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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 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

EDF's used fuel management french policy - 15199

Regarding back-end fuel cycle, EDF's strategy, implemented from the very beginning of French nuclear production, consists in reprocessing of used fuel; decision to recycle plutonium in French PWR was taken from 1984. This reprocessing and recycling strategy, relying on existing industrial facilities (AREVA La Hague and Melox plants), has proven to be reliable and efficient. Treatment leads to a stabilization of used fuel total inventory in interim storage. Vitrification process ensures a safe packaging for high level and long-lived waste under a reduced volume. Recycling of fissile materials saves natural uranium requirements. Concentration of Pu in MOX used fuel leaves open the option to reuse it for starting next Generation IV systems beyond 2040. In 2008, EDF and AREVA, supported by French Government, agreed on commercial terms covering long term fuel reprocessing and recycling (up to 2040). Currently, EDF fuel management strategy relies on flow equilibrium in order to respect fuel cycle consistency and sustainability. The underlying principle is to strictly adjust separated plutonium production to recycling possibilities (i.e. licensed reactors and MOX fuel fabrication facilities). Given that fuel cycle policy, any trend in NPP core management that may impact the fuel cycle should be anticipated within times in order to avoid technical or administrative incompatibilities, and guarantee a high level of safety and robustness. In collaboration with industrial operators, EDF have to perform periodic 'Cycle Impact' analyses and report to the French Nuclear Safety Authority (ASN); next exercise is scheduled by the end of 2016. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Adsorption Modeling at the Front and Back Ends of the Nuclear Fuel Cycle

Adsorption is a diverse and complex physicochemical process that has many different applications in chemical, environmental, and nuclear sciences. For example, adsorption can be found at both the front and back ends of the nuclear fuel cycle for (i) capturing uranium from seawater to use as a new source of nuclear fuel and (ii) treatment of the off-gases produced when reprocessing used nuclear fuel. Accurate modeling of these processes is important for the design of efficient capture, recovery, and removal systems for the particular aqueous or gaseous species of interest. A set of algorithms and methods designed for flexibility and portability will greatly improve the ability to rigorously model such a complex set of systems. This approach allows more specific models to be built upon the more generic methods, which gives the overall modeling effort larger breadth and greater usability. Due to the complexity of adsorption processes, it is necessary to develop relatively complex mechanistic models to accurately predict the behavior of engineered adsorption systems. Capture and recovery systems designed for nuclear energy need to have a high level of accuracy to reduce uncertainties and aid in the implementation of separations technologies for efficient nuclear energy systems. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Multinational underground nuclear parks

Newcomer countries expected to develop new nuclear power programs by 2030 are being encouraged by the International Atomic Energy Agency to explore the use of shared facilities for spent fuel storage and geologic disposal. Multinational underground nuclear parks (M-UNPs) are an option for sharing such facilities. Newcomer countries with suitable bedrock conditions could volunteer to host M-UNPs. M-UNPs would include back-end fuel cycle facilities, in open or closed fuel cycle configurations, with sufficient capacity to enable M-UNP host countries to provide for-fee waste management services to partner countries, and to manage waste from the M-UNP power reactors. M-UNP potential advantages include: the option for decades of spent fuel storage; fuel-cycle policy flexibility; increased proliferation resistance; high margin of physical security against attack; and high margin of containment capability in the event of beyond-design-basis accidents, thereby reducing the risk of Fukushima-like radiological contamination of surface lands. A hypothetical M-UNP in crystalline rock with facilities for small modular reactors, spent fuel storage, reprocessing, and geologic disposal is described using a room-and-pillar reference-design cavern. Underground construction cost is judged tractable through use of modern excavation technology and careful site selection. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Clement, J. D.↗
DOE OSTI2020

Evaluation of Material Balance Approaches for Hanford Direct-Feed Low Activity Waste Processing - 20022

The Hanford Site has accumulated millions of gallons of tank waste from reprocessing spent fuel to recover plutonium, uranium, cesium, and strontium. The supernatant from the accumulated tank waste will be treated using a Direct Feed Low-Activity Waste approach. The supernatant will be treated to remove solids and cesium in the Tank-Side Cesium Removal process in the Hanford tank farm, then vitrified in a semi-batch process in the Hanford Waste Treatment and Immobilization Plant (WTP). Currently, each batch of feed is sampled at three locations prior to being fed to the melter: the feed qualification tank in the Hanford tank farm as well as the concentrate receipt vessel (CRV) and melter feed preparation vessel (MFPV) in the WTP. The feed qualification sample is taken from a large batch of accumulated feed, only two to three samples are expected each year. Approximately 275 samples from the CRVs and 1100 samples from the MFPV are expected each year. An evaluation was performed to determine if a material balance based on the feed qualification sample could replace most of the sampling in the CRVs and MFPVs. The evaluation consisted of three elements: (1) determination of the practicality of using a material balance to estimate the stream composition of the CRV and MFPV contents, (2) evaluation of whether the material balance could be automated using the existing process control system, and (3) determination of the uncertainty in glass composition using the material balance approach. It is assumed that periodic sampling at the CRV and MFPV would be performed periodically to re-baseline the material balance, evaluations are in progress to determine the frequency of this periodic sampling. Process sample locations downstream of the melter were reviewed as well, but the partitioning of semi-volatile species in the melter was determined to preclude extending the material balance approach past the melter. It was determined that replacement of the CRV sample location was feasible and did not increase process uncertainty or significantly impact waste loading. Replacement of the MFPV sample was also determined to be feasible, but that measurement of the glass former chemical addition may be needed prior to addition of these chemicals to the MFPV. This measurement could be performed by an in situ laser-induced breakdown spectroscopy (LIBS) system. Limited tests were performed to evaluate LIBS for direct measurements of the low-activity waste melter feed. The use of a material balance would eliminate over 1200 samples each year if only 10% of the CRV and MFPV batches are sampled and could likely allow the WTP laboratory to operate on days only versus 24/7 operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Influence of elevated temperature and oxygen on the capture of radioactive iodine by silver functionalized silica aerogel

Reprocessing is considered a competent strategy for spent nuclear fuel management, yet radioiodine ( 129 I) is emitted in reprocessing off-gas as a hazardous byproduct. Silver functionalized silica aerogel (Ag 0 -aerogel), a promising iodine capture material, experiences a reduction in its capacity after prolonged exposure to off-gas components at elevated temperatures, a phenomenon termed as aging. To fully understand this process, we isolated the contribution of each aging factor, exposing Ag 0 -aerogel samples to N 2 and dry air gas streams, respectively, at 150 °C for different time periods. Aged samples were loaded with I 2 to examine the capacity change and comprehensively characterized to investigate the evolution of their properties. Results show that temperature alone did not alter Ag 0 -aerogel's capacity but triggered Ag 0 nanoparticles sintering and generated organic sulfur species. The presence of O 2 reduced the capacity by ~20 %, causing (i) formation of silver sulfide (Ag 2 S) crystals and (ii) oxidation of Ag-thiolate (Ag-S-r) to Ag sulfonate (Ag-SO 3 -r). Given that Ag 2 S readily adsorbs I 2 , the formation of Ag-SO 3 -r is the major inhibitor for iodine adsorption. This hypothesis was supported by density functional theory (DFT) simulations. These findings unraveled key mechanisms of Ag 0 -aerogel aging, which are useful in the development of materials that withstand realistic spent-nuclear-fuel-reprocessing off-gas conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

A Systems Processing Model for Molten Salt Reactors

The chemistry group within the U.S. Department of Energy Molten Salt Reactor Campaign is developing a model for the molten salt reactor system for assessing the suitability of various used fuel processing schemes. In the long run, the developed model is expected to be a tool for better understanding the interplay between the components of the molten salt reactor system including uranium resources, the molten salt reactor, the used fuel processor (e.g., reprocessing plant), and waste management. In this regard, the report proposes a theoretical framework for integrating the various components of the molten salt reactor system. Rather than being buried in implementation details for seemingly heterogeneous components of the reactor system asking for various modeling techniques, this report proposes to approximate and integrate all components with a unified mathematical framework, a discrete-time linear system. An abstracted molten salt reactor system is given as an example to illustrate the effectiveness of the proposed theoretical framework. A fully developed molten salt reactor system model (based on the proposed framework) is expected to be beneficial for various objectives including the transient and equilibrium analysis of fuel cycle scenarios and the identification of potential technology bottlenecks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Color Sensor Comparison for Solvent Extraction Processes

Idaho National Laboratory is building Beartooth, a test bed to advance research into nuclear fuel cycle reprocessing. This test bed will give researchers the opportunity to study solvent extraction processes. Beartooth extraction operations will include the use of centrifugal contactors designed for the recovery and purification of special nuclear material from spent fuel. This new test bed facility provides researchers with the opportunity to study innovative technologies. As part of that initiative, this project will integrate non-traditional (atypical to a solvent extraction process) measurement sensors into a system of contactors to determine the state-of-health. These non-traditional sensors have the potential to enhance nuclear safeguarding efforts and other process activities. The non-traditional sensors include accelerometers, acoustic microphones, colorimetric, pH, conductivity, viscosity, density, and infrared cameras. This report documents the testing and evaluation of two different colorimetric RGB sensors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Improving actinides recycling, a key step towards a more sustainable nuclear energy - 15617

Nuclear energy is thought to be one of the energy sources that could help mitigating the global climate change together with the renewables, due to its low green-house-gases emissions, its reliability and its high load power. However, nuclear energy will only develop if future nuclear energy systems meet the criteria of sustainability in terms of durability, bearability and liveability. Nowadays, many countries chose the so-called once-through cycle which basically considers spent nuclear fuel as a waste, whereas others reprocess their spent fuel to recover the energetically-valuable material Pu (and partially U) to produce Mixed Oxide Fuel (MOX) to be irradiated in a second cycle (twice-through cycle). Both options can be considered relevant, fitting diverse criteria; but if we consider a long lasting prospective for nuclear options, the once-through cycle sustainability is clearly lower (low consumption of natural resource, no recycling of potentially valuable energetic material, larger volume of waste...). Reversely, twice-through cycle can be seen as a first step towards an improved sustainability with a partial recycling of U/Pu and the reduction of the waste volume and toxicity. However, it is likely not sufficient for the long-term and implementing more efficient recycling is a key-driver to improve the environmental friendship, the economics and the social acceptability of nuclear energy. Future nuclear fuel cycles will have to better preserve the natural resource by improving the actinides recycling, to minimize the volume and toxicity of nuclear waste, to ensure their long-term confinement and to prevent any proliferation-risk. Such a gradual and progressive process calls for implementing appropriate technologies, either for reactors but also for fuel treatment/recycling facilities. This paper will describe the successive step that would allow improving the overall sustainability. Indeed, recycled Pu can be once recycled in MOX fuel feeding similar PWRs (Pressurized Water Reactor). Together with the recycling of reprocessed uranium, this first step already allows preserving ∼17% of the natural U resource and to confine ultimate wastes without Pu in a tailored durable wasteform. Subsequent Pu multi-recycling requires introducing Fast Reactors (FR) to efficiently use U-238 available in natural and depleted uranium stockpiles. In parallel, recycling processes would have to be adapted for treating spent LWR and SFR MOX spent fuels, which is currently under study thanks to an ambitious R and D program dedicated on MOX treatment and recycling. For the longer term, minor actinides recycling would allow decreasing the waste burden towards future generations: (i) Am-recycling leads to decrease the waste residual heat-power and increase the repository density and lifespan, (ii) Cm-recycling leads to decrease waste long-term toxicity. The stepwise implementation of these different processes defines a path towards sustainability for nuclear energy that are described in this paper from the fuel cycle viewpoint. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of an MC&A toolbox for liquid-fueled molten salt reactors with online reprocessing (Final Report)

A critical barrier to the deployment of MSRs is the absence of a well-defined nuclear material control and accounting (MC&A) approach, a vital prerequisite to meet NRC licensing requirements as well as facilitating future international exports. Liquid-fueled MSR variants (especially those incorporating online refueling or reprocessing) present a unique set of challenges to traditional nuclear material control and accountancy. Unlike solid-fueled light-water reactors, traditional item-counting methods cannot be applied as an accountancy strategy. Rather, MC&A approaches to MSR variants (including both uranium and thorium fueled designs) are more analogous to bulk material handling facilities (e.g., enrichment and reprocessing); yet further complicating matters is the fact that fuel medium is also highly radioactive. Meanwhile, the space of MSRs covers a broad range of design parameters, including thermal and fast spectra designs; operation in actinide breeder or burner modes, choice of actinide fuel used (e.g., 235 U, 232 Th / 233 U, denatured 233 U, etc.), pool or loop-type configuration, and even different salt chemistry. Each of these design choices introduce significant challenges to MC&A approaches within MSR facilities. We propose to bridge this gap for liquid-fueled MSRs by developing a modular, component-based test framework for evaluating viable process monitoring and MC&A techniques specifically suited to liquid-fueled MSR system variants employing online refueling or reprocessing. This test platform will consist of a toolbox of independent process modules representing discrete physical units (such as the reactor core, off-gas processing, decay tanks, and actinide separation units), each with its own self-contained physics responsive to the input mass flow, along with appropriate measurement models that can be coupled to key flow points. These dynamic physical signatures thus afford the ability to test the viability and efficacy of potential accountancy techniques under the full range of reactor operating conditions. As process modules are connected via mass flows, the result is a reconfigurable, generic MSR mass flow model capable of serving as an MC&A test platform for a broad spectrum of possible MSR configurations. The resulting MSR MC&A toolbox will enable robust assessment of accountancy strategies for this unique facility type, including analysis of physical feedbacks arising both from depletion of the fuel over time as well as from potential off-normal events, including those introduced by equipment failures (e.g., a pump failure) as well as by malicious action (i.e., attempts to divert material). The proposed toolbox both addresses a critical needs area for the MPACT analysis toolkit while leveraging existing MPACT-sponsored tools, especially with respect to simulation of measurement and accountancy techniques for advanced fuel cycle facilities. Beyond enabling new analysis capabilities for MSR systems, the design of this toolbox will be to such to maximize compatibility with existing MPACT tools, such to enhance existing facility MC&A analysis capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Electrochemically Modulated Separation for Plutonium Safeguards

Accurate and timely analysis of plutonium in spent nuclear fuel is critical in nuclear safeguards for detection of both protracted and rapid plutonium diversions. Gamma spectroscopy is a viable method for accurate and timely measurements of plutonium provided that the plutonium is well separated from the interfering fission and activation products present in spent nuclear fuel. Electrochemically modulated separation (EMS) is a method that has been used successfully to isolate picogram amounts of Pu from nitric acid matrices. With EMS, Pu adsorption may be turned "on" and "off" depending on the applied voltage, allowing for collection and stripping of Pu without the addition of chemical reagents. In this work, we have scaled up the EMS process to isolate microgram quantities of Pu from matrices encountered in spent nuclear fuel during reprocessing. Several challenges have been addressed including surface area limitations, radiolysis effects, electrochemical cell performance stability, and chemical interferences. After these challenges were resolved, 6 µg Pu was deposited in the electrochemical cell with approximately an 800-fold reduction of fission and activation product levels from a spent nuclear fuel sample. Modeling showed that these levels of Pu collection and interference reduction may not be sufficient for Pu detection by gamma spectroscopy. The main remaining challenges are to achieve a more complete Pu isolation and to deposit larger quantities of Pu for successful gamma analysis of Pu. If gamma analyses of Pu are successful, EMS will allow for accurate and timely on-site analysis for enhanced Pu safeguards.

Pratt, Sandra H.↗
DOE OSTI2021

Acid Dissolution Behavior of Ferritic FeCrAl Tubes Candidates for Nuclear Fuel Cladding

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance to high-temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT and C26M) in three acids (H 2 SO 4 , HNO 3 , and HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS and 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Tritium Distribution and Capture Requirements for UNF Reprocessing

Four radionuclides have been identified as being sufficiently volatile in reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements. These radionuclides are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. Previous studies have discussed the need for treating off-gas streams to control {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I from the reprocessing of used nuclear fuel (UNF) and the regulations that set the limits for releases from such a facility. These regulations ultimately prescribe the decontamination factor (DF) that must be met in the facility. Jubin et al. examined the role of off-gas pathways since the gaseous radionuclides can partition to different reprocessing off-gas streams. Based on this report and subsequent analysis, the required DF for {sup 129}I could be 1000 to 3000. The required DF for {sup 3}H could be as high as 200, depending on the age of the fuel processed. The DF for {sup 85}Kr could be up to ∼35, also depending on fuel age. The required DF for {sup 14}C is 1 in many cases (no treatment required) but could be higher in specific cases. While iodine has been the focus of a number of studies due to the high DF requirements on multiple escape pathways, it is clear from the studies noted above that the control of tritium will also be a significant issue for future reprocessing plants, especially if processing relatively short-cooled fuels (<10 yr post reactor discharge). Tritium contained in a UNF matrix will be released into the liquid and off-gas streams during fuel reprocessing. The amount released into the process streams is a function of the type and burnup of the fuel processed, and the chemical processes employed within the reprocessing plant. Previous studies have examined the tritium capture requirements for the reprocessing facility assuming that all of the tritium is volatilized in the head-end operations, i.e., tritium pretreatment is employed and tritium is completely released from the cladding. This is clearly a worst case scenario from the standpoint of the amount of tritium that could be released, but the scenario tends to limit the release pathway to the head-end off-gas stream. In this analysis, the volume of HTO and the HTO to H{sub 2}O ratio for each of the three plant-wide tritium DFs were determined. The overall amount of HTO that had to be captured from the off-gas streams was virtually the same for the three DF cases evaluated; 98%, 99% and 99.5% of the tritium in the fuel entering the plant needed to be recovered for overall plant DFs of 50, 100 and 200, respectively. The use of the 'tritium barrier' to limit the spread of HTO from the head-end systems to the balance of the plant results in an increase in tritiated waste concentration in the head-end recycle acid systems which results in a decrease in the total amount of HTO + H{sub 2}O that must recovered from the gaseous effluent from the plant. The use of the 'tritium barrier' decreases the waste volume due to the higher HTO:H{sub 2}O ratio in the treated off-gas stream. The results also show that increasing the acid recycle percentage increases the concentration of HTO within the UNF reprocessing plant. This is then manifested in less HTO + H{sub 2}O that must be recovered from the gaseous effluents. Finally, the application of tritium recovery on the recycle acid reduces the HTO concentration in the recycle acid stream and generates a separate HTO + H{sub 2}O stream that must be managed. However, the resulting volume of tritium-containing waste generated plant-wide is a function of the selectivity of the tritium recovery system and the ability of the tritium recovery system to limit the accumulation of HTO in the recycle acid such that the HTO + H{sub 2}O recovered from gaseous effluents does not contribute significantly to the overall tritium waste volume. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Thermally and chemically responsive nanoporous materials for efficient capture of fission product gases.

The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel. We achieved synthesis, characterization and detailed modeling of the materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.

36 MATERIALS SCIENCE↗