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DOE OSTI2015

Bio-sorption of thorium by bone meal - 15435

In the recent years, there is an increasing interest in the application of biological materials for the removal of heavy metal ions from aqueous solutions. There are a variety of biological materials that have been proven efficient bio-sorbents such as agriculture residues, plants, algae, etc. The use of these materials is attractive and advantageous because they are inexpensive in comparison with commercial adsorbents. Bone meal is a biological material that reunites low cost and efficiency. It is a natural material that contains large amount of calcium phosphate in the form of hydroxyapatite. Studies with hydroxyapatite showed that this material is efficient in removing heavy metals from polluted soils due to their ability to adsorb molecules and can be used to remove radionuclides from liquid aqueous solutions, as liquid radioactive wastes generated in many activities. Bone meal was purchased in local market and chopped and sieved to obtain particle size between 0.297 mm and 0.125 mm. Batch bio-sorption experiments were performed to determine the metal uptake capacity and equilibrium time. Fixed concentrations of thorium solutions were prepared by dissolving thorium nitrate in distillate water. The bone meal was suspended in 5 mL of thorium solutions in a ratio of 0.2% w/v and left in contact during different times. The contact times adopted were 30 min, 1, 2 and 4 hours. After the contact time, the bone meal was removed by filtration and the supernatant, analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES). The results were evaluated using isotherms and kinetics models. The maximum capacity for the bone meal was 11.5 mg/g in 2 h. These results suggest that bio-sorption with bone meal can be applied in the treatment of aqueous solutions and liquid radioactive wastes containing thorium. (authors)

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

Diversity and the citizens' advisory board: aligning disparate views for better results - 15469

The Northern New Mexico Citizens' Advisory Board (NNMCAB) is a group of -citizen experts,- i.e., experts on being citizens who take the time and focus their attention to grow more informed about nuclear science, nuclear waste remediation, and environmental monitoring issues. The Board digests complex information associated with these topics in light of the Los Alamos National Laboratory's (LANL) and Department of Energy's (DOE) activities in Northern New Mexico and discusses and conveys this information clearly to a diverse regional stakeholder population that includes a 400+ year old Hispanic culture, a 1,000+ year old Native American pueblo culture, a highly educated scientific and technical community of federal workers and their families, rural agricultural communities, creative/artistic communities, and political and peace activists. Our Board has either representation from each of these diverse New Mexican sectors or enjoys public comment from these sectors. Projects that involve such stakeholder-expert exchanges include remediation of an underground chromium plume, continued investigation and remediation associated with a radiation leak at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, the transport of Transuranic Waste off the Los Alamos 'Hill,- the ultimate disposition of -33 Shafts,- currently in place at Los Alamos National Laboratory, and the distilled voice of the citizen for continued and sufficient federal funding each year for responsible clean-up and monitoring of nuclear activities. Through the NNMCAB's communication venue of regular public meetings, mis-characterization is prevented, public input is incorporated into the official record, and recommendations to the DOE are promulgated. Scientific and technical experts join the bi-monthly NNMCAB meetings and monthly committee meetings to present and discuss current topics. The availability of these experts and their deeper reference materials is key to the interchange, as is the respectful and patient dialogue that exists between board members and the subject matter experts, both in the public forums and between meetings. Meetings are professionally organized and orchestrated to maximize the value of the experts' and board members' participation time. The professional format of the meetings ensures maximum learning and relationship-building which fosters effective channels for public dialogue. (authors)

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

Degradation of Simulated HLW Glass and International Simple Glass immersed partially in water - 16474

Water degradation of glass waste forms destined for geological disposal has been studied in a variety of conditions including of bulk glass immersed completely in static or dynamic water. During the course of repository life, the canisters may be only partially immersed in the initial stages of contact with water. Under these conditions glass degradation occurs on the glass immersed in water and on the glass above water by vapor condensation. In this work degradation of International Simple Glass (ISG) and Simulated High Level Waste Glass (HSG) is examined by immersing partially in distilled water during the course of test. Under static experimental conditions at 90 deg. C, the condensation of vapor on the ISG glass surface formed Ca-rich precipitates and pits after degrading for 1 and 3 weeks respectively. Whereas HSG precipitated Ca-rich phases predominantly after 1 week and no pits were found on the surface. When compared, HSG was found to change surface chemistry faster than ISG. (authors)

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

A Literature Review and Compilation of Nuclear Waste Management System Attributes for Use in Multi-Objective System Evaluations - 16618

The purpose of this work was to compile a comprehensive initial set of potential nuclear waste management system attributes. This initial set of attributes is intended to serve as a starting point for additional consideration by system analysts and planners to facilitate the development of a waste management system multi-objective evaluation framework based on the principles and methodology of multi-attribute utility analysis. The compilation is primarily based on a review of reports issued by the Canadian Nuclear Waste Management Organization (NWMO) and the Blue Ribbon Commission on America's Nuclear Future (BRC), but also an extensive review of the available literature for similar and past efforts as well. Numerous system attributes found in different sources were combined into a single objectives-oriented hierarchical structure. This study provides a discussion of the data sources and the descriptions of the hierarchical structure. A particular focus of this study was on collecting and compiling inputs from past studies that involved the participation of various external stakeholders. However, while the important role of stakeholder input in a country's waste management decision process is recognized in the referenced sources, there are only a limited number of in-depth studies of the stakeholders' differing perspectives. Compiling a comprehensive hierarchical listing of attributes is a complex task since stakeholders have multiple and often conflicting interests. The BRC worked for two years (January 2010 to January 2012) to 'ensure it has heard from as many points of view as possible'. The Canadian NWMO study took four years and ample resources, involving national and regional stakeholders' dialogs, internet-based dialogs, information and discussion sessions, open houses, workshops, round-tables, public attitude research, web site, and topic reports. The current compilation effort benefited from the distillation of these many varied inputs conducted by the previous studies. The initial set of attributes is intended to provide a starting point for considering potential objectives for evaluating nuclear waste management system options. It is envisioned that this initial set will evolve and be revised as new and additional input is received from stakeholders on various attributes and their relative importance. (authors)

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

International Governmental Collaboration on Best Practice in Aging Infrastructure Management - 18027

Legacy nuclear sites in the US, UK and Canada are all facing a similar challenge of completing their operational and cleanup missions in parallel in a safe, efficient and cost-effective manner. Compounding the challenge is the fact that many important facilities are old with a significant level of deferred maintenance/capital needs, resulting in the need for a strategic approach to infrastructure management to ensure that maintenance and capital funding is directed at the highest priority items and that neither operations nor ongoing Decontamination and Decommissioning (D and D) activities are compromised. To help address this challenge, a Technical Information Exchange meeting was held in Ottawa, Canada in early June 2017 which brought together responsible parties from; - Government agencies including the UK Nuclear Decommissioning Authority (NDA), US Department of Energy Environmental Management (DOE EM), US National Nuclear Security Administration (NNSA), Atomic Energy of Canada Ltd (AECL) and Canadian Department of Defense - Implementers and other Government entities including Savannah River National Laboratory, Idaho National Laboratory and Canadian Nuclear Laboratories (CNL) - Associations including representation on behalf of the Institute of Asset Management and; - Contractors including Mission Support Alliance, Mott McDonald, and Longenecker and Associates The objective of this meeting was to share information about respective 'best practices' and to begin to develop common approaches to the prioritization and management of aging infrastructure investment across the organizations. The exchange focused on identifying common critical issues which were discussed through a series of presentations and small group discussions. The outcome from the meeting was an improved understanding by all participants of the latest approaches being applied to Infrastructure Asset Management and the formation of four International Working Groups which now meet on a regular (4-6 week) basis to share lessons learned and to develop international best practice in their area of interest. This paper details; - the issues and challenges that were identified during the meeting - how these were assimilated and distilled into discrete areas to enable formation of specific Working Groups - progress made by the Working Groups to date. (authors)

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

Effects of liquid subcooling and initial rod temperature on the minimum film boiling temperature

This paper presents an experimental study on the minimum film boiling temperature (T{sub min}) and its sensitivity to liquid subcooling and initial rod temperature. T{sub min} is measured through a quenching experiment using a vertical stainless steel rod that is quenched in a distilled water bath. The cylindrical stainless steel rod has a flat circular end with diameter of 9.5 mm similar to that of a fuel rod and length of 25 cm. Physically, T{sub min} is used to define the boundary between film boiling and transition boiling, beyond which temperature liquid would not be able to come into physical contact with the solid surface. Results show that the quenching curves are affected by varying the liquid subcooling but are not significantly altered by the initial rod temperature. Additionally, the experiments include visualization of the boiling regimes. The visualization gives insight into the film flow behavior that is found to depend strongly on the liquid subcooling. It also provides an understanding of the phenomenon for quench front axial propagation (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Unique Facilities, Innovative Treatment Processes, and a Talented Workforce Allow Fluor Idaho to Successfully Disposition Transuranic Waste - 18388

Fluor Idaho is utilizing state-of-the-art equipment, treatment technologies, and employee expertise at the Department of Energy's (DOE) Idaho Site to treat contact-handled transuranic (CH-TRU) waste, as well as extremely challenging remote-handled transuranic (RH-TRU) waste forms, to prepare the material for shipment and permanent disposal at the Department's Waste Isolation Pilot Plant (WIPP). Nowhere else in the DOE Complex - perhaps the world - will one find the compilation of tools and an experienced workforce with an unparalleled history of safely retrieving, characterizing, treating, repackaging, certifying, and shipping TRU waste. Since 2003, the Advanced Mixed Waste Treatment Project (AMWTP) has used both conventional and unique retrieval concepts, high-tech characterization equipment - including real-time radiography and radioassay - robotics, and automated treatment processes to retrieve and treat 65,000 cubic meters of above-ground CH-TRU waste. Additionally, engineered controls and specialized personal protective equipment have allowed employees to make thousands of entries into high-radiation areas without a skin or clothing contamination incident - evidence of the employees' commitment to safety. AMWTP has encountered and found solutions for many challenging waste forms, including polychlorinated biphenyl-contaminated wastes, TRU waste containers with up to 1,000 fissile gram equivalent plutonium-239, oversized TRU boxes, and partially absorbed organic sludge waste. By utilizing additional facilities and equipment at the Accelerated Retrieval Project V and VII, Fluor Idaho has developed new processing lines and capabilities to ensure worker safety and milestone completion. These two soft-sided buildings, located at a facility adjacent to AMWTP, were originally constructed to exhume Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) TRU wastes from a section of a 97-acre landfill. After a total of 1,150 cubic meters of targeted waste were removed, personnel, working in collaboration with state regulators, successfully implemented a unique regulatory strategy that allowed these CERCLA facilities to be modified and permitted to treat Resource Conservation and Recovery Act (RCRA) wastes from the AMWTP - a savings of tens of millions of dollars over the cost of constructing new buildings. Adding to the AMWTP's technological uniqueness is its ability to treat wastes from other facilities in the DOE Complex. Two prior decision documents and an agreement with the state of Idaho have allowed AMWTP to treat wastes from 15 other DOE sites. It is the only project in the complex to have done so. At another project located next to AMWTP, targeted waste exhumation within the Subsurface Disposal Area is approximately two years ahead of schedule. Crews have removed more than 8,600 cubic meters of CH-TRU and hazardous wastes from eight exhumation areas that equal a combined footprint of 4.75 acres. Less than one acre remains to be exhumed to satisfy the 2008 Record of Decision waste exhumation end state between the DOE, Environmental Protection Agency, and state of Idaho. Construction of the final building where targeted waste exhumation will continue was completed during the summer of 2017. Exhumation will be completed within the next few years. Following exhumation completion, an evapotranspiration barrier (or cap) will be constructed over the entire landfill. Fluor Idaho continues to make progress on a more challenging project involving an inventory of RH-TRU waste that also contains the reactive metal sodium. Crews are utilizing hot cells, once used for spent nuclear fuel reprocessing and high-level waste treatment at the Idaho Nuclear Technology and Engineering Center, to render this challenging waste form safe for disposal in a national geologic waste repository. The treatment methods being employed on this RH-TRU waste are unique in the DOE Complex. Minute concentrations of water are sprayed onto the sodium-contaminated material in a process called 'spritzing' to chemically change the sodium from a reactive metal to a non-reactive salt. In other cases, sodium-contaminated waste is dipped in water to initiate a reaction in a process called 'immersion.' Lastly, for more challenging waste shapes that could not be successfully treated with the two prior methods, crews lower the sodium-contaminated RH-TRU waste into a one-of-a-kind distillation unit that uses high temperature and vacuum to volatilize the sodium and recapture it in a condenser unit for disposal. The RH-TRU waste treated with this technology is then safe for permanent disposal. Fluor Idaho is proud to operate DOE's CH- and RH-TRU waste treatment facilities, as well as use innovative processes and its talented workforce to complete this important work-scope. Such capabilities allow the contractor to meet the commitments made by the DOE to the state of Idaho. (authors)

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

Capacity of ETS-10 supported carbon nanosorbent for removal of multicomponent from off-gas stream

Nuclear fission results in fission and activation products, some of which are volatile under the conditions of used nuclear fuel (UNF) reprocessing. Several volatile radionuclides are released into the atmosphere through offgas stream, which includes {sup 3}H, {sup 14}C, {sup 129}I and noble gases, Kr and Xe. {sup 129}I is a highly mobile and volatile contaminant. It has half-life period about 1.57 x 107 years that tends to bioaccumulate, concentrating in the thyroid glands and affecting bio-metabolism in humans. {sup 85}Kr is a chemically inert radioisotope with a half-life of 10.7 years that continuously accumulate in the environment. Control of emissions of these contaminants is necessary to comply with the stringent regulatory limits placed on their releases to protect human health and environment. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbent shows the most promise results. Up to now, several sorbents such as silver impregnated mordenite, silver ETS-10, chalcogen based aerogels, and metal organic framework, MOFs have been developed for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. Present research work is focused on the determination of the removal capacity of C'at'ETS-10 sorbent for a multicomponent (iodine and krypton) off-gas stream at ambient temperature. Our previously reported work showed that the 10 wt% C'at'ETS-10 sorbent was active for adsorption of krypton as well as iodine at ambient temperature and pressure. Therefore, 10 wt% C'at'ETS-10 sorbent was synthesized by wet-impregnation technique. The performance evaluation of the prepared nanosorbent was examined for the removal of iodine and krypton from the breakthrough curves obtained using bench scale experimental setup. ETS-10 supported carbon nanoparticles, C'at'ETS-10 nanosorbent was successfully synthesized by wet-impregnation technique. The 1/8 inch pelletized sorbent was used to minimize the pressure drop in adsorption column. The capacity of sorbent was found to be 0.143 mg g{sup -1} for krypton and 24.5 mg g{sup -1} for iodine. This experimental results indicates the C'at'ETS-10 is the effect sorbent for removal of multicomponent, krypton as well as iodine from the off-gas stream at ambient temperature. (authors)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

Estimation of Deposition Behavior of Supersaturated Silicic Acid around Geological Disposal Facility - 19137

For safety evaluation about the geological disposal system, the leakage of radionuclides with groundwater flow is estimated by considering the groundwater flow through the manmade and natural barriers. In such an estimate, we should also consider the influence of cement materials used for the construction of the disposal facility system. Cement material including high alkalinity alters groundwater condition up to 13 in pH and dissolves numerous silicate materials, which release silicic acid to groundwater. Since this groundwater mixes with natural groundwater downstream, supersaturated silicic acid will be produced with the decrease of pH to 8 around the repository system. Thus, in this study, the behavior of supersaturated silicic acid in a flow system was examined by using the column packed with amorphous silica particles under the condition of constant hydraulic head. The packed length is 20 mm or 40 mm, and the supersaturated silicic acid solution was continuously injected into the column. Here, the supersaturated silicic acid was adjusted by mixing with distilled water of pH 8 just before the solution was injected into the column. In the results, the supersaturated silicic acid quickly deposited in the packed column, and colloidal silicic acid was not detected at the outlet of the column. The deposition amount of supersaturated silicic acid in the column decreased with time, and became larger in the longer column. Besides, the flow rate decreased with the deposition of supersaturated silicic acid. Thus, this study monitored the decrease rate of permeability with the supply amount of silicic acid. In the results, the change rate did not depend on packed length. This suggested that even with a short packed length, by considering the amount of precipitation of supersaturated silicic acid, it is possible to estimate the change in permeability similarly to the case of a long packed length. Such a decrease effect in permeability may contribute to retarding the migration of radionuclides around the repository. (authors)

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

Rehabilitation of Westbay{sup R} Wells with PCE-Contaminated Tubing Fluid - 19250

In November 2015, tetrachloroethylene, also known as PCE, was discovered in groundwater samples collected from a Westbay{sup R} multi-level groundwater monitoring well at the Idaho National Laboratory (INL).a The result was unexpected due to the remote location of the well. Subsequent investigation determined that PCE was not present in the groundwater. The detected PCE was due to inadvertent contamination of the tubing fluid in the Westbay well-monitoring system. Further investigation showed that three of the 11 Westbay wells at the INL contained PCE-contaminated tubing fluid at concentrations above the U.S. Environmental Protection Agency maximum contamination level (5 μg/L); the three wells were identified for rehabilitation as part of the INL well maintenance program. Well rehabilitation would either remove and replace the PCE-contaminated tubing fluid or remove the PCE from the tubing fluid. The interior of the Westbay well casing is isolated from the surrounding groundwater and filled with tubing fluid (either distilled or potable water). Groundwater samples are collected by opening sealed ports using a specialized sampling device. The unique construction prevents the direct interaction of the groundwater and tubing fluid. The tubing fluid also supports the structural integrity of the Westbay wells. The unique construction of the Westbay wells-i.e., the small diameter (5.08 cm [2 in.]) and the depths of the tubing fluid ranging from 144.8 to 419 m (475 to 1,376 ft) below ground surface-posed unique challenges to the rehabilitation. Because the tubing fluid provides structural integrity to the Westbay well, the fluid cannot be removed and then replaced without damaging the integrity of the Westbay well. Additionally, the small diameter of the systems and the associated depths limit the availability of equipment that can be used for the rehabilitation. After an evaluation of potential methods, it was determined that the use of an in-well, granular-activated-carbon filtration swab deployed via a wire line would provide the safest and most cost-effective method for rehabilitation. Rehabilitation commenced during the summer of 2018 and is ongoing with measurable success. (authors)

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

Development of Processing Technology for Radioactively Contaminated Water at Chernobyl Nuclear Power Plant - 19656

During the operation of the Chernobyl Nuclear Power Plant, distillation by evaporation was the main method for processing radioactively contaminated water. When the plant was decommissioned, the evaporation facility was mothballed and floor drain treatment was suspended in 2016. At present, the existing tanks with the capacity of 26,000 m{sup 3} are more than 60% full with radioactively contaminated floor drains, which requires alternative processing methods. The current process flow for radioactive waste management at State Specialized Enterprise 'Chernobyl NPP' includes processing of accumulated liquid radioactive wastes at a liquid radioactive wastes processing plant. Under the processing technology, liquid radioactive waste is immobilized at the plant by cementation and the resulting solid radioactive waste is then buried in a near-surface storage. Key factors for the selection of alternative radioactively contaminated water processing methods are the possibility to clear the processed water (no control monitoring), which requires the same degree of processing as for drinking water, and minimization of radioactive waste generation due to the lack of free space for solid waste disposal in the near-surface storage. Floor drains that require processing contain radionuclides such as {sup 137}Cs, {sup 90}Sr, {sup 60}Co and {sup 40}K. The current level of the total beta and gamma activity is about 4*105 Bq/kg. The following criteria are set for the Chernobyl NPP site for water to be disposed to the sewerage system: - Max. 37 Bq/kg for the total beta and gamma containing radionuclide activity; - Max. 0.7 Bq/kg for the total alpha containing radionuclide activity. METOIL produces inorganic sorbents for removing radionuclides from liquids and develops sorbent-based technology for a variety of applications related to liquid radioactive waste processing. In 2017-2018, experts from the Chernobyl NPP and METOIL ran several joint laboratory tests to evaluate the efficiency of inorganic sorbents made by METOIL for processing floor drains at the Chernobyl NPP. The test procedure included treatment of radioactively contaminated water in a two-stage sorption process: static sorption (sorbents are mixed in a liquid) and dynamic sorption (the liquid is filtered through a sorbent column). NIKET and MODIKS sorbents were provided for the tests. The results of the test measurements showed that the total alpha containing radionuclide activity is less than 0.026 Bq/kg and the total beta and gamma containing radionuclide activity is 9.2 Bq/kg. The physical and chemical parameters after processing also met the water disposal criteria. High sorption capacity of the sorbents used for the tests allows processing hundreds of water volumes with just one sorbent volume. At present, preparations are underway for full-scale tests at a pilot plant with a capacity of minimum 10 m{sup 3}/day. At a later stage, spent sorbents may be immobilized by inclusion into the cement matrix, which will substantially reduce the amount of radioactive waste to be buried. (authors)

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

Mercury Speciation of Hanford 241-AP-107 Tank Waste Samples - 20384

The concentration of key mercury species in the Hanford tank waste has not been previously measured. Based on process knowledge, an estimated inventory of 2,100 kg of mercury is assumed and was distributed in varied amounts across 149 single-shell tanks (SSTs) and 28 double-shell tanks (DSTs). Limited analyses of individual SSTs and DSTs for total mercury were all below detection, yielding upper limits in concentration estimates based solely on instrument detection limits, resulting in an uncertain mercury inventory for Hanford tanks. Accurate and reliable chemical speciation and quantification is desired to anticipate, predict, and abate the mass movements of various mercury species through the Hanford waste processing flowsheet. Highly sensitive separation and quantification methods for total, elemental, and monomethyl mercury species, in tank waste, have been developed at the Pacific Northwest National Laboratory's (PNNL) Radiochemical Processing Laboratory (RPL). A non-radiological environmental method to quantify mercury species has been adapted for application to Hanford tank 241-AP-107 raw supernatant feed, filtered feed, and cesium-decontaminated samples. These samples represent feed inventory and pretreatment process effluents prior to immobilization by vitrification. This method separates and preconcentrates total, elemental, and monomethyl mercury onto selective solid substrates with little-to-no radioactive background. The solid samples can then be released from strict radiological control, allowing for analytical work to be conducted in low-level radiological facilities. Separation of mercury species was demonstrated at PNNL's RPL in Richland, Washington before transport and analytical quantification using cold vapor atomic fluorescence spectrometry (CVAFS) at PNNL's Marine Science Laboratory (MSL) in Sequim, Washington. Additional studies were conducted to assess the stability of monomethyl mercury in distilled matrices and preconcentrated onto the Carbotrap{sup R} solid substrate to assess sample degradation over time during transport. (authors)

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

Recovery of Cs-137 from Electric Arc Furnace Dust - 20562

The present work proposes the Cs-137 recovery from Electric Arc Furnace Dusts (EAFD) by leaching, in order to reduce the volume of radioactive waste to be managed. Leaching tests were performed on different samples of EAFD contaminated with Cs-137. Two leaching methodologies were tested: dynamic and static using as medium distilled water and in some cases NaOH. The results indicate the possibility of recovering Cs-137 from EAFD using the dynamic method; however, the recovery efficiencies of Cs-137 depend on EAFD's chemical composition. The leachates of Cs-137 were quantitatively adsorbed on exchange resins, showing its great potential to isolate or contain the Cs-137 recovery from EAFD. This study shows the feasible to apply the dynamic leaching methodology to recover Cs-137 from EAFD and the use of ion exchange resins for the isolation of removed Cs-137. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A review of human-system interface design issues observed during analog-to-digital and digital-to-digital migrations in U.S. nuclear power plants - 312

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is developing a scientific basis through targeted research and development (R and D) to support the U.S. nuclear power plant (NPP) fleet in extending their existing licensing period and ensuring their long-term reliability, productivity, safety, and security. Over the last several years, human factors engineering (HFE) professionals at the Idaho National Laboratory (INL) have supported the LWRS Advanced Instrumentation, Information, and Control (II and C) System Technologies pathway across several U.S. commercial NPPs in analog-to-digital migrations (i.e., turbine control systems) and digital-to-digital migrations (i.e., Safety Parameter Display System). These efforts have included in-depth human factors evaluation of proposed human-system interface (HSI) design concepts against established U.S. Nuclear Regulatory Commission (NRC) design guidelines from NUREG-0700, Rev 2 to inform subsequent HSI design prior to transitioning into Verification and Validation. This paper discusses some of the overarching design issues observed from these past HFE evaluations. In addition, this work presents some observed challenges such as common tradeoffs utilities are likely to face when introducing new HSI technologies into NPP hybrid control rooms. The primary purpose of this work is to distill these observed design issues into general HSI design guidance that industry can use in early stages of HSI design. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Multicomponent adsorption of radioactive iodine and krypton using ETS-10 supported carbon nano-polyhedrons

Over 13% of the world's energy demand is supplied by nuclear power plants in 31 countries. This demand is growing; by 2030 the IAEA predicts a 100% increase in the demand for nuclear energy. The increase in nuclear energy production imposes a growing strain on the uranium fuel supply. At current consumption rates the world's supply of uranium is projected to last ∼100 years. Following the IAEA's prediction for increased nuclear energy, uranium supplies will dwindle even more quickly. However, this consequence can be limited as ∼95% of uranium is left in used nuclear fuel (UNF). Reprocessing of UNF will utilize this remaining uranium and extend the lifetime of the world's uranium supply. One of the concerns with nuclear reprocessing is that it frees radionuclides created by the fission of uranium to be released into the atmosphere through off-gas. The major constituents of off-gas include: {sup 85}Kr, {sup 129}I, {sup 14}C, {sup 3}H, and various Xe isotopes. {sup 85}Kr and {sup 129}I present the largest threat due to their harmful environmental effects and abundance. They are the most practical contaminants to be removed to meet requirements for radionuclide release. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbents shows the most promising results. Until now, several sorbents such as silver impregnated mordenite, silver Engelhard titanosilicate 10, chalcogen based aerogels, and metal organic framework, MOFs have been investigated for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. This work is focused on the multicomponent of {sup 85}Kr and {sup 129}I from a simulated off-gas with an Engelhard titanosilicate 10 (ETS-10) supported carbon nano-polyhedron sorbent. Adsorption studies were performed to determine iodine sorption capacity as a function of temperature and iodine inlet concentration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uncertainty Quantification by Monte Carlo Analysis using CFD Simulations for GEMIX Benchmark Activities

Uncertainty quantification (UQ) methods can be applied to computational fluid dynamics (CFD) simulations for determining the uncertainty bands of the simulations results. In order to determine the applicability of these different methods, simple benchmark activities have been conducted by various group for comparison. The GEMIX (Generic Mixing experiment) test facility was used to develop benchmark data for this purpose by the Paul Scherrer Institute (PSI). The facility involves the mixing of two fluids that are initially separated. One of the two fluids is water with sugar dissolved in it to create a fluid with similar fluid properties to water, except for density. The other fluid was distilled water to reduce the uncertainty in the fluid's properties. During the experiments, velocity fields and concentration measurements were acquired at several planes of interest. The velocity fields were acquired using particle image velocimetry (PIV) and concentrations of each fluid were acquired using laser induced fluorescence (LIF). The inlet velocity profiles were acquired using laser Doppler velocimetry (LDV). The activity provided the opportunity to vary turbulence models, meshing methodologies, and numerical schemes for creation of uncertainty estimates. The present study investigates the usage of a simple Monte Carlo (MC) technique for varying the inlet boundary conditions of each inlet. The velocity profiles and densities of each inlet were specifically varied based on experiment measurement uncertainty. The intent was to isolate the uncertainty due to inlet quantities and not related to meshing, numerical scheme, or turbulence models. The comparisons of experimental and simulation results were found to be reasonably accurate. Though deficiencies were identified in the modeling approach to be corrected in future works by the authors. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Chabazite Zeolite Membranes for Krypton/Xenon Separation: Control of Membrane Properties and Process Modeling

Nuclear energy is considered to be a clean and economical means of carbon-free energy. As part of the overall process for recycling used nuclear fuel (UNF), it is necessary to develop technologies to separate the off-gases formed during the fission of uranium and plutonium in nuclear power reactors. The off-gas stream is generated during the shearing and chopping, voloxidation, and acid dissolution steps, and would have a flow rate of ∼2500 L/min assuming a typical 800 tons heavy metal/yr reprocessing facility. Kr separation from Xe is one of the steps in a proposed UNF off-gas processing flowsheet. The Kr/Xe mixture is approximately 10/90 by volume and hence the separation of Kr would reduce the volume of radioactive {sup 85}Kr waste by a factor of 10. Kr/Xe separation is challenging since they are both chemically inert species and physically very similar (0.036 nm size difference). There is no existing process for this separation that is suitable for use in UNF recycling since conventional cryogenic distillation requires highly intensive power and large system volume. Membrane-based separation using radiation-stable materials such as zeolites could be an attractive separation method, if membranes with sufficient Kr flux and selectivity can be fabricated. Zeolite membranes offer an opportunity to exploit the size difference of Kr (kinetic diameter 0.36 nm) and Xe (0.396 nm) to achieve preferential permeation of the minority component (Kr) from the mixture, leading to an efficient separation with very low energy requirements. An ideal membrane process would permeate a stream of pure Kr for storage and also produce pure Xe which has commercial value (> $8000/kg at 99+% purity). SAPO-34 is a silico-alumino-phosphate zeolite molecular sieve with CHA framework topology and containing uniform micropores (0.38 nm), intermediate in size between the kinetic diameters of Kr and Xe. It was hypothesized that SAPO-34 membranes can separate krypton and xenon by diffusivity differences. Previously we fabricated SAPO-34 membranes and showed that they are capable of separating Kr from Xe. Our previous work also indicated that there is much potential for engineering the separation properties of these membranes with a view to improving permeance and selectivity and evaluating the expected process performance. The main objectives of the present study are: 1) development of SAPO-34 membranes with enhanced membrane performance via thickness reduction and ion exchange, and 2) process modeling for technic-economic analysis of zeolite membrane systems for Kr/Xe separation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Effect of Moisture on the Adsorption of Iodine and Krypton in Multicomponent Streams on the 10 WT% C/ETS-10 Sorbent

Over 13% of the world's energy demand is supplied by nuclear power plants in 31 countries. This demand is growing; by 2030 the IAEA predicts a 100% increase in the demand for nuclear energy. The increase in nuclear energy production imposes a growing strain on the uranium fuel supply. At current consumption rates the world's supply of uranium is projected to last ∼200 years. With the growing energy consumption, uranium supplies will dwindle even more quickly. However, ∼95% of uranium is left in used nuclear fuel (UNF) and may be utilized through reprocessing. A main concern with reprocessing is the release of off-gas. The major constituents of off-gas include {sup 85}Kr, {sup 129}I, {sup 14}C, {sup 3}H, and various Xe isotopes. {sup 85}Kr and {sup 129}I present the largest threat due to their harmful environmental effects and abundance. Several methods have been studied for the removal of these contaminants, including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Adsorption on porous solid sorbents has the advantages of being simple, effective, and economic. Thus many studies have been performed investigating several sorbents such as silver impregnated mordenite, silver Engelhard titanosilicate 10, chalcogen-based aerogels, and metal organic frameworks. Despite these advances, many of these studies have been conducted under ideal conditions: cryogenic temperature, helium as a carrier gas, and single component adsorption. The need for investigating adsorption under conditions more representative of practical use is needed. Thus our work has been devoted to the study of adsorption of {sup 85}Kr and {sup 129}I from a simulated off-gas on an Engelhard titanosilicate 10 (ETS-10) supported carbon nanopolyhedron sorbent (C/ETS-10) in a multicomponent system at ambient temperatures. Previously we have reported sorption capacities of C/ETS-10 for single component krypton, and iodine. The effects multi-contaminant iodine and krypton and their respective inlet concentrations on the sorption capacity of 10 wt% C/ETS-10 have also been studied. Investigations into the effect of the presence of moisture in the multicomponent gas mixture on the sorption on 10 wt% C/ETS-10 are described in this paper.

36 MATERIALS SCIENCE↗