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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.
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)
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
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)
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.
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.
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.
In proposed method, nuclear and chemical waste quickly encapsulated in hollow silicon spheres by forming, filing and sealing sphere. Process entirely automated.
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.
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.
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.
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.
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.
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 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)
An analytical method is required to distinguish between the pertechnetate and non-pertechnetate forms of technetium; currently, the methods available only report the total technetium present in a sample. The overall objective of this effort is to develop a method for routinely analyzing Hanford tank waste for technetium in the pertechnetate and the non-pertechnetate forms. A phased approach will be deployed to accomplish this objective: Phase I Comparison of existing technetium analysis methods with modification; Phase II Demonstration of modified methods using non-pertechnetate spiked simulants; and, Phase III Demonstration of chosen method on Hanford tank sample containing non-pertechnetate. This report describes the Phase I work, providing a comparison of Aliquat 336 and TEVA(R)1 in the removal of pertechnetate and discussing the subsequent analysis for technetium in both alkaline and acidic environments without oxidation. The effort was executed under LAB-PLN-13-00004, The Determination of Pertechnetate and Non-Pertechnetate Species in Hanford Tanks Phase I.
The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. 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 spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.
The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria and, specifically, the ETF Delisting Modification treatability envelope limits. Consequently, there is a need to assess possible mitigation approaches, one of which involves identification and assessment of potential treatment processes to destroy the acetonitrile, which is the subject of the present report.