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Stability, Composition, and Core-Shell Particle Structure of Uranium(IV)-Silicate Colloids

Uranium is typically the most abundant radionuclide by mass in radioactive wastes and is a significant component of effluent streams at nuclear facilities. Actinide(IV) (An(IV)) colloids formed via various pathways, including corrosion of spent nuclear fuel, have the potential to greatly enhance the mobility of poorly soluble An(IV) forms, including uranium. This is particularly important in conditions relevant to decommissioning of nuclear facilities and the geological disposal of radioactive waste. Previous studies have suggested that silicate could stabilize U(IV) colloids. Here the formation, composition, and structure of U(IV)-silicate colloids under the alkaline conditions relevant to spent nuclear fuel storage and disposal were investigated using a range of state of the art techniques. The colloids are formed across a range of pH conditions (9-10.5) and silicate concentrations (2-4 mM) and have a primary particle size 1-10 nm, also forming suspended aggregates <220 nm. X-ray absorption spectroscopy, ultrafiltration, and scanning transmission electron microscopy confirm the particles are U(IV)-silicates. Additional evidence from X-ray diffraction and pair distribution function data suggests the primary particles are composed of a UO2-rich core and a U-silicate shell. U(IV)-silicate colloids formation correlates with the formation of U(OH)3(H3SiO4)3 2- complexes in solution indicating they are likely particle precursors. Finally, these colloids form under a range of conditions relevant to nuclear fuel storage and geological disposal of radioactive waste and represent a potential pathway for U mobility in these systems.

Neill, Thomas↗

Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

A Study Of Geological Disposal Facility (GDF) Ageing Phenomena In Bentonite - 17605

Since the 1950's, the UK's nuclear activity in energy, defence and medicine has led to the accumulation of a large amount of higher activity radioactive waste (HAW). Most of nuclear waste arises in a wide variety of physical and chemical forms with different radioisotopes and loadings. Due to complexity in HAW, it is imperative that there be a rigorous management system that ensures population and environmental safety over extended periods. The concept of Deep Geological Disposal of waste in a stable geological formation is, nowadays, an internationally accepted solution as the surest and the most viable long term method for management. Any future Geological Disposal Facility (GDF) will comprise, among other artificial and natural engineered barriers, a barrier immediately around the waste containers compounded by a sealing material. The long-term stability of a GDF will be determined by the design and construction of the engineered barriers, and the mechanical, hydraulic and geochemical interactions that may happen between the different engineered components as a result of combined processes such as the heat produced by radioactive decay and the ion water composition of the surrounding host rock. Bentonite clay has been selected as the most promising buffer and backfill material in many disposal programmes because of its swelling capacity, low permeability, plasticity and cation exchange capacity. In the nuclear scope, bentonite testing at different scale-lengths (laboratory and field scale) has been performed and implemented in order to examine the thermo-hydro-mechanical (THM) and thermo-hydro-geochemical (THG) processes that may happen in the engineered barriers and geological settings associated with a GDF. Two major European field scale tests have been performed to better understand bentonite performance and behaviour immediately after HAW emplacements. The Full-Scale Engineered Barriers Experiment (FEBEX) at the Grimsel Test Sit (Switzerland) and the Alternative Buffer Material (ABM) Project at the Aespoe Hard Rock Laboratory (Sweden). The Febex Experiment was a pilot project designed for 'field research' of HAW repository hosted in crystalline rock where surrogate HAW canisters were placed horizontally in drifts, surrounded by compacted blocks of bentonite, sealed by a concrete plug and maintaining a constant canister surface temperature of 100 deg. C. After five years of operation, a partial dismantling was carried out in 2002. Heater processes occurring in the surrounding bentonite. The dismantling of the heater no. 2 (after 18 years of operation) started in late 2014 and is planned to be finished at the end of 2016 (termed FEBEX-DP). The FEBEX Project utilised bentonite from the Cortijo de Archidona deposit (Almeria, Spain). This 'Febex' bentonite is sourced from calcium-magnesium montmorillonite with good properties in relation to swelling capacity, permeability, adsorption capacity, and thermal conductivity. The ABM Project is a large-scale programme that has been running since 2006. The programme is designed to study a suite of possible buffer materials and correlate the physical and chemical properties of eleven different types of bentonites to their fundamental mineralogical properties. The ABM test consists in three 'packages' containing a central heater tube surrounded by different pre-compacted bentonite blocks and instrumentation. Each 'package' was installed in separate boreholes. The target temperature for the surface of the heated tubes in contact with the bentonite was 130 deg. C. Both FEBEX and ABM test the bentonite was saturated with water percolating from surrounding host rock (FEBEX and ABM) and also via artificial supply systems (ABM), being the water itself acting as a long term corrosion agent. For both experiments and emerging interest is in the interaction of bentonite with metallic components, under geological disposal conditions are not considered to be thermodynamically stable and hence arising corrosion products could interact with the buffer. These interactions depend on various factors such as the redox potential, temperature, pH, solute concentration and buffer material characteristics. During the operational phases of the FEBEX and ABM and after its 'closure' a significant volume of air was found to be retained in the bentonite. Under these conditions a period of initial aerobic corrosion by the air in the bentonite barrier would have occurred followed by establishment of anaerobic corrosion would be the dominant process ongoing. The anaerobic corrosion began when the bentonite became water saturated. Under these conditions, the water acted as the major oxidant. The present project focuses on the characterization of samples arising from the FEBEX and ABM Projects as well as performing additional independent experiments designed to complement these prior studies. Characterization will focus mainly on corrosion related alteration combined with thermal effects. Through detailed sample characterization, the intention is to achieve a greater knowledge regarding the changes to the bentonite under the experimental conditions to which it was exposed, and plausible alterations in its THM and THG properties occurring during the experiment due to blended effects of temperature,water content, engineering structural components and solutes are also to be measured. The fulfilment of a large-scale 'in-situ' test is intricate, time-consuming and with an abundance of external variables that are not always well controlled. For this reason, the University of Bristol is performing a series of well constrained time-resolved laboratory tests which bentonite is being subjected to different possible scenarios that may happen whilst in a HAW repository. In virtue of this experiment, the UoB aims to understand the illitisation and cementation process and their effects on bentonite safety properties. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Catalytic oxidation of waste materials

Aqueous stream of human waste is mixed with soluble ruthenium salts and is introduced into reactor at temperature where ruthenium black catalyst forms on internal surfaces of reactor. This provides catalytically active surface to convert oxidizable wastes into breakdown products such as water and carbon dioxide.

Jagow, R. B.↗

Development of a specific methodology to assess suitable sites to receive a repository for L/ILW waste in the Portuguese territory

Portugal does not have nuclear power plants but records an increasing production of radioactive waste from medical, industrial and research applications of radioactive materials in the form of sealed and unsealed sources; the country totals include also the spent fuel of one nuclear reactor for research purposes. Since radioactive waste management policies and practices in Portugal will have to comply with the Council Directive 2011/70/Euratom and the IAEA Joint Convention, the search for scientific and technological solutions to deal with radioactive wastes produced in the country started some years ago. The research carried out recently under the scope of a national funded project (KADRWaste, PTDC/CTEGEX/ 82678/2006) represents a firm step in this commitment. Indeed, the transfer of methodologies tested and validated in this project allowed for improvement to a procedure to assess suitable sites to receive a near-surface repository for 'Low and Intermediate Level Waste, Short Live and Long Live' (LILWSL, LL) wastes in Portugal mainland. Although the main stages of the procedure can be of universal usage, details were designed according to the intrinsic geological, geomorphic and meteorological features of previously selected target-areas. All the requirements exhaustively listed in many reports of the IAEA were fulfilled and, in addition, the application of mineralogical, geochemical and textural criteria is strongly advised. The proposed procedure is based on a 5 key-steps approach preceded by clarification of the boundary conditions to be imposed, which are crucial to the inventory of various compulsory technical requirements. This analysis requires the adoption of stringent criteria, many of them of multidisciplinary nature, including tests of vulnerability and assessment of uncertainty, besides the environmental impact risk. As a result, priority targets that are not excluded will integrate different classes and, depending on the existing knowledge, it will be possible to select locations suitable for the repository installation, taking into account also the political, social and administrative dimensions behind this decision. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Disposition Options for Sodium Cooled Fast Reactor (A White Paper)

The sodium-cooled fast reactor (SFR) design concept is one of the six classes of nuclear reactors in the GenIV initiative. SFRs are uranium or plutonium-fueled reactors operating in the fast neutron spectrum using liquid sodium as the coolant. SFRs can be designed as a breeder reactor or actinide-burning reactor in addition to operating the thorium fuel cycle. While having different fuel designs, the anticipated waste streams, and the necessary management strategies for spent nuclear fuel (SNF) and radioactive wastes from SFRs are very similar. This includes the SNF, activated sodium coolant, in-core stainless-steel components, piping, resins and filters, solidified liquid waste, contaminated equipment, and other radioactive wastes. Modern SFR designs are based on a long and rich operating history of several liquid-metal-cooled fast reactors with sodium coolant. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the SNF and radioactive wastes associated with operating these reactors. This white paper will identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste and identify the key radioactive waste streams from SFRs. Idaho National Laboratory has significant experience in the management of the SNF from the SFR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new SFR-type small modular reactor or microreactor intended for deployment at the Idaho National Laboratory Site

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Regulation of the Shutdown Defueling and Decommissioning Reactors in the United Kingdom - 16576

Nuclear policy, including for decommissioning, in the United Kingdom is set by the UK Government. The Nuclear Decommissioning Authority, NDA, is responsible for managing the clean-up of the UK's nuclear legacy which includes the first generation of Magnox power stations. Magnox Ltd (MXL), owned by Cavendish Fluor Partnership, is the licensee and the management and operations contractor responsible for ten Magnox reactor sites and two restoration sites in the UK, working for the sites' owner, NDA. The work of the company is diverse covering electricity generation at Wylfa (until December 2015), defueling, and preparation for an interim end state, known as Care and Maintenance (C and M) which will precede the final site clearance. The C and M phase is expected to last over 75 years during which the sites are maintained in a passively safe and secure state while radiation levels are left to decay naturally. Preparations for this phase include removing fuel from the reactors and storage ponds; retrieving intermediate level waste from current stores and treating it or converting it to forms suitable for long-term storage or disposal and making sure of the integrity of critical safety systems of the de-fuelled reactors. The Decommissioning, Fuel and Waste programme within the Office for Nuclear Regulation, ONR, regulates nuclear safety in the 12 Magnox nuclear licensed sites. ONR works with the environmental regulators who are responsible for environmental protection on nuclear sites. The focus of this paper is the regulation of nuclear safety by ONR across the Magnox sites and the interactions with nuclear security and the environmental aspects required to achieve effective, proportionate and efficient regulation of these sites in the UK. ONR's strategy for regulating the Magnox defueling and decommissioning sites is focused on enabling the licensee to complete the initial stages of decommissioning and reduce the nuclear and radioactive hazards on its sites. To this effect, ONR's revised inspection guidance is centred on inspecting the (remaining) important systems for safety identified in the safety cases. Other interventions focus on engaging with the licensee on safety significant decommissioning projects to gain evidence of the licensee's readiness prior to and during these activities rather than formal assessment of the associated safety cases. In light of reduced hazards and risks across the de-fueled and decommissioning Magnox reactors, ONR's revised regulatory expectation regarding evolution of emergency arrangements during decommissioning is based on key transition points that follow electricity generation. Implementation of these revised strategies as well as organisational restructuring within the team responsible for regulation of the Magnox fleet has resulted in improved efficiency and consistency in regulatory approach whilst securing the required safety standards in managing decommissioning activities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Sustainability of carbon delivery to an algal biorefinery: A techno-economic and life-cycle assessment

Supplementation of carbon is critical for high productivity cultivation of most microalgae. Moreover, using microalgae for atmospheric CO 2 mitigation to combat climate change is promising as waste sources and atmospheric CO 2 can be utilized to produce useful products. The challenge is developing technologies, processes, and strategies that utilize carbon efficiently such that the overall system is sustainable. Through engineering system modeling combined with techno-economic and life-cycle assessments, this study examined the implications of various delivery methods of carbon to a production-scale algal biorefinery. Five primary carbon sources were considered: atmospheric CO 2 ; CO 2 from direct chemical or power plant waste emissions; CO 2 that has been concentrated from waste sources and compressed; inorganic carbon in the form of hydrogen carbonate; and organic carbon in the form of cellulosic sugars derived from corn stover. Each source was evaluated assuming co-location as well as pipeline transportation up to 100 km. Sustainability results indicate that economics are more prohibitive than energy and emissions. Furthermore, of the scenarios evaluated, only two met both the economic and environmental criteria of contributing less than 0.50 US-$ GGE -1 and 20 g CO 2-eq MJ -1 to the overall system, respectively: uncompressed, pure sources of gaseous CO 2 with pipeline transportation of 40 km or less; and compressed, supercritical CO 2 from pure sources for pipeline transportation up to 100 km. A first order scalability assessment of algal biofuels based on these results shows carbon to be a limiting nutrient in an algal biorefinery with a total US production capability of 360 million gallons of fuel per year.

42 ENGINEERING↗

Spectral changes in Si–O–Si stretching band of porous glass network upon ingress of water

Water ingress into porous glass can induce internal stress, resulting in strains in chemical bonds of the network. For silica and silicate glasses, the position of Si–O–Si stretching mode in infrared spectroscopy is known to vary with the degree of strain in the Si–O network. Then, one could hypothesize that the stress due to water ingress could be probed with infrared spectroscopy. We tested this hypothesis using porous layers formed through aqueous corrosion on a model nuclear waste glass. The porosity and thickness of the porous layer were determined using ellipsometry. The humidity-dependent infrared spectra of the samples showed red-shifts of the Si–O–Si stretching band; however, it was difficult to deconvolute the spectral change due to variation of effective refractive index of the sample upon water ingress. Thus, it was infeasible to unambiguously determine the internal stress of porous glass network upon water ingress using the infrared spectroscopy alone.

36 MATERIALS SCIENCE↗

Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Criticality Safety Evaluation for 700 and 350 Gram Moderated Fissionable Equivalent (MFE) Criticality Control Areas (CCAs)

This evaluation is intended to generically apply to facilities and/or areas designated as CCAs that store and/or handle greater than 15 grams of fissionable material but no more than 700 grams of MFE. This evaluation includes facilities and/or areas designated as CCAs limited to no more than 350 grams of MFE. Activities performed in these CCAs may include general laboratory activities to include work in radiological areas, chemical procurement, material handling, sampling, imaging, non-destructive analysis, and disposing of nuclear and radiological materials. Materials included in these activities may include radioactive-contaminated oil, polychlorinated biphenyls (PCBs), and/or equipment. Fissionable materials may consist of transuranic (TRU) waste, sources containing fissionable material, uranium in various forms such as metal, oxide and silicide, however other forms of uranium and plutonium may be present. Fissionable materials may consist of fines from cutting, be solid forms mounted on a ring and embedded in epoxy or be in small volumes of dilute solutions resulting from cleaning or other processes. Major equipment located in these CCAs may include inert gas environment gloveboxes, fume hoods, ultrasonic testing equipment, hot presses, rolling mills, furnaces, gamma irradiators, autoclaves, machining equipment, tensile testing equipment, shears, forklifts and cranes.

07 ISOTOPE AND RADIATION SOURCES↗

Reinforced radiological containment bag

Radiological containment bags for use in storing alpha particle emitting waste are described. The radiological containment bags are formed of a multilayer film with an inner layer of the film being a sacrificial layer that lies between the interior of the bag and a structural polymeric layer. The sacrificial layer includes a flexible composite material that in turn includes a polymeric matrix and an alpha particle energy absorber incorporated in the polymeric matrix. The sacrificial layer can include a chromophore that can detect early degradation of the sacrificial material.

Washington, II, Aaron L.↗

Characterization Techniques Investigated for Characterization of Anomalous Materials in Plutonium Oxide – 25666

The DOE has adopted a Dilute and Dispose approach for processing surplus plutonium which consists of blending plutonium oxide with adulterants and packaging it in a form which is acceptable for disposal at the Waste Isolation Pilot Plant. The feed material for the downblend process is intended to be pure plutonium oxide powder, however other objects are occasionally encountered within the oxide, particularly with legacy material. Potential technologies which could assist in the resolution of incidents where such anomalies are encountered have been investigated. Resolution of these incidents requires characterization of the anomalous object so that plutonium oxide processing can resume and so that a disposition pathway can be determined for the anomalous material. Technologies including gamma ray spectroscopy, alpha-gamma coincidence, LiDAR volumetric measurements, and surface conductivity measurements were investigated for this purpose, with a focus on systems which can be easily introduced into the processing environment when an anomaly is encountered, then removed from the environment after resolution to avoid impeding normal processing activities. This places an emphasis on small and portable measurement systems and systems that can operate in a high-background, oxide-processing environment. Gamma ray systems investigated include the GR1™a CZT detector and the MicroGe™a germanium detector, with a focus on detecting characteristic gamma rays from Pu-239 and other actinides. A gamma ray and alpha particle coincidence method was investigated with the goal of identifying actinides in the presence of a gamma ray background produced by adjacent plutonium oxide material. Leica™b BLK360 G1 and Keyence™c LJ-X8300 LiDAR systems were investigated for use in conjunction with mass measurements to gain accurate material density values, and a Foerster Sigmatest™d 2.070 was tested to determine surface conductivity. The combination of these properties would allow improved identification and characterization of a wide variety of potential anomalous material.

Munson, Justin M.↗

Reactions of Studtite UO 4 ·4H 2 O in Alkali Hydroxides: Isolation of Single-Crystal Uranate Phases under Mild Hydrothermal Conditions

Uranyl peroxide complexes, particularly studtite (UO 4 ·4H 2 O), are important phases within the nuclear fuel cycle, forming through radiolysis-induced reactions on the surfaces of spent fuel and in waste environments. Studtite has been identified in Hanford’s irradiated fuel storage basins, Chornobyl’s corium lavas, and is anticipated on Fukushima’s fuel debris. While its formation and stability have been extensively studied, the reactivity of UO 4 ·4H 2 O in highly alkaline environments such as those encountered in high-level nuclear waste remains underexplored. These environments contain molar concentrations of [OH - ] and present a chemically dynamic and reactive environment where actinide behavior is not well understood. Reported here are investigations of uranium reactions in concentrated alkali hydroxides under mild hydrothermal conditions (<200 °C) that have resulted in the isolation of the alkali metal uranates Li 2 UO 4 , α-Na 2 UO 4 , γ-Na 2 U 2 O 7 , and K 2 U 2 O 7 . In contrast to conventional solid-state methods (>800 °C) that typically yield polycrystalline powders, our approach enables the isolation of these uranates as single crystals, allowing us to provide single-crystal structure solutions of certain uranates for the first time. Our findings match the results of high-level waste tank sampling, confirming that sodium diuranate (Na 2 U 2 O 7 ) is the most persistent uranium phase, potentially forming through reactions of uranyl peroxide intermediates with NaOH under radiolytic and highly alkaline conditions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

New Perspectives in Thermoelectric Energy Recovery System Design Optimization

Abstract: Large amounts of waste heat are generated worldwide in industrial processes, automotive transportation, diesel engine exhaust, military generators, and incinerators because 60-70% of the fuel energy is typically lost in these processes. There is a strong need to develop technologies that recover this waste heat to increase fuel efficiency and minimize fuel requirements in these industrial processes, automotive and heavy vehicle engines, diesel generators, and incinerators. There are additional requirements to reduce CO2 production and environmental footprints in many of these applications. Recent work with the Strategic Environmental Research and Development Program office has investigated new thermoelectric (TE) materials and systems that can operate at higher performance levels and show a viable pathway to lightweight, small form-factor, advanced thermoelectric generator (TEG) systems to recover waste heat in many of these applications. New TE materials include nano-composite materials such as lead-antimony-silver-telluride (LAST) and lead-antimony-silver-tin-telluride (LASTT) compounds. These new materials have created opportunities for high-performance, segmented-element TE devices. New higher-performance TE devices segmenting LAST/LASTT materials with bismuth telluride have been designed and fabricated. Sectioned TEG systems using these new TE devices and materials have been designed. Integrated heat exchanger/TE device system analyses of sectioned TE system designs have been performed creating unique efficiency-power maps that provide better understandings and comparisons of design tradeoffs and nominal and off-nominal system performance conditions. New design perspectives in optimization of sectioned TE design approaches are discussed that provide insight on how to optimize such sectioned TE systems. System performance analyses using ANSYS® TE modeling capabilities have integrated heat exchanger performance models with ANSYS® TE models to extend its analysis capabilities beyond simple constant hot-side and cold-side temperature conditions . Analysis results portray external resistance effects, matched load conditions, maximum power vs. maximum efficiency points simultaneously.

Energy Conversion↗

THE EFFECT OF THE PRESENCE OF OZONE ON THE LOWER FLAMMABILITY LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'.

08 HYDROGEN↗

The Effect of the Presence of Ozone on the Lower Flammability Limit (LFL) of Hydrogen in Vessels Containing Savannah River Site High Level Waste - 12387

The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'. The LFL of hydrogen in air was determined and is in good agreement with the literature data. Ozone-oxygen mixtures were found to be flammable at concentrations above 8.3 vol.% based on the ASTM E918 7% pressure rise criteria for flame propagation. This result is lower than previously reported values which can be explained through the variations in the test setup and procedure. It is believed that the lower values obtained in this work are a result of improvements of the test methodology. Tests performed with hydrogen in various concentrations of ozone in oxygen have shown that the LFL of hydrogen decreases as the concentration of ozone in the mixture increases. This testing was designed to provide data under the conditions considered most optimal to produce deflagration. The geometry and materials of construction of the testing vessel; the location of the fuse wire; the magnitude of the supplied energy; the careful minimization of diluents and other contaminants; and meticulous procedural detail to maintain integrity of the ozone to the maximum extent practical, result in data that reflect not the expected process conditions, but those that enhance the possibility of flame propagation. For this reason, there is believed to be considerable conservatism in the indicated results. Per the vendor, the maximum possible ozone concentration producible by the planned ECC Ozone generator is 8 volume percent (the typical maximum operating setpoint concentration is 6.8 vol%), less than the 8.3 minimum volume % concentration shown to be flammable in a 99.999% pure O{sub 2} environment at the optimally conservative conditions established in this testing. Further, the feed to the ECC ozone generator is only 87% oxygen, the remainder, water vapor and nitrogen, both powerful diluents. It is, thus, believed not credible that deflagration can occur at this maximum feed concentration condition. In addition, once the ozone stream contacts the waste stream, the many simultaneous oxidizing reactions will rapidly decompose available ozone to well below flammable levels. Further, because the radiolytically generated hydrogen quantity is negligible compared to the supplied ozone/oxygen stream (0.0004 moles per minute H{sub 2} vs 76 moles per minute ozone/oxygen), even the total H{sub 2}/O{sub 3} mixture, without crediting decomposition reactions, does not approach flammable concentrations. Finally, even at the 'end' of the ECC batch cycle, when most of the metal oxalates have been decomposed, testing has indicated that the ozone concentration in the vapor space of the ECC process vessel reaches a concentration of no more than 3 vol%, remaining well below concentrations of concern. The major issue for the ECC operation established by this testing is the impact of the data when applied to off normal conditions. While it is possible to discontinue ozone addition to the reaction vessel at any time, the radiolytic hydrogen generation rate continues, varying slightly as ambient pressures and temperatures change. Relative to the data generated and analyzed in this testing, the ECC hazards analysis team will re-evaluate off normal conditions (e.g. those during which process exhaust ventilation is lost) such that issues involving mixtures of hydrogen and ozone in the vapor space can be appropriately controlled. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗