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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Preliminary Preclosure Accident Consequence Analysis for the APM Conceptual Design - 18090

The Nuclear Waste Management Organization (NWMO) is implementing Adaptive Phased Management (APM), Canada's plan for the long-term management of used nuclear fuel. The APM approach encompasses centralized containment and isolation of the used fuel in a Deep Geological Repository (DGR) in a suitable rock formation, such as crystalline rock or sedimentary rock, in an informed and willing host. This paper describes a preliminary analysis of the potential dose consequences to the public under accident conditions associated with the operation of a DGR and related surface processing facilities. These results are for a generic site and considers exposure to a person standing at a potential fenceline location under conservative atmospheric conditions. Radiological dose consequences to the public were calculated for accidents identified in the preliminary hazard identification. The presence or absence of ventilation system HEPA filters was also considered in combination with specific accident scenarios. A Gaussian dispersion model considering the most stable atmospheric condition (Pasquill stability category F) was used to estimate the exposure of a member of the public in the direct plume path at the location of maximum airborne radionuclide concentrations. The public dose was calculated via the inhalation, air immersion and ground exposure pathways. For all accidents considered, the calculated public dose was low, up to about 0.8 mSv for a fall of an UFTP with container breach. The total dose to the public was also calculated at various distances from the Used Fuel Packaging Plant, Ventilation Shaft, and Main Shaft. For all accidents considered, the calculated dose to the public was low, below 1 mSv even at the minimum 100 m distance evaluated. Sensitivity cases were also carried out to determine the effect of stack release height, the effluent exit velocity, and the release orientation on the calculated minimum site boundary distance. (authors)

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

The criticality safety program of the Swedish nuclear fuel and waste management Co (SKB)

SKB is co-owned by the owners of the Swedish nuclear power plants (NPP) and is responsible for the back-end of all Swedish spent nuclear fuel and radioactive waste from the Swedish nuclear program. The responsibility starts from when the fuel leaves the NPP and ends at the planned final repository. That means that SKB has the responsible for the safety at the intermediate pool storage (Clab), the encapsulation plant (Clink), the Spent Fuel Repository and the transportation between these sites. In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several complements to the criticality safety analysis in the application and about the same time they also required Clab to modernise the Safety Analysis Report. Our owners (the NPPs) had also requested SKB to take care of higher enrichments and higher burnup fuel. A prerequisite to meet this was extended Burnable Absorber credit (peak reactivity) and Burn-up credit. Up to this stage SKB had relied heavily on consultants to do criticality safety analysis and had no specific criticality safety program or methodology. However it was now evident to SKB that we needed to take an overall view on the criticality safety area and not only lift the quality of the actual criticality analysis but also to handle organisational and competence issues. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

What does the public really want when engaging with the nuclear industry? - 15087

The nuclear industry continues to evolve how it engages the public. A question to constantly ask is, 'What does the public really want when engaging with the nuclear industry?' This paper helps answer this question most specifically in the area of nuclear waste management, disposal and remediation. It offers multiple theories from the social sciences, most notably the theory of 'Dominating Concepts'. This is the theory that ideas that dominate the general thinking in an industry exert a powerful influence on how people think and act. By examining an industry's Dominating Concepts, it becomes possible to recognize the unintended (undesirable) consequences. This awareness can be a catalyst for the adoption of changes to the general thinking --- of new and more useful Dominating Concepts. In this paper, one Dominating Concept widely held in the nuclear industry is examined and it becomes possible to understand why the public may not respond to scientific explanations, why the public may distrust the nuclear industry, or why the public may engage the political process. In lieu of this Dominating Concept that produces these undesirable consequences, four new Dominating Concepts that lead to more desirable consequences are proposed. (authors)

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

Evaluation of waste arising from future nuclear fuel cycle - 15197

A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein. The fundamental characteristics of nuclear fuel cycles were used to create a comprehensive set of fuel cycle options. These options were categorized into 40 Evaluation Groups based on similarities in their physics-based performance. These Evaluation Groups, listed below, comprehensively represent all fuel cycle options: - 8 once-through Evaluation Groups: 5 uranium-based fuel cycles and 3 thorium-based fuel cycles - 10 limited-recycle Evaluation Groups: 6 uranium-based fuel cycles and 4 thorium-based fuel cycles, and - 22 continuous-recycle Evaluation Groups: 14 uranium-based fuel cycles and 8 thorium-based fuel cycles. The study focused on the quantity and characteristics of radioactive wastes generated by the different fuel cycles, examining 5 metrics characterizing waste generation: - mass of spent nuclear fuel (SNF) plus high level waste (HLW) disposed per energy generated, - activity of SNF + HLW (at 100 years) per energy generated, - activity of SNF + HLW (at 100,000 years) per energy generated, - mass of depleted uranium (DU) + recovered uranium (RU) + recovered thorium (RTh) disposed per energy generated, and - volume of low level radioactive waste (LLW) per energy generated. (authors)

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

Submersible ROVs pioneer decommissioning of a legacy nuclear fuel storage pond - 15368

Submersible Remote Handling Vehicles (ROVs) are being used in innovative ways at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK in a legacy fuel storage pond in support of pond decommissioning. A wide range of sensors and tooling attached to submersible ROVs are being used to characterise the contents of the pond and perform a wide range of maintenance and material retrieval tasks. Visual surveys, sludge and liquor sampling and radiological surveys are just some of the characterisation projects undertaken with the help of the ROVs. The ROVs have also been deployed with tooling including manipulators which can retrieve debris from the pond floor. A variety of mini and work class ROVs are used in the ponds and these can access areas of the pond that the traditional technologies such as long reach tooling and manipulators from overhead cranes cannot reach safely, efficiently and cost effectively. The ROVs are controlled at a distance from the pond to minimise operator exposure to radiation and avoid potential radiological contamination. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2011

Engineering Glass Passivation Layers -Model Results

The immobilization of radioactive waste into glass waste forms is a baseline process of nuclear waste management not only in the United States, but worldwide. The rate of radionuclide release from these glasses is a critical measure of the quality of the waste form. Over long-term tests and using extrapolations of ancient analogues, it has been shown that well designed glasses exhibit a dissolution rate that quickly decreases to a slow residual rate for the lifetime of the glass. The mechanistic cause of this decreased corrosion rate is a subject of debate, with one of the major theories suggesting that the decrease is caused by the formation of corrosion products in such a manner as to present a diffusion barrier on the surface of the glass. Although there is much evidence of this type of mechanism, there has been no attempt to engineer the effect to maximize the passivating qualities of the corrosion products. This study represents the first attempt to engineer the creation of passivating phases on the surface of glasses. Our approach utilizes interactions between the dissolving glass and elements from the disposal environment to create impermeable capping layers. By drawing from other corrosion studies in areas where passivation layers have been successfully engineered to protect the bulk material, we present here a report on mineral phases that are likely have a morphological tendency to encrust the surface of the glass. Our modeling has focused on using the AFCI glass system in a carbonate, sulfate, and phosphate rich environment. We evaluate the minerals predicted to form to determine the likelihood of the formation of a protective layer on the surface of the glass. We have also modeled individual ions in solutions vs. pH and the addition of aluminum and silicon. These results allow us to understand the pH and ion concentration dependence of mineral formation. We have determined that iron minerals are likely to form a complete incrustation layer and we plan to look more closely at Vivianite [Fe3(PO4)2-8(H2O)] and Siderite [FeCO3] in the next stage of the project.

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

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated

Worrall, Andrew↗
DOE OSTI2019

First Road / River Barge Security Transport of SNF under the New German Transport Security Guidelines - 19102

A transfer of 342 spent fuel assemblies from NPP Obrigheim to NPP Neckarwestheim, both located alongside the river Neckar in the south of Germany, is required to promote the decommissioning process of NPP Obrigheim, which started in 2008 after the NPP had been shut down three years before. The spent fuel storage facility in Neckarwestheim offers sufficient storage space to accept the 342 SNF assemblies, stored in fifteen dry casks type CASTOR{sup R} 440/84 mvK. After several transport feasibility studies had been conducted, Daher Nuclear Technologies GmbH was awarded the commission to plan, prepare and conduct the SNF transport. Since the combined road / river barge transport promised to be the most suitable mode of transportation Daher Nuclear Technologies started to prepare its security concept, a prerequisite to apply for and attain the transport license mandatory for transporting spent nuclear fuel in Germany. In March 2014 the company applied for the transport license, which, after more than three years and numerous amendments to the security concept, was finally granted to Daher Nuclear Technologies on May 16, 2017 by the German Federal Office for the Safety of Nuclear Waste Management (successor organization of the BfS). Simultaneously to the license application phase the transport equipment - amongst others the heavy-duty trucks and trailers, the pusher tugs, the river barge and transport frames - had to be provided and modified to meet the intensified requirements of the New German Transport Security Guidelines Road / Rail as well as several other legal regulations referring to the transport of dangerous goods by road and inland waterways (e.g. ADR - European Agreement concerning the International Carriage of Dangerous Goods by Road and ADN - European Agreement concerning the International Carriage of Dangerous Goods by Inland Waterways). The planning process involved various meetings with representatives of different authorities and administrative bodies incl. accredited experts as well as subcontractors. Tremendous training efforts had to be undertaken to train more than 100 transport participants regarding security measures, health, safety and environmental measures, as well as the proper handling of radioactive material. This also included the training of the security staff, that had to be qualified in compliance with the new Transport Security Guidelines Road and Rail. Finally, in June 2017, when all preparations where complete, all requirements met, and all permits obtained, the first ever security SNF transport on a German river was performed. Three laden CASTOR{sup R} 440/84 mvK transport and storage casks were transported in one batch via heavy-duty truck-trailer from the NPP Obrigheim to the NPP's own harbor, to be rolled on to the river barge, that would carry the three casks upstream the river Neckar to the approx. fifty kilometer distant NPP Neckarwestheim. The journey on the river took about eleven hours, in which several bridges and locks had to be passed. The entire transport was escorted by qualified security staff and auxiliary police forces on waterways and alongside on roads. It attracted nationwide public attention, due to its pioneer status and the fact, that there hadn't been any SNF or HLW shipments in Germany since 2011. After the convoy safely arrived at NPP Neckarwestheim's pier the three laden CASTOR{sup R} 440/84 mvK casks were rolled off the river barge via heavy-duty truck-trailers to be stored in the on-site interim storage facility. This first transport was a success and was followed by four additionally successful security transports in September, October, November and December 2017. The public attention receded significantly during the transport campaign, which ended on December 19, 2017 with the transfer of the last three of fifteen casks to the on-site interim storage facility in Neckarwestheim. With the last cask safely transferred to the interim storage facility Daher Nuclear Technologies successfully completed the transport campaign on schedule, meeting one of EnBW's fundamental objectives to finish the transfer of the 342 SNF assemblies by the end of 2017. (authors)

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

Study of treatment as an alternative WM pathway: examples of I-graphite and metallic components such as steam generators - 15212

EDF is working continuously to optimize and secure its waste management pathways, leading in some case to develop appropriate treatment solutions that could be implemented parallel to reference solutions, complementary or as alternative solutions to already existing paths. This approach is illustrated by the examples of steam generators replaced during the operation of French PWR power plants, and of irradiated graphite (stacks and sleeves), coming from the decommissioning of UNGG reactors, which have been shut down. Concerning graphite wastes, the near surface disposal facility intended for this waste does not exist yet, and is currently giving rise to the research of a suitable site, and an evaluation of its performances. While accompanying ANDRA (French national agency for nuclear waste management), in charge of the development of the facility, and while waiting for more data to be available, EDF is continuing the consolidation of graphite wastes radiological inventory, and is working on the development of an alternative management solution, based on a thermal treatment process. Depending on the nature of the treatment, and on the choice of operating parameters, this treatment may contribute to reduce graphite wastes volume and/or promote its acceptance in a disposal facility. Reducing the radiological inventory by decontamination is one of the means to achieve these objectives. Concerning replaced steam generators, nearly 200 of them will have been removed by 2030, and stored on site waiting for their treatment, which represents about 65 000 tons of metal. Initially intended to be cut and packed on NPPs' sites during their dismantling, EDF is studying the economic value of having the steam generators treated in a centralized facility, which allows reducing the volume of final wastes by optimizing cutting operations, while considering recycling after melting. However, this studied pathway implies a number of difficulties to be overcome. In particular, the shipment of such large components to a centralized facility may be complex. Moreover, French regulatory framework does not allow recycling outside nuclear industry, therefore limiting recycling opportunities to storage containers for radioactive wastes. (authors)

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

Chemical durability and corrosion-induced microstructure evolution of compositionally complex titanate pyrochlore waste forms with uranium incorporation

Chemical durability of compositionally-complex rare-earth titanate pyrochlore solid solutions with or without uranium incorporation was investigated by semi-dynamic leaching testing. A Ti-enriched amorphous passivation film formed on the surface due to the surface alteration-reorganization mechanism with preferential release of the weakly oxygen-bonded rare earth elements. Elemental release was found to be strongly correlated with chemical disorder, with a negligible correlation with entropy. The release rate of the uranium gradually decreased with increasing numbers of components, implying a possibility of designing and optimizing waste form performance with enhanced chemical durability by controlling composition complexity and chemical disorder for effective nuclear waste management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Corrosion of Carbon Steel Inside Used Nuclear Fuel Containers - 19492

This paper presents an overview of the carbon steel (CS) corrosion research being conducted by the Radiation Assisted Materials Performance Science (RAMPS) group at Western University, in collaboration with the Canadian Nuclear Waste Management Organization (NWMO) and Canada Natural Science and Engineering Research Council (NSERC). The objective of our studies is to develop a high-fidelity CS corrosion dynamic model that can be used to assess the integrity of the current Canadian used fuel container (UFC) design with confidence. Our key findings, the implications of the findings on corrosion science in general, and the potential consequences of these results on the predictions for long-term UFC integrity are described. Results are presented on the following five themes: chemical kinetics of γ-radiolysis of liquid water and humid air, aqueous corrosion of CS under γ-radiation, corrosion of CS under small water volumes and water filled gaps, Liesegang ring formation in a gel medium, and radiation-induced iron oxide nanoparticle formation from dissolved metal ions. These studies together represent a detailed and integrated examination of the whole CS corrosion system, and in particular, its behaviour under γ-radiation. The key finding of this research is that under certain conditions, the CS corrosion system can exhibit non-linear dynamic behaviours - which can be seen in periodic precipitation patterns and oscillations in measurable system properties such as corrosion potential. This throws into question existing corrosion models that do not account for these types of behaviours. We describe the preliminary steps we have taken in integrating these new understandings into a CS corrosion dynamic model that will be able to make long-term predictions of CS corrosion in the UFC environment. (authors)

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

Integrated Systems Analysis for a Regional Deep Borehole Repository for Commercial Spent Nuclear Fuel - 16291

Some nuclear waste stakeholders have suggested private interim storage facilities (ISF), but little work has been done to consider the cost or logistics for nuclear waste management or disposal facilities smaller than the national scale. This paper performs an integrated system analysis for a deep borehole repository system which would accept waste from a regional subset of the national commercial reactor fleet. The regional subset is comprised of a centrally-located host state with a large SNF inventory surrounded by states with small SNF inventories. The chosen region will contain between one-fifth and one-quarter of the national commercial SNF inventory. Using best estimates of past and future SNF discharges in the sample region, the SNF inventory in the chosen region is characterized by reactor of origin. Basic parameters are then chosen for the regional repository, such as the annual rate of waste emplacement and the time necessary to process borehole-specific waste canisters. Scenarios chosen for analysis test the sensitivity of a variety of system parameters, including the annual rate of waste acceptance, the division of annual waste acceptance among sites, the use of bare fuel transportation canisters, and the size of the regional subset of reactors. Estimated costs for the facility are $13 billion for around 27,000 MT of SNF and $16 billion for around 35,000 MT. This translates to $490,000 per MT and $466,000 per MT, respectively. These unit costs are comparable to the $370,000 per MT estimate for the full 140,000 MT national inventory. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Remote Radiation Sensing Using Aerial and Ground Platforms

Remote sensing of ionizing radiation has a significant role in waste management, nuclear material management and nonproliferation, and radiation safety. Robotic platforms can surpass the number of tasks that are achieved by humans. With this technique, the operator's radiation exposure can be decreased. Remote sensing allows for the evaluation and monitoring of radiological contamination. Gamma-ray and neutron sensors were integrated onto the robotic platforms. This approach allows for the radiation sensor data to be dynamically tracked and mapped thus enabling further analysis of the radiation flux in temporal and spatial domains. The goal is to complete scheduled tasks while the robot is being irradiated. To achieve this, electronic components must be shielded and radiation hardened. CZT Detector: Cadmium Zinc Telluride (CZT) detector technology has been a promising solution for gamma-ray and x-ray measurements. Detector data is transferred to the Odroid minicomputer that controls and powers the module via the USB. Robot Operating System (ROS) was utilized for data acquisition and data fusion. The Mariscotti method was employed for the spectrum analysis. A function was programmed in ROS for the automatic identification of photopeaks. CLYC Detector: A Cs{sub 2}LiYCl{sub 6}:Ce{sup 3+} (CLYC) detector was used for simultaneous medium-resolution gamma-ray measurements and neutron counting. A 2.54 cm diameter photomultiplier tube (PMT) was equipped with a high voltage supply and a miniature digitizer. Gamma-ray excitation: fast core-to-valence luminescence (CVL) with 1 ns decay constant, and prompt Ce{sup 3+} emission with 50 ns decay constant. Neutron excitation: slow cerium self-trapped excitation (Ce{sup 3+} STE), 1000 ns decay constant. Radiation Source Localization: Maximum Likelihood Estimation (MLE) and gradient-based methods were used to locate the position of a radiation source based on measured radiation intensities. Multi-Particle Transport Code FLUKA: Estimation of radiation damage of the electronic components is important in order to optimize the robot's operational time while it is irradiated. Displacement per atom (DPA) represents the radiation damage in materials exposed to the ionizing radiation. Various shielding layers of different thickness t were analyzed (< 5% statistical error). The model of the controller of the UAS was designed in FLUKA. Conclusion: CZT and CLYC detectors were integrated onto the robotic platforms. Radiation source localization and contour mapping using robotic platforms were studied. Functions for data analysis and fusion were developed in ROS. FLUKA code was utilized to analyze DPA values. Layers of low-density and high-density materials were used to shield the UAS electronics.

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

Technetium Management for Hanford Tank Waste Processing and Disposition - 16399

Technetium-99 (Tc) generated from the fission of {sup 235}U and {sup 239}Pu in high yields is one of the most difficult contaminants to be addressed at the U.S. Department of Energy (DOE) Hanford Site. Among radioactive constituents in the tank waste, Tc presents a unique challenge in that it is radio-toxic (β = 292 keV), has a long half-life (T{sub 1/2} = 2.11x10{sup 5} y), and exists predominantly in the liquid fraction of the alkaline tank waste, generally in the anionic form of pertechnetate TcO{sub 4}{sup -}, which is highly volatile at low-activity waste vitrification melter temperatures and mobile in the subsurface environment. This makes immobilization of Tc into high-durability waste forms a critical technical challenge for nuclear waste management. The complex behavior of Tc under storage, treatment, and immobilization conditions significantly affects its management options. The overall objective of the Technetium Management program is to provide the DOE Office of Environmental Management with practical solutions to Tc needs encountered in the processing of Hanford tank waste. To reduce the technical uncertainty associated with the long-term environmental impact of Tc, a systems-based approach is needed to address Tc management needs by evaluating and developing science and engineering options for treatment, immobilization, and disposition of Tc contained in Hanford tank wastes. This paper describes major elements of Technetium Management program, including: - Identifying the chemical forms of Tc in tank waste through analytical speciation of low-valent Tc(I) species in alkaline solutions. - Performing proof-of-concept demonstration for detection of low-valent Tc(I) by fluorescence spectroscopy. - Identifying alternative non-glass Tc immobilization methods. (authors)

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

Crystallization of Sodium Phosphate Dodecahydrate and Re-crystallization to Natrophosphate in Simulated Hanford Nuclear Waste - 16188

The nuclear waste at the Hanford site near Richland, WA, has large concentrations of phosphate in the form of the phosphate ion, sodium phosphate dodecahydrate (Na{sub 3}PO{sub 4}.12H{sub 2}O.(1/4)NaOH) and natrophosphate (Na{sub 7}F(PO{sub 4}){sub 3}.19H{sub 2}O). Sodium phosphate dodecahydrate can form a gel and natrophosphate can crystallize large particles, complicating the processing of slurries of both salts. The gel is regarded as more problematic, so natrophosphate has historically been crystallized to prevent phosphate gelling. This study determined if natrophosphate crystals can grow to large size in short time periods (a few weeks), time periods relevant to short process shutdowns. Solutions of NaOH and NaAl(OH){sub 4} were blended at different ratios with stock solutions containing NaOH, NaF, and Na{sub 3}PO{sub 4} at 50 deg. C. The mixtures were allowed to cool to 22 deg. C, and the crystal growth was monitored by Polarized Light Microscopy (PLM) for 17 days. Four of the six blends investigated gelled rapidly due to Na{sub 3}PO{sub 4}.12H{sub 2}O.(1/4)NaOH precipitation. The gel slowly dissipated over time as the solids recrystallized into natrophosphate. In one case, the natrophosphate reached sizes of greater than 1000 microns in diameter in just 4 days. This rapid gelling and crystallization kinetics is important to engineers trying to manage nuclear wastes high in phosphates. Hanford tank farm engineers are developing sampling plans to support temperature and process control strategies for preventing the formation of solid phosphates. They are also developing methods of suspending large natrophosphate crystals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Panel Session 4B: Best Practices in Project Communication Involving Stakeholders

This panel focused on a sustainable, holistic and integrated approach to structure nuclear waste management, disposal and remediation decisions. This panel had participation from internal and external stakeholders and discuss various approaches, tools, and techniques to effectively communicate factors that inform decisions. The panelists presented a summary of their experiences and share key points that were helpful in gaining common understanding with stakeholders. Presentations proposed include: 1 - The Evolution of U.S. Department of Energy and Stakeholder Relationships at the Fernald Preserve, Ohio, Site (Penny Borgman, Navarro Research and Engineering, Inc. (USA)); 2 - The Site Specific Advisory Board Experience - A Members Perspective - 19069 (Belinda Price, Oak Ridge Site Specific Advisory Board (USA)); 3 - Preserving the Mound Laboratory Legacy-The Mound Cold War Discovery Center - 19550 (Susan Smiley, US DOE Office of Legacy Management (USA); Melissa Lutz, Navarro (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗