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

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Stack Characterization System Development and Testing

Oak Ridge National Laboratory, as well as the rest of the U.S. Department of Energy community, has numerous off-gas stacks that need to be decommissioned, demolished, and packaged for disposal. Disposal requires a waste disposition determination phase. Process knowledge typically makes a worst-case scenario decision that may place lower-level waste into a more expensive higher-level waste disposal category. Truly useful radiological and chemical sampling can be problematic on old stacks due to their inherent height and access hazards, and many of these stacks have begun to deteriorate structurally. A remote stack characterization system (SCS) that can manage sample and data collection removes people from the hazards and provides an opportunity for access to difficult to reach internal stack areas. The SCS is a remotely operated articulated radiological data recovery system designed to deploy down into off-gas stacks from the top via crane. The battery-powered SCS is designed to stabilize itself against the stack walls and move various data recovery systems into areas of interest on the inner stack walls. Stabilization is provided by a tripod structure; sensors are mounted in a rotatable bipod underneath the tripod. Sensors include a beta/gamma/alpha detector, a removable contaminant multi-sample automated sampler, and a multi-core remote core drill. Multiple cameras provide remote task viewing, support for sampling, and video documentation of the process. A delay in funding has delayed project delivery somewhat. Therefore, this paper describes the technology and shows fabrication and testing progress to the extent that data is available.

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

Classification of Dissolution Events Using Fusion of Effluents Measurements and Classifiers

Classifiers for dissolution events at a radiochemical processing facility are studied using gamma spectra measurements of effluents collected by a high purity germanium detector located at its off-gas stack. Data sets collected at the Oak Ridge National Laboratory’s Radiochemical Engineering Development Center under a Pu dissolution campaign spanning a three months period are utilized. Features corresponding to the activity levels of 15 radionuclides, including isotopes of iodine, krypton, and xenon, that are indicated by the target decay chains, are computed from the spectra at 1 hour intervals. A conceptualization diagram is developed to reflect the steps from the source to measurement to feature computation that depend on fission products indicated by decay chains, chemical processing, and effluents transport to the off-gas stack. A diverse set of eight classifiers based on different design principles are trained using the ground truth data for this campaign, and the outputs of top three classifiers, namely, classification trees, Ensemble of Trees (EOT), and k-nearest neighbor, are combined using EOT classifier-fuser. Our results show that for 5-fold cross validation, features associated with isotopes of xenon provide the lowest classification error among the different elements across the classifiers; the classification error is furthered improved when all 15 isotope features are used by each classifier, and it is again improved by the fusion of three classifiers. Further reduction in classification error is achieved by using a measurement window of 1-2 days which is identified based on half-life time estimates of the isotopes; it is long enough for the stabilization of feature estimates while being short enough not to be affected by the follow on dissolution events. As a net result of feature and classifier fusion, combined with the incorporation of decay chain and isotope half-life information, this approach achieves 98% detection rate while maintaining a false alarm rate under 2% for this data set.

Rao, Nageswara↗
DOE OSTI2011

Remote System for Characterizing, Monitoring and Inspecting the Inside of Contaminated Nuclear Stacks

The Stack Characterization System (SCS) is a collaborative project with the Robotics and Energetic Systems Group (RESG) at Oak Ridge National Laboratory (ORNL) and the Applied Research Center (ARC) at Florida International University (FIU). The SCS is a robotic system that will be deployed into off-gas stacks located around the central campus at ORNL. The system will consists of surveying equipment capable of taking surface contamination samples, radiation readings, core samples and transmit live video to its operators. Trade studies were conducted on varying concrete materials to determine the best way of retrieving loose contamination from the surface. The studies were performed at the ARC facility by DOE Fellows, where traditional cloth wipes were compared to adhesive material. The adhesive material was tested on the RESG s smear sampler to record how much loose surface material could be retrieved. The DOE Fellows completed a summer internship during which conceptual designs were created for a deployable radiation detector and core drill capable of retrieving multiple core samples.

Vargas, Mario↗
DOE OSTI2018

On/Off State Classification of a Reactor Facility Using Gas Effluence Measurements

Inferring on/off operational state of a reactor facility using measurements from an independent monitoring system is critical to the assessment of its compliance to agreements. In this paper, we consider the problem of inferring on/off state of a reactor facility using measurements of Ar-41, Cs-138, and Xe-138 gas effluence types collected at the facility’s off-gas stack. We present classifiers based on thresholding measurements of individual effluence types, and then present fusers that combine their outputs or measurements. We present five fusers based on simple majority rule, Chow’s pattern recognition function, Fisher’s combined p-value statistic, physics-based Poisson radiation counts model, and correlation coefficient method. In addition, we also test five machine learning methods based on non-linear classifiers, which are available as R packages. Finally, our results show that: (i) these gas effluence measurements are effective in inferring on/off state of a reactor facility; for example, best fusers achieve ~97% detection at ~1% false alarm rate, and (ii) fusers that combine all effluence types based on physics-based models, correlation coefficient and Fisher’s method outperform simple majority rule, Chow’s fusers and machine learning methods, as well as when they are applied to individual and pairs of effluence types.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Facility Activity Inference Using Radiation Networks

We consider the problem of inferring the operational status of a reactor facility using measurements from a radiation sensor network deployed around the facility’s ventilation off-gas stack. The intensity of stack emissions decays with distance, and the sensor counts or measurements are inherently random with parameters determined by the intensity at the sensor’s location. We utilize the measurements to estimate the intensity at the stack, and use it in a one-sided Sequential Probability Ratio Test (SPRT) to infer on/off status of the reactor. We demonstrate the superior performance of this method over conventional majority fusers and individual sensors using (i) test measurements from a network of 21 NaI detectors, and (ii) effluence measurements collected at the stack of a reactor facility. We also analytically establish the superior detection performance of the network over individual sensors with fixed and adaptive thresholds by utilizing the Poisson distribution of the counts. We quantify the performance improvements of the network detection over individual sensors using the packing number of the intensity space.

Rao, Nageswara S.↗
DOE OSTI2019

Classifiers for Dissolution Events in Processing Facility Using Effluents Measurements

Inferring occurrences of dissolutions at a radiochemical processing facility using measurements from an independent monitoring system has potential for verifying operator declarations and compliance with nuclear technology use agreements. We consider the identification of such dissolution events using measurements of radioactive effluents collected at the facility’s off-gas stack. By using simple models for the source term based on radioactive decay-chains and sensor measurements of transported effluents, we formulate a simplified on/off classification problem. There are many available machine learning classification methods applicable to this problem, but they are subject to potential over-fitting. We mitigate this effect by fusing four promising but disparate classifiers: the ensemble of trees, support vector machines, naive Bayes, and k-nearest neighbors. Each classifier is trained to distinguish using effluents measurements between dissolutions of irradiated 237Np/238Pu and Cm/Cf targets from other (or no) radiochemical processing activities. We present Chow and ensemble of trees methods to fuse the classifier outputs to produce a single classification output, and thereby incorporate their diversity of design. We study the performance of these classifiers and fusers using measurements of87Kr,135Xe,138Xe, and138Cs isotopes,and present their training accuracy estimates using individual isotope measurements and all four isotopes taken together.

Rao, Nageswara S.↗
DOE OSTI2014

Air Sample Conditioner Helps the Waste Treatment Plant Meet Emissions Standards

The air in three of the Hanford Site Waste Treatment and Immobilization Plant (WTP) melter off-gas discharge stacks will be hot and humid after passing through the train of emission abatement equipment. The off-gas temperature and humidity levels will be incompatible with the airborne emissions monitoring equipment required for this type of stack. To facilitate sampling from these facilities, an air sample conditioner system will be installed to introduce cool, dry air into the sample stream to reduce the temperature and dew point. This will avoid thermal damage to the instrumentation and problematic condensation. The complete sample transport system must also deliver at least 50% of the particles in the sample airstream to the sample collection and on-line analysis equipment. The primary components of the sample conditioning system were tested in a laboratory setting. The sample conditioner itself is based on a commercially-available porous tube filter design. It consists of a porous sintered metal tube inside a coaxial metal jacket. The hot gas sample stream passes axially through the porous tube, and the dry, cool air is injected into the jacket and through the porous wall of the inner tube, creating an effective sample diluter. The dilution and sample air mix along the entire length of the porous tube, thereby simultaneously reducing the dew point and temperature of the mixed sample stream. Furthermore, because the dilution air enters through the porous tube wall, the sample stream does not come in contact with the porous wall and particle deposition is reduced in this part of the sampling system. Tests were performed with an environmental chamber to supply air with the temperature and humidity needed to simulate the off-gas conditions. Air from the chamber was passed through the conditioning system to test its ability to reduce the temperature and dew point of the sample stream. To measure particle deposition, oil droplets in the range of 9 to 11 micrometer aerodynamic diameter were injected into the environmental chamber and drawn through the conditioning system, which included a filter to capture droplets that passed through the conditioner. The droplets were tagged with a fluorescent dye which allowed quantification of droplet deposition on each component of the system. The tests demonstrated the required reductions in temperature and moisture, with no condensation forming when heat tracing was added on the upstream end of the sample conditioner. Additionally, tests indicated that the system, operating at several flow rates and in both vertical and horizontal orientations, delivers nearly all of the sampled particles for analysis. Typical aerosol penetration values were between 98 and 99%. PNNL, Bechtel National Inc., and the instrument vendor are working to implement the sample conditioner into the air monitoring systems used for the melter off-gas exhaust streams. Similar technology may be useful for processes in other facilities with air exhaust streams with elevated temperature and/or humidity.

sample conditioning↗
DOE OSTI2018

Reactor Facility Operational State Classification Using Gas Effluents and Radiation Measurements

Inferring the operational state of a reactor facility, using measurements from an independent in-situ monitoring system, is critical to the assessment of its compliance to agreements. We consider the problem of inferring the on/off operational state of a facility by using gas effluents and radiation measurements collected at the reactor's off-gas ventilation stack and within its surrounding area. Effluent release concentrations of three noble gases, namely, Ar-41, Cs-138, and Xe-138, are collected inside the ventilation stack and provided every four hours. Gamma spectra is recorded every second by a radiation detector located about 100 meters offset from the stack, which is affected by weather and nearby conditions. We study classifiers based on the correlation coefficient (CC) method [1] applied to individual modalities and their combinations. We implement CC classifiers using ground truth measurements collected over 2 operational cycles of High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) in year 2017.

Ramirez Aviles, Camila A.↗
DOE OSTI2020

GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} for Fukushima Daiichi Water Treatment Secondary Wastes - 20212

The Japanese government is supporting development work implemented jointly by Veolia subsidiaries Kurion Japan, K.K., Veolia Nuclear Solutions (VNS), Inc., and Veolia Nuclear Solutions Federal Services, LLC for treating radioactive waste generated from Fukushima Daiichi Nuclear Power Station (NPS) water treatment using the GeoMelt{sup TM} In-Container Vitrification (ICV){sup TM} technology. The initial work consisted of glass formulation and engineering-scale testing which was completed in 2018, in the frame of an IRID (International Research Institute for Nuclear Decommissioning) program as part of a project subsidized by Japan's Ministry of Economy, Trade and Industry (METI). The Fukushima Daiichi NPS Mid- and Long-Term Road-map requires investigation of methods to stabilize solid wastes (and to immobilize radioisotopes in the wastes) generated as a result of emergency response and decommissioning activities. Cooling water treatment has resulted in a significant amount of solid and slurry secondary wastes (mostly adsorbents and ion-exchange materials) which will require processing at some point. GeoMelt{sup R} ICV{sup TM} is a joule-heated melter technology which uses a refractory-lined single-use container combining the melter and disposal container. There is no pouring required nor concerns with refractory corrosion which allows the process to accommodate a wide range of waste chemistries and high waste loadings. The testing described here consisted of three engineering-scale melts, each processing between 212 kg and 240 kg of waste simulants, glass formers, and non-radioactive cesium (Cs) and strontium (Sr) tracers. Continuous isokinetic stack sampling of off-gas emissions was performed for each test in order to calculate Cs and Sr retention in the glass wasteform. Single-pass retention of Cs in the final glass wasteform ranged from 91.46 to 99.30%, and single-pass retention of Sr ranged from 99.76 to 100%. Planned particulate recycle will increase these retention levels. Melt 1 processed a mixture of KUR-EH (a zeolite-based ion-exchange material), simulated Advanced Liquid Processing System (ALPS) Carbonate and Iron Slurries, and glass additives. Melt 2 processed a mixture of KUR-EH, KUR-TSG (a titanate-based adsorbent), and glass additives. Melt 3 processed a mixture of KUR-EH, simulated barium sulfate/iron ferrocyanide sludge (AREVA sludge), and glass additives. Waste loadings for these melts ranged from 70 weight percent (wt%) to 82 wt%. Vitrification produces a waste form much denser than the stored water treatment secondary waste wastes, resulting in significant volume reduction. Volume reductions for the three tests ranged from 74 to 79 vol%. Vitrification produces a chemically durable wasteform. Pacific Northwest National Laboratory (PNNL) tested three glass samples from each engineering-scale melt) by the Materials Characterization Center 1 (MCC-1) test, an international standard leach test of the chemical durability of nuclear waste glasses. PNNL also obtained one U.S. reference glass (EA Glass) and two Japan reference glasses P0798) and tested these under the same MCC-1 conditions (90 deg. C, 10 m-1, DIW, and 7, 14, 28-day) as the GeoMelt{sup R} ICV{sup TM} glasses. The GeoMelt{sup R} ICV{sup TM} glasses exhibited lower total normalized releases and 14- to 28-day normalized release rates than the three reference glasses. These results suggest that the GeoMelt{sup R} ICV{sup TM} glasses have durabilities on par with high-level waste glasses under standard test conditions. Post-melt process sampling and analysis indicated no Cs migration into the melter refractory materials and very little deposition of Cs or Sr onto the melter hood or off-gas piping internals. The results of the testing indicated good Cs retention in the glass, high volume reduction and waste loadings, and excellent chemical durability. These factors are important to minimize treatment costs and to protect workers and the environment. (authors)

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

Silicon Carbide Formation Enhanced by In-Situ-Formed Silicon Nitride: An Approach to Capture Thermal Energy of CO-Rich Off-Gas Combustion

Carbothermic smelting of ores to produce metals or alloys in alternating current open/semiclosed and closed submerged arc furnaces, or in closed direct current furnaces, results in large volumes of CO-rich off-gas being generated. Most of the CO-rich off-gas is cleaned and flared on stacks, since the storing of large volumes is problematic due to the associated toxic and explosive risks. Flaring of CO-rich off-gas results in significant wastage of energy. In this study, an alternative method to partially capture the thermal energy associated with off-gas combustion, in the form of silicon carbide (SiC) generated from waste materials (quartz and anthracite fines), is proposed. SiC can partially replace conventional carbonaceous reductants used to produce alloys such as ferrochromium. The influences of quartz and anthracite particle size, treatment temperature, and gaseous atmosphere (nitrogen or air) on SiC formation were investigated. A quartz-anthracite mixture with 90 pct of the particles < 350.9 µm carbothermically treated at 1873.15 K (1600 °C) resulted in almost complete conversion of quartz to SiC in both nitrogen and air atmospheres. The study indicated significant potential for industrial application of the process.

36 MATERIALS SCIENCE↗
DOE OSTI2014

PRELIMINARY ENVIRONMENTAL, HEALTH AND SAFETY RISK ASSESSMENT ON THE INTEGRATION OF A PROCESS UTILIZING LOW-ENERGY SOLVENTS FOR CARBON DIOXIDE CAPTURE ENABLED BY A COMBINATION OF ENZYMES AND VACUUM REGENERATION WITH A SUBCRITICAL PC POWER PLANT

The results of the preliminary environmental, health and safety (EH&S) risk assessment for an enzyme-activated potassium carbonate (K2CO3) solution post-combustion CO2 capture (PCC) plant, integrated with a subcritical pulverized coal (PC) power plant, are presented. The expected emissions during normal steady-state operation have been estimated utilizing models of the PCC plant developed in AspenTech’s AspenPlus® software, bench scale test results from the University of Kentucky, and industrial experience of emission results from a slipstream PCC plant utilizing amine based solvents. A review of all potential emission species and their sources was undertaken that identified two credible emission sources, the absorber off-gas that is vented to atmosphere via a stack and the waste removed from the PCC plant in the centrifuge used to reclaim enzyme and solvent. The conditions and compositions of the emissions were calculated and the potential EH&S effects were considered as well as legislative compliance requirements. Potential mitigation methods for emissions during normal operation have been proposed and solutions to mitigate uncontrolled releases of species have been considered. The potential emissions were found to pose no significant EH&S concerns and were compliant with the Federal legislation reviewed. The limitations in predicting full scale plant performance from bench scale tests have been noted and further work on a larger scale test unit is recommended to reduce the level of uncertainty.

Fitzgerald, David↗
DOE OSTI2015

A Literature Survey to Identify Potentially Volatile Iodine-Bearing Species Present in Off-Gas Streams

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3H, 14C, 85Kr, and 129I. Of these, 129I has the longest half-life and potentially high biological impact. Accordingly, control of the release of 129I is most critical with respect to the regulations for the release of radioactive material in stack emissions. It is estimated that current EPA regulations (EPA 2010) would require any reprocessing plant in the United States to limit 129I release to less than 0.05 Ci/MTIHM for a typical fuel burnup of 55 gigawatt days per metric tonne (GWd/t) (Jubin 2011). The study of inorganic iodide in off-gas systems has been almost exclusively limited to I2 and the focus of organic iodide studies has been CH3I. In this document, we provide the results of an examination of publically available literature that is relevant to the presence and sources of both inorganic and organic iodine-bearing species in reprocessing plants. We especially focus on those that have the potential to be poorly sequestered with traditional capture methodologies. Based on the results of the literature survey and some limited thermodynamic modeling, the inorganic iodine species hypoiodous acid (HOI) and iodine monochloride (ICl) were identified as potentially low-sorbing iodine species that could present in off-gas systems. Organic species of interest included both short chain alkyl iodides such as methyl iodide (CH3I) and longer alkyl iodides up to iodododecane (C10H21I). It was found that fuel dissolution may provide conditions conducive to HOI formation and has been shown to result in volatile long-chain alkyl iodides, though these may not volatilize until later in the reprocessing sequence. Solvent extraction processes were found to be significant sources of various organic iodine-bearing species; formation of these was facilitated by the presence of radiolytic decomposition products resulting from radiolysis of tri-n-butyl phosphate and dodecane. Primarily inorganic iodine compounds were expected from waste management processes, including chlorinated species such as ICl. Critical knowledge gaps that must still be addressed include confirmation of the existence and quantification of low-sorbing species in the off-gas of reprocessing facilities. The contributions from penetrating forms of iodine to the plant DF are largely unknown and highly dependent on the magnitude of their presence. These species are likely to be more difficult to remove and it is likely that their sequestration could be improved through the use of different sorbents, through design modifications of the off-gas capture system, or through chemical conversion prior to iodine abatement that would produce more easily captured forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Investigating realistic anode off-gas combustion in SOFC/ICE hybrid systems: mini review and experimental evaluation

Solid oxide fuel cells (SOFCs) have been deployed in hybrid decentralized energy systems, in which they are directly coupled to internal combustion engines (ICEs). Prior research indicated that the anode tailgas exiting the SOFC stack should be additionally exploited due to its high energy value, with typical ICE operation favoring hybridization due to matching thermodynamic conditions during operation. Consequently, extensive research has been performed, in which engines are positioned downstream the SOFC subsystem, operating in several modes of combustion, with the most prevalent being homogeneous compression ignition (HCCI) and spark ignition (SI). Experiments were performed in a 3-cylinder ICE operating in the latter modus operandi, where the anode tailgas was assimilated by mixing syngas (H 2 : 33.9%, CO: 15.6%, CO 2 : 50.5%) with three different water vapor flowrates in the engine’s intake. While increased vapor content significantly undermined engine performance, brake thermal efficiency (BTE) surpassed 34% in the best case scenario, which outperformed the majority of engines operating under similar operating conditions, as determined from the conducted literature review. Nevertheless, the best performing application was identified operating under HCCI, in which diesel reformates assimilating SOFC anode tailgas, fueled a heavy duty ICE (17:1), and gross indicated thermal efficiency ([Formula: see text]) of 48.8% was achieved, with the same engine exhibiting identical performance when operating in reactivity-controlled compression ignition (RCCI). Overall, emissions in terms of NO x and CO were minimal, especially in SI engines, while unburned hydrocarbons (UHC) were non-existent due to the absence of hydrocarbons in the assessed reformates.

Engineering↗
DOE OSTI2015

Milestone Report - M3FT-15OR03120213 - A Literature Survey to Identify Potentially Problematic Volatile Iodine-Bearing Species Present in Off-Gas Streams

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3 H, 14 C, 85 Kr, and 129 I. Of these, 129 I has the longest half-life and potentially highest biological impact. Accordingly, control of the release of 129 I is most critical with respect to U.S. regulations for the release of radioactive material in stack emissions. Current U.S. Environmental Protection Agency regulation governing nuclear facilities (40 CFR 190) states that the total quantity of radioactive materials entering the general environment from the entire uranium fuel cycle, per gigawatt-year of electrical energy produced by the fuel cycle, must contain less than 5 mCi of 129 I. The study of inorganic iodide in off-gas systems has been almost exclusively limited to I 2 , and the focus of organic iodide studies has been CH 3 I.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Requirements and Conceptual Design of Off-gas Systems for the Reprocessing of Metallic Fuels

An assessment has been conducted to determine how key regulations regarding volatile radionuclide emissions to the atmosphere may apply to the off-gas streams associated with electrochemical reprocessing. The scope of this assessment was based upon a generic electrochemical reprocessing scheme with a throughput rate of 200 MTIHM/y applied to metallic fuel discharged from a sodium fast reactor (SFR), but the findings are able to be translated to other advanced nuclear scenarios as merited. Air dispersion modeling was performed using the EPA CAP-88 model and evaluated the uncontrolled decontamination factors (DFs) that would be required to achieve regulatory compliance with the dose-based limits set forth by EPA regulation 40 CFR 190.10(a). These DFs were compared to those required by fuel cycle–based limits set forth by EPA regulation 40 CFR 190.10(b). Two theoretical sites with disparate climatological conditions were selected for air dispersion modeling (Idaho and Tennessee). The radionuclides modeled included 3 H, 85 Kr, 129 I, and selected alpha-emitting transuranic isotopes (referred to here as 239 Pu-TRU <1y ). It was found that the fuel cycle-based limits in 40 CFR 190.10(b) are most restrictive for 85 Kr and 239 Pu-TRU <1y , with DFs of 3 and 6.1E+09, respectively. The dose-based limit as derived from 40 CFR 190.10(a) could require mitigation of tritium in some scenarios, with an estimated DF of about 3 for the reference scenarios. The fuel cycle-based limit for 129 I resulted in a DF of about 240 for the reference scenario. The need for iodine mitigation based on dose to the public depended upon the physical form of iodine as either particulate or vapor-phase species. Emission of iodine from the facility as a vapor necessitated DFs of about 2 but emission as a particulate would require DFs >6,000 to meet thyroid dose-based limits. Effects of physical form on needed iodine mitigation are significant, but the understanding of speciation of iodine both during electrochemical reprocessing and after release to the atmosphere is limited. The electrochemical processing unit operations were evaluated to identify potential release points for the volatile radionuclides and to assess the potential for retention of the radionuclides within the process (thus decreasing the need for mitigation). Mitigation strategies for 3 H, 85 Kr, 129 I, and 239 Pu-TRU <1y were identified. In all cases, there are reasonably achievable pathways to regulatory compliance, although in some cases additional R&D is merited to verify the chemical speciation of these isotopes and to develop and demonstrate potential treatment technologies for this application. Whether or not additional off-gas controls (beyond common operations such as HEPA filtration and oxygen and moisture control) are needed for any of these regulated or volatile radionuclides depends on the (a) type of facility (NRC-regulated or DOE), (b) used fuel process rate, (c) used fuel burnup and composition, (d) speciation and retention of volatile radionuclides in the process and in the cell gas cleanup system, (e) site-specific parameters such as location, meteorology, stack height, and site boundaries, and (f) levels of conservatism and safety factors used in assessing compliance to air emissions regulations. Performance of this assessment revealed several areas where information is lacking or additional research is required in order to better determine if or what kinds of off-gas control might be needed. First, and most significantly, the understanding of the chemical speciation and physical form and partitioning of iodine during electrochemical processing operations is lacking and prevents the ability to accurately assess the potential iodine mitigation requirements. Future research in this area should be multifaceted and include thermodynamic modeling of iodine speciation in different process steps, experiments to quantify the kinetics of vapor-phase and melt-phase transitions, bench-scale experiments to determine the potential chemical and physical form of iodine emissions from the electrorefining process, and verification of iodine behavior with experiments utilizing operational facilities. Similarly, an improved understanding of iodine behavior in the environment after release from the facility stack will be required to refine dose estimations, as particulate and vapor-phase emissions result in significantly different doses to the MEI.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Fuel Age Impacts on Gaseous Fission Product Capture During Separations

As a result of fuel reprocessing, volatile radionuclides will be released from the facility stack if no processes are put in place to remove them. The radionuclides that are of concern in this document are 3 H, 14 C, 85 Kr, and 129 I. The question we attempt to answer is how efficient must this removal process be for each of these radionuclides? To answer this question, we examine the three regulations that may impact the degree to which these radionuclides must be reduced before process gases can be released from the facility. These regulations are 40 CFR 61 (EPA 2010a), 40 CFR 190(EPA 2010b), and 10 CFR 20 (NRC 2012), and they apply to the total radonuclide release and to the dose to a particular organ – the thyroid. Because these doses can be divided amongst all the radionuclides in different ways and even within the four radionuclides in question, several cases are studied. These cases consider for the four analyzed radionuclides inventories produced for three fuel types—pressurized water reactor uranium oxide (PWR UOX), pressurized water reactor mixed oxide (PWR MOX), and advanced high-temperature gas-cooled reactor (AHTGR)—several burnup values and time out of reactor extending to 200 y. Doses to the maximum exposed individual (MEI) are calculated with the EPA code CAP-88 (Rosnick 2007, 1992). Two dose cases are considered. The first case, perhaps unrealistic, assumes that all of the allowable dose is assigned to the volatile radionuclides. In lieu of this, for the second case a value of 10% of the allowable dose is arbitrarily selected to be assigned to the volatile radionuclides. The required decontamination factors (DFs) are calculated for both of these cases, including the case for the thyroid dose for which 14 C and 129 I are the main contributors. However, for completeness, for one fuel type and burnup, additional cases are provided, allowing 25% and 50% of the allowable dose to be assigned to the volatile radionuclides. Because 3 H and 85 Kr have relatively short half-lives, 12.3 y and 10.7 y, respectively, the dose decreases with the time from when the fuel is removed from the reactor to the time it is processed (herein “fuel age”). One possible strategy for limiting the discharges of these short halflife radionuclides is to allow the fuel to age to take advantage of radioactive decay. Therefore, the doses and required DFs are calculated as a function of fuel age. Here we calculate, given the above constraints and assumptions, the minimum ages for each fuel type that would not require additional effluent controls for the shorter half-life volatile radionuclides based on dose considerations. With respect to 129 I doses, we find that the highest dose is calculated with iodine as a fine particulate. The dose scales as the fraction of the total 129 I that is particulate. Therefore, we assume for all of our calculations that 100% of the 129 I is particulate and allow the user of the results given here to scale our calculated doses to their needs. To summarize the data given in the body and appendices of this report, we find that the principal isotopes of concern are 3 H and 129 I, the latter requiring the highest DFs. The maximum DF value for 129 I is 8000 for the illustrated cases. The required DF for 3 H could be as high as 720, depending on the age of the fuel processed. The DF for 85 Kr could be up to ~60, depending on fuel age. The DF for 14 C is in many cases 1 (no treatment required) but could be as high as 30. The DFs required are within the range of DFs that are reported for the capture technologies that are available for the volatile radionuclides. Achieving the required 129 I and 3 H DFs is more challenging. Variations in stack design and other design factors may also significantly impact the DF requirements.

carbon-14↗
DOE OSTI2021

Performance of a Natural Gas Solid Oxide Fuel Cell System With and Without Carbon Capture

The fuel cell program at the United States Department of Energy (DOE) National Energy Technology Laboratory (NETL) is focused on the development of low-cost, highly efficient, and reliable fossil-fuel-based solid oxide fuel cell (SOFC) power systems that can generate environmentally-friendly electric power with at least 90 percent carbon capture. NETL’s SOFC technology development roadmap is aligned with near-term market opportunities in the distributed generation sector to validate and advance the technology while paving the way for utility-scale natural gas (NG)- and coal-derived synthesis gas-fueled applications via progressively larger system demonstrations. The present study represents a part of a series of system evaluations being carried out at NETL to aid in prioritizing technological advances along research pathways to the realization of utility-scale SOFC systems, a transformational goal of the fuel cell program. In particular, the system performance of utility-scale NG fuel cell (NGFC) systems with and without carbon dioxide (CO2) capture is presented. The NGFC system analyzed features an external auto-thermal reformer (ATR) feeding the fuel to the SOFC system consisting of planar anode-supported SOFC with separated anode and cathode off-gas streams. In systems with CO2 capture, an air separation unit (ASU) is used to provide the oxygen for the ATR and for the combustion of unutilized fuel in the SOFC anode exhaust along with a CO2 purification unit to provide a nearly pure CO2 stream suitable for transport for usage in enhanced oil recovery operations or for storage in underground saline formations. Remaining thermal energy in the exhaust gases is recovered in a bottoming steam Rankine cycle while supplying any process heat requirements. A reduced order model (ROM) developed at the Pacific Northwest National Laboratory (PNNL) is used to predict the SOFC performance. The ROM, while being computationally effective for system studies, provides other detailed information about the state of the stack, such as the internal temperature gradient, generally not available from simple performance models often used to represent the SOFC. Such additional information can be important in system optimization studies to preclude operation under off-design conditions that can adversely impact overall system reliability. The NGFC system performance was analyzed by varying salient system parameters, including the percent of internal (to the SOFC module) NG reformation—ranging from 0 to 100 percent—fuel utilization, and current density. The impact of advances in underlying SOFC technology on electrical performance was also explored.

solid oxide fuel cell (SOFC), natural gas fuel cel↗
DOE OSTI2020

Test for Reconditioning RA Waste with Simulated Bitumen and Concrete in a 1,2 MW Plasma Test Facility - 20092

The operation and maintenance of nuclear power plants, the non nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken in nuclear operation In May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2 mSv/h. Different mixtures of radioactive waste packed in 200 l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissioning fund (KIDSF) administrated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50 kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200 kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 200 l (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 200 l drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 200 l (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describe the test facility, volume reduction factor (VRF) of different waste streams and most important parameters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗