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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding (Milestone Report)

To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Electrochemical behavior of simulated debris from a severe accident using a molten salt system

In a severe nuclear accident, the fuel in the reactor may melt, forming debris, which contains a UO{sub 2}-ZrO{sub 2} stable oxide mixture and parts of the reactor, such as Zircaloy and iron components. Proper handling of the debris is a critically important issue. The debris does not have the same composition as spent fuel, and so it is impossible to apply conventional reprocessing technology directly. In this study, we successfully separated Zr and Fe from simulated debris using NaCl-KCl molten salt electrolysis, and we selectively recovered the Zr and Fe. The simulated debris was made from Zr, Fe, and CeO{sub 2}. The CeO{sub 2} was used for simulating stable UO{sub 2}-ZrO{sub 2}. With this approach, it should be possible to reduce the volume of the debris by recovering metals, which can then be treated as low level radioactive wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1992

Aluminum-fueled rockets for the space transportation system

Aluminum-fueled engines, used to propel orbital transfer vehicles (OTV's), offer benefits to the Space Transportation System (STS) if scrap aluminum can be scavenged at a reasonable cost. Aluminum scavenged from Space Shuttle external tanks could replace propellants hauled from Earth, thus allowing more payloads to be sent to their final destinations at the same Shuttle launch rate. To allow OTV use of aluminum fuel, two new items would be required: a facility to reprocess aluminum from external tanks and an engine for the OTV which could burn aluminum. Design of the orbital transfer vehicle would have to differ substantially from current concepts for it to carry and use the aluminum fuel. The aluminum reprocessing facility would probably have a mass of under 15 metric tons and would probably cost less that $200,000,000. Development of an aluminum-burning engine would no doubt be extremely expensive (1 to 2 billion dollars), but this amount would be adequately repaid by increased STS throughput. Engine production cost is difficult to estimate, but even an extremely high cost (e.g., $250,000,000 per engine) would not significantly increase orbit-raising expenses.

Cutler, Andrew Hall↗
DOE OSTI2013

New approaches for MOX multi-recycling

Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the used assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The Proposed Operations Master Plan of Cigeo - 17013

The proposed operations master plan of Cigeo describes the so-called 'reference progression' of the Cigeo project as envisaged by Andra at the end of the preliminary design studies. It aims at clarifying the objectives of the industrial pilot phase and present the choices offered by the reversibility for the management of the Cigeo project. The reference progression includes the inventory of the waste to be disposed of in Cigeo used for the dimensioning (73 600 m{sup 3} of ILW-LL and 10 100 m{sup 3} of HLW). Cigeo is designed to store the waste already produced and those that will be produced in the future by the existing nuclear installations and those whose creation was authorized, until the predictable end of their operation and their dismantling. All spent fuel is assumed to be reprocessed. The reference progression also includes the duration and the projected milestones of the construction and operation phases of Cigeo until termination. Across the operating life of Cigeo, the extensions of the Cigeo repository areas proceed by successive incremental phases, from the start of the industrial pilot phase (2025) to the final closure operations planned by 2150. Reversibility is defined as the ability to offer to the next generations different choices in terms of radioactive waste long-term management. The implementation of the reversibility principle is based on both governance and technical project management tools. (authors)

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

Mass Inventory Balance and Fuel Breeding Analysis on Transuranic Recycling Option of FBR

Maintaining sustainability of nuclear fuel can be maintained by utilizing new fresh fuel resources, utilization of spent fuel resource, and obtaining fuel breeding resource in the reactors. Utilization of used fuel from recycling resource and fuel breeding resource becomes the key issue which affects to the reprocessing and refueling facilities. Some minor actinides (MA) have been developed to be recycled as well as some different compositions of plutonium for some purposes such as converted fuel for extending longer reactor operation, burning and transmutation of MA, fuel breeding of FBR as well as for proliferation resistance purposes. This evaluation intends to investigate the mass inventory balance focused on plutonium (Pu) and minor actinide (MA) inventory and reactor behavior in relation to reactor criticality, excess reactivity and fuel breeding for several fuel types. Time dependence as well as irradiation process during reactor operation gives the significant impact for changing the behavior of fuel composition which affects to the behavior of reactor itself. Reactor operation time effect on this analysis has been analyzed from the beginning of cycle up to the equilibrium condition of each fuel type compositions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

The formation of crystals in glasses containing rare earth oxides

Korean spent nuclear fuel will reach the capacity of the available temporary storage by 2016. Pyroprocessing and direct disposal seems to be an alternative way to manage and reuse spent nuclear fuel while avoiding the wet reprocessing technology. Pyroprocessing produces several wastes streams, including metals, salts, and rare earths, which must be converted into stabilized form. A suitable form for rare earth immobilization is borosilicate glass. The borosilicate glass form exhibits excellent durability, allows a high waste loading, and is easy to process. In this work, we combined the rare earths waste of composition (in wt%) 39.2Nd{sub 2}O{sub 3}–22.7CeO{sub 2}–11.7La{sub 2}O{sub 3}–10.9PrO{sub 2}–1.3Eu{sub 2}O{sub 3}–1.3Gd{sub 2}O{sub 3}–8.1Sm{sub 2}O{sub 3}–4.8Y{sub 2}O{sub 3} with a baseline glass of composition 60.2SiO{sub 2}–16.0B{sub 2}O{sub 3}–12.6Na{sub 2}O–3.8Al{sub 2}O{sub 3}–5.7CaO–1.7ZrO{sub 2}. Crystallization in waste glasses occurs as the waste loading increases. It may produce complicate glass processing and affect the product quality. To study crystal formation, we initially made glasses containing 5%, 10% and 15% of La{sub 2}O{sub 3} and then glasses with 5%, 10% and 15% of the complete rare earth mix. Samples were heat-treated for 24 hours at temperatures 800°C to 1150°C in 50°C increments. Quenched samples were analyzed using an optical microscope, scanning electron microscope with energy dispersive spectroscopy, and x-ray diffraction. Stillwellite (LaBSiO{sub 5}) and oxyapatite (Ca{sub 2}La{sub 8}Si{sub 6}O{sub 26}) were found in glasses containing La{sub 2}O{sub 3}, while oxyapatite (Ca{sub 2}La{sub 8}Si{sub 6}O{sub 26} and NaNd{sub 9}Si{sub 6}O{sub 26}) precipitated in glasses with additions of mixed rare earths. The liquidus temperature (T{sub L}) of the glasses containing 5%, 10% and 15% La{sub 2}O{sub 3} were 800°C, 959°C and 986°C, respectively; while T{sub L} was 825°C, 1059°C and 1267°C for glasses with 5%, 10% and 15% addition of mixed rare earth oxides. The component coefficients T{sub B2O3}, T{sub SiO2}, T{sub CaO}, and T{sub RE2O3} were also evaluated using a recently published study.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Fiber Optic Sensor for Corrosion Monitoring in Molten Salt Irradiation Experiments

Molten salt reactors (MSRs) are one of the more promising advanced reactor concepts because they offer passive safety, high-temperature process heat, and the potential for improved fuel cycle economics through online reprocessing [1]. However, before MSRs can be reliably deployed, structural materials must be identified that can survive extended exposure to highly-corrosive salts during irradiation at high temperatures. The U.S. Department of Energy (DOE) is funding the design and eventual construction of a Versatile Test Reactor (VTR) that will include experimental facilities for testing under conditions representative of several advanced reactors, including MSRs [2]. These facilities will allow testing of candidate structural materials during irradiation with flowing salt. If properly designed, these experiments will allow for post-irradiation characterization of weight loss in the structural material coupons. However, significantly more information could be gained if the corrosion could be monitored in situ.There are currently no commercially available sensors for in situ corrosion monitoring at temperatures in the range of 500–900°C [3]. An in situ corrosion sensor would also have to survive high-dose neutron irradiation and extended exposure chemically aggressive media (e.g., molten salts). While there are ongoing efforts to use the change in magnetic susceptibility of Cr-containing alloys, this technique is limited to Cr-containing structural materials and there are concerns when using a magnetic susceptibility sensor at temperatures approaching the Curie temperature of iron (770°C) [4].

Petrie, Christian M.↗
DOE OSTI2021

Uranyl oxalate species in high ionic strength environments: stability constants for aqueous and solid uranyl oxalate complexes

Uranyl ion, UO 2 2+ , and its aqueous complexes with organic and inorganic ligands can be the dominant species for uranium transport on the Earth surface or in a nuclear waste disposal system if an oxidizing condition is present. As an important biodegradation product, oxalate, C 2 O 4 2– , is ubiquitous in natural environments and is known for its ability to complex with the uranyl ion. Oxalate can also form solid phases with uranyl ion in certain environments thus limiting uranium migration. Therefore, the determination of stability constants for aqueous and solid uranyl oxalate complexes is important not only to the understanding of uranium mobility in natural environments, but also to the performance assessment of nuclear waste disposal. In this work, we developed a thermodynamic model for the UO 2 2+ –Na + –H + –Cl – –ClO 4 – –C 2 O 4 2– –NO 3 – –H 2 O system to ionic strength up to ~11 mol•kg –1 . We constrained the stability constants for UO 2 C 2 O 4 (aq) and UO 2 (C 2 O 4 ) 2 2– at infinite dilution based on our evaluation of the literature data over a wide range of ionic strengths up to ~11 mol•kg –1 . We also obtained the solubility constants at infinite dilution for solid uranyl oxalates, UO 2 C 2 O 4 •3H 2 O, based on the solubility data over a wide range of ionic strengths. The developed model will enable for the accurate stability assessment of oxalate complexes affecting uranium mobility under a wide range of conditions including those in deep geological repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Installation and Commissioning of the Netzsch Skimmer

The DOE molten salt reactor (MSR) program has been focusing on the thermochemistry and transport properties of fuel, diluent/carrier, coolant, flush, and reprocessing salts proposed for various molten salt reactor technologies. Data from the chemistry project inform models of salt behavior used in dynamic models of nuclear salt reactors, for both high fidelity and systems simulations. Reactor and salt processing systems operate well above salt melting points (>600°C), so salt behavior can be well described by thermodynamics. However, much is not known about salt systems that contain two, three, or even more components. The Netzsch skimmer (commissioned this year) can be used in several configurations to collect data on molten systems. The skimmer uses mass spectrometry coupled with a thermogravimetric analyzer to accurately analyze systems up to 2000°C. Because the mass analyzer is coupled directly to the furnace through two nested orifices, there are no losses due to condensation. As a thermogravimetric analyzer, the skimmer measures weight loss and reaction energies simultaneously with evolved gas analysis. The mass spectrometry identifies gas phase species, and these volatility measurements can be used to derive vapor pressures. The skimmer was purchased in fiscal year 2018, delivered in April, installed in May 2019, and now is fully operational as verified by the analysis of known standards. The skimmer is unique because it will be available for study of salt systems that contain actinides or beryllium, although the additional work control to accommodate these hazards is still under review.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Toward a Fieldable Atomic Mass Spectrometer for Safeguards Applications: Sample Preparation and Ionization

The International Atomic Energy Agency’s (IAEA’s) long-term research and development plan calls for the development of new methods to detect misuse at nuclear fuel cycle facilities such as reprocessing and enrichment plants. At enrichment plants, for example, the IAEA’s contemporary safeguards approaches are based on a combination of routine and random inspections that include collection of UF6 samples from in-process material and selected cylinders for subsequent analyses. These analyses include destructive analysis (DA) in a laboratory (typically by mass spectrometry [MS]) for isotopic characterization, and environmental sampling (ES) for subsequent laboratory elemental and isotopic analysis (also both typically by MS). One area of new method development includes moving this kind of isotope ratio analytical capability for DA and ES activities into the field. Some of the reasons for these developments include timeliness of results, avoidance of hazardous material shipments, and guidance for additional sample collecting. However, this capability does not already exist for several reasons, such as that most lab-based chemical and instrumental methods rely on laboratory infrastructure (highly trained staff, power, space, hazardous material handling, etc.) and require significant amounts of consumables (power, compressed gases, etc.). In addition, there are no currently available, fieldable instruments for atomic or isotope ratio analysis. To address these issues, Pacific Northwest National Laboratory (PNNL) and collaborator, Clemson University, are studying key areas that limit the fieldability of isotope ratio mass spectrometry for atomic ions: sample preparation and ionization, and reducing the physical size of a fieldable mass spectrometer. PNNL is seeking simple and robust techniques that could be effectively used by inspectors who may have no expertise in analytical MS. In this report, we present and describe the preliminary findings for three candidate techniques: matrix-assisted laser desorption/ionization (MALDI) MS, liquid sampling-atmospheric pressure glow discharge (LS-APGD), and laser ablation/ionization (LAI) MS at atmospheric pressure. Potential performance metrics for these techniques will be presented, including detectability, response, isotope ratio accuracy and precision, and ease of use.

Barinaga, Charles J.↗
DOE OSTI2023

Dissolution of Mechanically Mixed UO2-PuO2 and Insoluble Residue Characteristics

During the reprocessing operations of mixed oxide fuel the dissolution step plays a very important role in the attainment of an acceptable plutonium recovery yield. The importance of this step on the overall fuel cycle was recognized by many authors but was especially treated in depth by Lerch [1] and Nicholson [2].

De Regge, P.↗
DOE OSTI2013

The JRC-ITU approach to the safety of advanced nuclear fuel cycles

The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimental facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Combined Assay Systems for the Characterisation of Historical ILW at of the Dounreay Shaft and Silo-18140

In 1955 the Dounreay nuclear power development establishment was created on the northern coast of Scotland primarily to pursue the UK Government policy of developing Fast Breeder Reactor (FBR) technology. The site operated then by the United Kingdom Atomic Energy Authority (UKAEA) saw the development, test and operation of 3 nuclear reactors, two of them FBRs, alongside fabrication and reprocessing facilities for the FBR fuel elements. From the late 1950's until 1977 Intermediate Level radioactive Waste (ILW) generated on site was disposed of into a 65 m deep, 4.6 m diameter shaft. In January 1977, an explosion thought to be caused by hydrogen generated from sodium and potassium wastes reacting with water occurred in the upper part of the shaft, thus stopping any further waste disposal. A second ILW storage facility, the ILW-Silo, started operation in 1971 to deal with the increasing volume of intermediate level waste generated on site in the mid-1960's and continued receiving waste until 1998. The decommissioning of the Shaft and Silo facilities is a challenging engineering task due to the complexity and potential hazards linked with the retrieval and handling of historical waste. Following recovery it is intended to shred the majority of the waste and load it into 200-l assay drums. Sludges would be separated and solidified within a special annulus drum to make them suitable for assay. To support the decommissioning strategies put forward by Dounreay Site Restoration Limited (DSRL), Mirion Technologies Canberra UK Ltd have designed and completed the manufacture of a pair of combined non-destructive waste assay (NDA) systems. The two identical automated ILW counters are intended to provide a means to characterise the waste materials by performing active and passive neutron measurement as well as segmented high resolution gamma spectrometry (HRGS) of 200-l sacrificial waste drums. The active neutron technique is based on the Differential Die Away (DDA) technique using a pulsed Deuterium-Tritium neutron generator to interrogate fissile material present in the waste. The active neutron measurements support, as well as the passive neutron coincidence measurements, a matrix correction capability by use of an Add-a-Source mechanism. The publication will describe in broad lines the approach proposed by DSRL in the processing of the waste retrieved from both the Shaft and Silo facilities. It will also present a description of the overall design of the NDA counters, illustrating how the measurements to be performed will allow DSRL to safely characterise the retrieved waste materials. (authors)

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

An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that states will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗