The Savannah River Site (SRS) is owned and administered by the US Department of Energy (DOE). SRS covers an area of approximately 900 square kilometers. The General Separation Area (GSA) is located roughly in the center of the SRS and includes: radioactive material chemical separations facilities, radioactive waste tank farms, a variety of radioactive seepage basins, and the radioactive waste burial grounds. Radioactive wastes were disposed in the GSA from the mid-1950s through the mid-1990s. Radioactive operations at the F Canyon began in 1954; radioactive operations at H Canyon began in 1955. Waste water disposition to the F and H Seepage Basins began soon after operations started in the canyons. The Old Radioactive Waste Burial Ground (ORWBG) began operations in 1952 to manage solid waste that could be radioactive from all the site operations, and ceased receiving waste in 1972. The Mixed Waste Management Facility (MWMF) and Low Level Radioactive Waste Disposal Facility (LLRWDF) received radioactive solid waste from 1969 until 1995. Environmental legislation enacted in the 1970s, 1980s, and 1990s led to changes in waste management and environmental cleanup practices at SRS. The US Congress passed the Clean Air Act in 1970, and the Clean Water Act in 1972; the Resource Conservation and Recovery Act (RCRA) was enacted in 1976; the Comprehensive Environmental Response Compensation, and Liability Act (CERCLA) was enacted by Congress in 1980; the Federal Facilities Compliance Act (FFCA) was signed into law in 1992. Environmental remediation at the SRS essentially began with a 1987 Settlement Agreement between the SRS and the State of South Carolina (under the South Carolina Department of Health and Environmental Control - SCDHEC), which recognized linkage between many SRS waste management facilities and RCRA. The SRS manages several of the larger groundwater remedial activities under RCRA for facilities recognized early on as environmental problems. All subsequent environmental remediation projects tend to be managed under tri-party agreement (DOE, Environmental Protection Agency, and SCDHEC) through the Federal Facilities Agreement. During 25 years of environmental remediation SRS has stabilized and capped seepage basins, and consolidated and capped waste units and burial grounds in the GSA. Groundwater activities include: pump and treat systems in the groundwater, installation of deep subsurface barrier systems to manage groundwater flow, in situ chemical treatments in the groundwater, and captured contaminated groundwater discharges at the surface for management in a forest irrigation system. Over the last 25 years concentrations of contaminants in the aquifers beneath the GSA and in surface water streams in the GSA have dropped significantly. Closure of 65 waste sites and 4 RCRA facilities has been successfully accomplished. Wastes have been successfully isolated in place beneath a variety of caps and cover systems. Environmental clean-up has progressed to the stage where most of the work involves monitoring, optimization, and maintenance of existing remedial systems. Many lessons have been learned in the process. Geotextile covers outperform low permeability clay caps, especially with respect to the amount of repairs required to upkeep the drainage layers as the caps age. Passive, enhanced natural processes to address groundwater contamination are much more cost effective than pump and treat systems. SRS operated two very large pump and treat systems at the F and H Seepage Basins to attempt to limit the release of tritium to Fourmile Branch, a tributary of the Savannah River. The systems were designed to extract contaminated acidic groundwater, remove all contamination except tritium (not possible to remove the tritium from the water), and inject the tritiated groundwater up-gradient of the source area and the plume. The concept was to increase the travel time of the injected water for radioactive decay of the tritium. The two systems were found to be non-effective and potentially mobilizing more contamination. SRS invested approximately $50 million in construction and approximately $100 million in 6 years of operation. The H Seepage Basin pump and treat system was replaced by a series of subsurface barriers that alters the groundwater velocity; the F Seepage Basin pump and treat system was replaced by subsurface barriers forming a funnel and gate augmented by chemical treatment within the gates. These replacement systems are mostly passive and cost approximately $13 million to construct, and have reduced the tritium flux to Fourmile Branch, in these plumes, by over 70%. SRS manages non-acidic tritiated groundwater releases to Fourmile Branch from the southwest plume of the MWMF with a forest irrigation system. Tritiated water is captured with a sheetpile dam below the springs that caused releases to Fourmile Branch. Water from the irrigation pond is pumped to a filter plant prior to irrigation of approximately 26 hectares of mixed forest and developing pine plantation. SRS has almost achieved a 70% reduction in tritium flux to the Branch from this plume. The system cost approximately $5 million to construct with operation cost of approximately $500K per year. In conclusion, many lessons have been learned in 25 years of relatively aggressive remedial activities in the GSA. Geotextile covers outperform low permeability clay caps, especially with respect to the amount of repairs required to upkeep the drainage layers as the caps age. Passive, enhanced natural processes to address groundwater contamination are much more cost effective than pump and treat systems. In water management situations with non-accumulative contaminants (tritium, VOCs, etc.) irrigation in a forest setting can be very effective.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Nature and State of Groundwater Contamination at the Nevada National Security Site: What Have We Learned from Decades of Groundwater Analysis? - 20337

The regulatory framework for remediating radionuclide contamination from underground nuclear testing at the Nevada National Security Site (NNSS) is based on a combination of characterization and modeling studies, monitoring, and institutional controls [1]. Currently, tritium is the largest contributor (∼90%) to the estimated 44.6 million-curie radionuclide inventory resulting from underground testing [2]. Because of its short half-life (12.32 years), its relative contribution reduces below 10% of the total radiologic inventory over the next 120 years as a result of radioactive decay. Although tritium levels are observed well above the Safe Drinking Water Act (SDWA) maximum contaminant levels (MCLs) in groundwater, other radionuclides are well below their MCLs except within the nuclear test near-field (nuclear test cavity and chimney) environment. In fact, most device-derived radionuclides are below their MCL in groundwater even in samples collected from this near-field environment. The distribution of radionuclides following the nuclear detonation greatly influences the availability of potential contaminants for groundwater transport. Tritium is initially distributed in the gas phase, later as tritiated water in steam, and finally as liquid water, and is available to groundwater transport away from the near-field environment. Other radionuclides that are mobile in groundwater are {sup 14}C, {sup 36}Cl, {sup 99}Tc, and {sup 129}I though their radiologic inventory is small when compared to tritium. Many radionuclides (e.g. U, Pu, Am) are incorporated to a significant extent into the melt glass at the bottom of the cavity and are accessible to groundwater primarily through the slow process of glass dissolution. These radionuclides are also adsorbed to the surfaces of the crushed rock within the cavity and chimney which limits their migration in groundwater. Although colloid facilitated transport of radionuclides at the NNSS has been observed [3], radionuclide concentrations decrease with time and migration distances due to desorption and colloid filtration processes. Current studies indicate that radionuclides associated with colloids are unlikely to migrate downgradient from NNSS underground nuclear tests at concentrations above the SDWA MCL [4][5]. The results of over 50 years of sampling, along with an understanding of these post detonation processes, indicate that tritium is the only contaminant of concern downgradient of testing and that even tritium will not exceed its MCL in groundwater after ∼120 years. Although other longer-lived radionuclides may continue to be released slowly from the near-field environment they will likely never reach levels exceeding their MCLs in groundwater downgradient of the NNSS. Groundwater monitoring will continue to verify these observations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Bradwell - The first UK Nuclear Site to Enter Care and Maintenance - 18152

Magnox Ltd, a Cavendish Fluor company, is currently decommissioning twelve nuclear sites on behalf of the UK Nuclear Decommissioning Authority (NDA). Ten of these sites are former Power Stations; the other two are closed Research and Development facilities. An overall strategy for decommissioning the Magnox sites has been agreed between Magnox and the NDA. For all but one of these sites the decommissioning strategy involves an extended quiescent period, known as Care and Maintenance (C and M), during which remaining radiological hazards reduce naturally through radioactive decay. After this quiescent period the sites will be restored to an agreed end state through further decommissioning and clearance work. Sites in C and M will not be permanently manned. They will be passively safe such that prompt action will not be required in response to an alarm or fault. Preparing a nuclear site for C and M involves several inter-related strands of work including definition of the intended physical configuration of the site during C and M (the C and M entry state), establishing a safety case to support that configuration, carrying out the physical works needed to deliver the C and M entry state, preparing arrangements and a team to manage the site during C and M and consolidating the knowledge and information needed to support C and M and final clearance of the site. To enter C and M there needs to be a carefully controlled transition of site, personnel and management system into the agreed C and M configuration. Bradwell, a site on the east coast of England, was selected as part of the overall decommissioning strategy for 'early' entry into C and M; Bradwell is scheduled to enter C and M in late 2018, several years in advance of the next scheduled site, with the intent of carrying learning forward from Bradwell to the subsequent sites. Bradwell is now in the final stages of preparation for C and M. Decommissioning commenced on completion of defueling in 2005. A clearly defined C and M entry state was established early in decommissioning and, through interaction with Regulators and internal and external stakeholders, developed as a shared target, thus providing certainty to the scope of the overall decommissioning project. The site has subsequently been systematically decommissioned to meet this target C and M entry state. All legacy waste has been retrieved and packaged into robust disposal containers, the reactor buildings have a new weather envelope, and much of the site is empty space with concrete hardstanding where facilities used to exist and bare rooms where plant used to operate. The remaining radiological and environmental hazards are not significant. During C and M, which is anticipated to last for 70 years, Bradwell will be managed remotely without the requirement for personnel to be present on site, with the possible exception of an ongoing security presence when the site first enters C and M. In terms of UK decommissioning, many of the projects undertaken at Bradwell were first of a kind. The company operates under a programmed organisational and delivery model, which will enable innovation and learning from Bradwell to be carried forward across the whole Magnox fleet. In addition to decommissioning tools and techniques, processes to manage decommissioning, optimise the C and M entry state to reduce the burden on the UK taxpayer and to transition a site into C and M have been developed. This, in addition to taking people from Bradwell and seeding them into Magnox and the broader UK nuclear industry, should help to provide long term cost savings and delivery efficiencies across the entire NDA estate. This paper describes the journey of Bradwell Site through decommissioning and into C and M, highlighting the innovation and challenges that have contributed to the success of delivering the first UK nuclear site into Care and Maintenance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radiochemistry and nuclear chemistry workforce in the United States

The disciplines of radiochemistry and nuclear chemistry have direct applications in the fields of national security, nuclear medicine, nuclear power production, and environmental management. Although, often, nuclear and radiochemistry are grouped together and many experts work in both areas, the definition for each field is slightly different. For example, radiochemistry may be defined as the application of the phenomena of radioactive decay and techniques common to nuclear physics so as to solve problems in the field of chemistry. In contrast, nuclear chemistry may be defined as the application of procedures and techniques common to chemistry to study the structure of the atomic nucleus. This chapter provides a brief update of the current state of, and critical U.S. needs for, nuclear chemistry and radiochemistry expertise as the Assuring a Future U.S.-Based Nuclear and Radiochemistry Expertise report was published by National Academy of Sciences (NAS) in 2012.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Concept of an upright wearable positron emission tomography imager in humans

Abstract Background Positron Emission Tomography ( PET ) is traditionally used to image patients in restrictive positions, with few devices allowing for upright, brain‐dedicated imaging. Our team has explored the concept of wearable PET imagers which could provide functional brain imaging of freely moving subjects. To test feasibility and determine future considerations for development, we built a rudimentary proof‐of‐concept prototype (Helmet_ PET ) and conducted tests in phantoms and four human volunteers. Methods Twelve Silicon Photomultiplier‐based detectors were assembled in a ring with exterior weight support and an interior mechanism that could be adjustably fitted to the head. We conducted brain phantom tests as well as scanned four patients scheduled for diagnostic F 18‐ FDG PET / CT imaging. For human subjects the imager was angled such that field of view included basal ganglia and visual cortex to test for typical resting‐state pattern. Imaging in two subjects was performed ~4 hr after PET / CT imaging to simulate lower injected F 18‐ FDG dose by taking advantage of the natural radioactive decay of the tracer (F 18 half‐life of 110 min), with an estimated imaging dosage of 25% of the standard. Results We found that imaging with a simple lightweight ring of detectors was feasible using a fraction of the standard radioligand dose. Activity levels in the human participants were quantitatively similar to standard PET in a set of anatomical ROI s. Typical resting‐state brain pattern activation was demonstrated even in a 1 min scan of active head rotation. Conclusion To our knowledge, this is the first demonstration of imaging a human subject with a novel wearable PET imager that moves with robust head movements. We discuss potential research and clinical applications that will drive the design of a fully functional device. Designs will need to consider trade‐offs between a low weight device with high mobility and a heavier device with greater sensitivity and larger field of view.

60 APPLIED LIFE SCIENCES↗

Hybrid optimization and Bayesian inference techniques for a non-smooth radiation detection problem

In this paper, we propose several algorithms to recover the location and intensity of a radiation source located in a simulated 250 × 180 m block of an urban center based on synthetic measurements. Radioactive decay and detection are Poisson random processes, so we employ likelihood functions based on this distribution. Owing to the domain geometry and the proposed response model, the negative logarithm of the likelihood is only piecewise continuous differentiable, and it has multiple local minima. To address these difficulties, we investigate three hybrid algorithms composed of mixed optimization techniques. For global optimization, we consider simulated annealing, particle swarm, and genetic algorithm, which rely solely on objective function evaluations; that is, they do not evaluate the gradient in the objective function. By employing early stopping criteria for the global optimization methods, a pseudo-optimum point is obtained. This is subsequently utilized as the initial value by the deterministic implicit filtering method, which is able to find local extrema in non-smooth functions, to finish the search in a narrow domain. These new hybrid techniques, combining global optimization and implicit filtering address, difficulties associated with the non-smooth response, and their performances, are shown to significantly decrease the computational time over the global optimization methods. To quantify uncertainties associated with the source location and intensity, we employ the delayed rejection adaptive Metropolis and DiffeRential Evolution Adaptive Metropolis algorithms. Finally, marginal densities of the source properties are obtained, and the means of the chains compare accurately with the estimates produced by the hybrid algorithms.

42 ENGINEERING↗

Thermal behavior of polyhalite: a high-temperature synchrotron XRD study

As an accessory mineral in marine evaporites, polyhalite, K 2 MgCa 2 (SO 4 ) 4 ·2H 2 O, coexists with halite (NaCl) in salt formations, which have been considered as potential repositories for permanent storage of high-level nuclear wastes. However, because of the heat generated by radioactive decays in the wastes, polyhalite may dehydrate, and the released water will dissolve its neighboring salt, potentially affecting the repository integrity. Thus, studying the thermal behavior of polyhalite is important. In this paper, a polyhalite sample containing a small amount of halite was collected from the Salado formation at the WIPP site in Carlsbad, New Mexico. To determine its thermal behavior, in situ high-temperature synchrotron X-ray diffraction was conducted from room temperature to 1066 K with the sample powders sealed in a silica-glass capillary. At about 506 K, polyhalite started to decompose into water vapor, anhydrite (CaSO 4 ) and two langbeinite-type phases, K 2 Ca x Mg 2-x (SO 4 ) 3 , with different Ca/Mg ratios. XRD peaks of the minor halite disappeared, presumably due to its dissolution by water vapor. With further increasing temperature, the two langbeinite solid solution phases displayed complex variations in crystallinity, composition and their molar ratio and then were combined into the single-phase triple salt, K 2 CaMg(SO 4 ) 3 , at ~919 K. Rietveld analyses of the XRD data allowed determination of structural parameters of polyhalite and its decomposed anhydrite and langbeinite phases as a function of temperature. Finally, from the results, the thermal expansion coefficients of these phases have been derived, and the structural mechanisms of their thermal behavior been discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Time sequence determination of parent–daughter radionuclides using gamma-spectrometry

The acquisition of time-stamped list (TLIST) data provides additional information useful to gamma-spectrometry analysis. A novel technique is described that uses non-linear least-squares fitting and the Levenberg-Marquardt algorithm to simultaneously determine parent-daughter atoms from time sequence measurements of only the daughter radionuclide. This has been demonstrated for the radioactive decay of short-lived radon progeny (214Pb/214Bi, 212Pb/212Bi) described using the Bateman first-order differential equation. The calculated atoms are in excellent agreement with measured atoms, with a difference of 1.3 – 4.8% for parent atoms and 2.4% - 10.4% for daughter atoms. Measurements are also reported with reduced uncertainty. The technique has potential to redefine gamma-spectrometry analysis.

Burnett, J. L. (ORCID:0000000322586252)↗