Safeguards for Reactors and Spent Fuel
It is crucial for safeguards that the amount of plutonium going to the reprocessing plant or repository is what we believe it to be.
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SEARCH · Engineering Papers
Results for “fuel reprocessing”
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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Investigating Uranium Corrosion in Magnesium-containing Sludge Using X-ray Tomography - 20042
One of the current primary concerns within the British nuclear community is the decommissioning of legacy storage ponds in Sellafield Cumbria, where nuclear fuel cladding material (a Mg-Al alloy) resides along with fuel metallic swarf. Corrosion of the immersed metallic elements, which has been occurring over the storage period, has resulted in the formation of what is commonly known as Corroded Magnox Sludge. The behaviour of uranium metal within this particular environment and, more significantly, the identity of the corrosion products that may have formed is currently unknown. The present study attempts to shed some light on these aspects by investigating simplified surrogate systems, mimicking storage ponds. (authors)
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Sulfur Chloride Radiation Chemistry Activities
Presentation highlight the INL Center for Radiation Chemistry Research's contributions the Material Recovery & Waste Form Development Campaign on sulfur chloride processing technologies.
38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
Continuous removal of fission products from molten-salt-fueled reactors
A method based on separation of volatilized molten salt (MS) components, including by mass, is being developed to extract fission products (FPs) from operating molten-salt reactors. The initial application of this method is for accelerator-driven subcritical reactors fueled by (fluorinated molten salt) spent nuclear fuel (UNF) from any past, present, or future reactor. The actinides remain in the subcritical reactor to produce profitable energy and be transmuted while the extracted FPs can be buried without long-lived actinides such that a geologic repository is not necessarily needed to close the nuclear fuel cycle. By removing neutron-absorbing FPs and operating sub-critically, where the restrictive link between operation and criticality is broken, it is possible to envision complete burnup of the UNF fuel. The game-changing feature of continuously processing the molten salt inside the reactor while the reactor operates eliminates the need for a separate reprocessing plant. This feature also simultaneously improves the neutronics of the reactor, increasing the burnup of the fuel and extending its useful life for generating energy. Nuclear nonproliferation and need for geologic repositories are addressed by keeping actinides inside the reactor containment until they are consumed. (authors)
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Fission Gas Measurements in Spent Fuel: Literature Review
The fission-derived noble gases, xenon(Xe) and krypton(Kr), have long been proposed as a tool for supporting nuclear safeguards and declarations verification. A vast body of literature outlines how these non-reactive fission gases released during reactor operation and shutdowns correlate to burnup level, power history, other reactor specifications and reprocessing activities. However, irradiation conditions can vary appreciably both radially and axially within a fuel pin.
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IER 517: Molybdenum Critical Experiment Design [Slides]
In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.
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Refining a Novel Process Monitoring Method to Safeguard Continuously Cycling Designs Using Isotopic Ratios
In advanced reactor (AR) designs, a common feature is continuous chemical processing and circulation of the nuclear material. This work bridges a significant measurement gap in safeguarding reactors with circulating fuel or continuous refueling by leveraging and building on the isotope ratio method first developed by our team under an FY21 Advanced Reactors International Safeguards Engagement (ARISE) project (Uribe et al. 2021). In circulating fuel designs, the radioisotope inventory changes from traditional effects (e.g., radioactive decay, fission) but also includes material transport due to pressure and temperature gradients. Such designs may also require regular or continuous additions or removals during operation, which significantly increases the rate of inventory change compared to traditional pressurized water reactor (PWR)s. Thus, directly tracking the nuclear inventory is ineffective since the isotopes are continuously added and removed. The isotope ratio method instead focuses on detecting changes to the input and output flows of radioisotopes. Previous work showed that for well-chosen pairs of isotopes, the isotopic ratio provides a sensitive and lasting indicator of deviation from normal conditions (e.g., startup, shutdown, diversion). The isotope ratio method is a process monitoring method with potential for application in for forward-looking approaches to International Atomic Energy Agency (IAEA) safeguards. The original process monitoring method was developed for a specific case—the decay tank of a thorium-fueled molten salt breeder reactor. In this expanded work, we explored other types of reactors and processes with nonstationary (e.g., flowing) nuclear material, which are difficult to safeguard with traditional methods because of the transient nature of the systems. The goal of the present work is to generalize the isotope ratio method for use in processes with continuously flowing nuclear material. All continuous processes have an average time for isotopes to be replaced in the system. The isotope ratio method works by choosing radioisotopes with half-lives both above and below the processing time. The present work seeks to explore which isotopes are suitable for the method by simulating the nuclear inventory, radioisotope emissions, and detector responses for several classes of advanced reactors. While the method can in principle be applied to other processes (e.g., enrichment or reprocessing facilities), the present work limits scope to ARs with continuously flowing fuel. Section 2 details the mathematics supporting the isotope ratio method, and Section 3 introduces the representative ARs selected for this work. Section 4 discusses how each reactor was analyzed, and Section 5 showcases the results for each representative reactor. Finally, Section 6 provides concluding remarks and suggests pathways for further analysis.
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Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System
To better enable informed decision making regarding the back end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed SNF. DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified HLW similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.
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Direct Extraction of Uranium Oxides with N,N -Di(2-ethylhexyl)isobutyramide
The direct extraction of uranium from voloxidized nuclear fuel into an organic solvent offers several potential advantages over conventional hydrometallurgical reprocessing, including reducing the reprocessing plant footprint, providing an initial degree of decontamination from fission products, and minimizing the amount of secondary waste from nitric acid. In this work, the direct extraction of uranium oxides into 1.5 M N,N-di(2-ethylhexyl)isobutyramide (DEHiBA) in n-dodecane is examined. UV–vis spectra and distribution ratios of HNO 2 in 1.5 M DEHiBA as well as the equilibrium organic phase H 2 O concentrations in HNO 3 -loaded 1.5 M DEHiBA are also reported. Hypothesized reaction stoichiometries for the direct extraction of uranium from UO 2 , α-U 3 O 8 , and ε-UO 3 are verified through analysis of organic-phase U, HNO 2 , and HNO 3 concentrations after dissolution. Water generated by the dissolution results in the formation of a separate aqueous phase, which will need to be accounted for in future flowsheet design.
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Separated waste stream immobilization of iodine and off-gas caustic scrubber solution (Final Report)
In this work, the approach was to immobilize the CS solution in a glass-bonded composite of cancrinite/sodalite. The synthesis of ceramic waste forms was accomplished by low-temperature (~90°C) processes developed by the team in the framework of a previous NEUP project. By building on our earlier efforts for iodine waste forms, we minimized the schedule and technical risk while still providing novel process demonstrations for the off-gas waste forms. In the case of AgI-based sorbents, a proof-of-concept stripping process was shown to remove iodine from AgI by converting it to NaI(aq) followed by immobilization of these ions in iodosodalite (ISOD), which could be glass-bonded in the same way as for the CS solution cancrinite/sodalite powders.
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Long-length Scintillating Fibers for Nuclear Geological Repositories
This research focuses on the usage of plastic scintillating fibers as ruggedized detectors in long-length sections coupled to long-length standard optical fiber segments for the purpose of radiation detection and monitoring. This technology addresses a need for monitoring radiation in difficult-to-access or inaccessible areas, such as monitoring: radioactive waste within geological repositories, nuclear material storage facilities, spent fuel dry storage areas, inaccessible areas within nuclear facilities including reactors and reprocessing facilities, decommissioning of nuclear facilities, and other similar areas. The fiber can be placed along the tunnels of a geological repository to monitor the radiation field along its length. This can detect an increase of radiation due to movement, leakage, or unexpected opening of the containers, as well as decrease in the radiation due to removal of material. These capabilities have been demonstrated at INL’s Radioactive Scrap Waste Facility by successfully detecting the removal of an interim storage containers from the storage area. The poster focuses on explaining the technology and its capability in monitoring deep geological repositories (DGRs), especially in the domain of safety and international safeguards. As more countries consider the development of DGRs, technologies such as the one being presented can help establish continuity of knowledge of the material, maintain public confidence, and ensure safe operation of the facility. Furthermore, the poster looks at the future steps of the technology and its potential role in the international safeguards regime.
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Impact of Nd ions on the chemical kinetic behavior of radiolytic transients in molten LiCl-KCl-NdCl3 salt mixtures
Pyrochemical reprocessing technologies can be used to recover valuable materials from used nuclear fuel (UNF), such as uranium. However, other critical materials, such as fission product neodymium, are challenging to recover from molten chloride eutectic mixtures (LiCl-KCl) due to the presence of multivalent states (Nd2+/Nd3+), disproportionation reactions, and re-dissolution in the molten chloride salt. Understanding, predicting, and controlling these processes is further complicated by the presence of ionizing radiation fields, for which little is known on the interaction of neodymium ions with the radiolytic transients (es– and Cl2.–) in molten LiCl-KCl salt mixtures. Here, chemical kinetics, from integrated electron pulse irradiation and transient absorption spectroscopy, are presented for the reaction of Nd3+ ions with es– and Cl2.– as a function of temperature (400–600 ?).
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Initial Design of SABRE Fueled Molten Salt Experiment Irradiation Vehicle
Molten Salt Reactors (MSRs) are emerging as promising advanced reactor technologies, utilizing molten salts for both fuel and primary cooling. These reactors may offer advantages such as passive safety, enhanced economic viability, and efficient waste reprocessing while operating at high temperatures and low pressures, which leads to increased system efficiency and reduced mechanical stress on containment structures. However, challenges arise from the complex chemistry and high corrosion rates of fueled molten salts, alongside the volatility of certain fission products during irradiation. The accumulation of these fission products can alter the fuel salt chemistry, affecting corrosion potential and radioactive source terms. Therefore, understanding these phenomena under neutron irradiation is essential for future MSR designs. The SABRE (Salt and Actinide Burnup in a Reactor Environment) experiment, currently in conceptual design at Idaho National Laboratory (INL), aims to investigate these challenges by conducting a drop-in capsule experiment in the Advanced Test Reactor (ATR). The primary objectives include achieving a minimum burnup of 2 GWd/MTU and characterizing fission products while studying corrosion behavior in molten salt systems. This innovative experiment will leverage recent advancements in ATR capabilities, allowing for the safe irradiation of molten salts, thereby advancing the understanding and qualification of MSR technology for future nuclear power generation.
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Assessment of uranium nitride interatomic potentials
Uranium mononitride (UN) is a promising nuclear fuel due to its high fissile density, high thermal conductivity, and suitability for reprocessing. In this study, two uranium nitride interatomic potentials are assessed: Tseplyaev and Starikov's angular-dependent potential and Kocevski et al.'s embedded atom model potential. Predictions of the thermophysical and elastic properties of UN, UN 2 , and α- and β-U 2 N 3 computed using both potentials are assessed and compared to available experimental data. Notably, the Tseplyaev potential performs better with the energetic aspects of UN, e.g., specific heat capacity and point defect formation energies, whereas the Kocevski potential performs better with the structural aspects of UN, e.g., thermal expansion as well as with the elastic properties. The reasons why the Kocevski potential underestimates the UN specific heat are explained by examining the UN phonon properties modeled using both potentials. The Kocevski potential shows better identification of the mechanical stability ranges of UN, UN 2 , and α- and β-U 2 N 3 , reasonably predicting the melting point of UN and predicting stable structures for UN 2 and α- and β-U 2 N 3 . On the other hand, the Tseplyaev potential predicts a premature phase change of both UN and UN 2 and cannot stabilize α- nor β-U 2 N 3 . However, the Kocevski potential cannot predict a stable α-U phase and is thus not suitable for the calculation of formation energies for non-stoichiometric point defects.
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Thermodynamic modeling of KCl-PrCl 3 and KCl-LiCl-PrCl 3 systems
Molten salt electrolysis can recover the actinides from spent nuclear fuels, and it involves a eutectic LiCl-KCl in molten form as an electrolyte. During reprocessing, the concentration of fission products such as La, Nd, Pr, etc., increases and affects the recovery efficiency of the electrolyte. In this work, thermodynamic modeling of KCl-PrCl 3 and KCl-LiCl-PrCl 3 systems was carried out using the CALPHAD (Calculation of Phase Diagrams) approach for the first time. The thermodynamic functions for the pure salts were taken from the SGTE (Scientific Group Thermodata Europe) Substances (SSUB) database. The experimental thermochemical and phase equilibria data available in the literature were used as input for the assessment of KCl-PrCl 3 and KCl-LiCl-PrCl 3 systems. The model parameters for the KCl-LiCl system were adjusted to include the new Gibbs energy descriptions for the pure salts. In addition, the sublattice model for the liquid phase in the LiCl-PrCl 3 system was modified and reassessed to ensure the model compatibility for higher-order extrapolation. There is a good agreement between the experimental and calculated thermochemical and phase diagram data for all the optimized constituent binaries and the ternary system. Furthermore, this work will be beneficial for determining the solubility limit of PrCl 3 in molten LiCl-KCl electrolytes and their thermodynamic properties for improving the efficiency of the pyrochemical process.
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Safeguards Modeling for Advanced Nuclear Facility Design.
Future nuclear fuel cycle facilities will see a significant benefit from considering materials accountancy requirements early in the design process. The Material Protection, Accounting, and Control Technologies (MPACT) working group is demonstrating Safeguards and Security by Design (SSBD) for a notional electrochemical reprocessing facility as part of a 2020 Milestone. The idea behind SSBD is to consider regulatory requirements early in the design process to provide more optimized systems and avoid costly retrofits later in the design process. Safeguards modeling, using single analyst tools, allows the designer to efficiently consider materials accountancy approaches that meet regulatory requirements. However, safeguards modeling also allows the facility designer to go beyond current regulations and work toward accountancy designs with rapid response and lower thresholds for detection of anomalies. This type of modeling enables new safeguards approaches and may inform future regulatory changes. The Separation and Safeguards Performance Model (SSPM) has been used for materials accountancy system design and analysis. This paper steps through the process of designing a Material Control and Accountancy (MC&A) system, presents the baseline system design for an electrochemical reprocessing facility, and provides performance metrics from the modeling analysis. The most critical measurements in the electrochemical facility are the spent fuel input, electrorefiner salt, and U/TRU product output measurements. Finally, material loss scenario analysis found that measurement uncertainties (relative standard deviations) for Pu would need to be at 1% (random and systematic error components) or better in order to meet domestic detection goals or as high as 3% in order to meet international detection goals, based on a 100 metric ton per year plant size.
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Behavior of the Cadmium Pool in the INL Mk-IV Electrorefiner
The Mk-IV electrorefiner has been in service since 1996 recovering uranium from spent fuels from the Experimental Breeder Rector II and Fast Flux Test Facility sodium-cooled fast reactors. The electrorefiner vessel includes a cadmium pool beneath the salt. The voltage signal between the cadmium pool and reference electrode (Ag/AgCl type) provides information about the chemical condition of the cadmium pool with respect to uranium and zirconium saturation. This information is used to guide process control and sampling decisions. This paper describes how the voltage signal is related to chemistry by the analyses of data from Mk-IV operations and confirmatory experimentation.
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Electrochemical Salt Wasteform Development: A Review of Salt Treatment and Immobilization Options
Electrochemical reprocessing, also referred to as pyroprocessing, is a technique for recycling actinides from used nuclear fuel (UNF) to produce fuel for future reactors. Here, UNF is dissolved in a molten salt (e.g., LiCl-KCl eutectic) within an electrorefiner. After UNF dissolution, fission products are released into the electrolyte salt and converted to chlorides. This paper discusses wasteform options for processing the base electrolyte salt with the fission product salts as well as just the rare-earth fission products (as RECl 3 , REOCl, or REO x ) with the intent of finding optimal methods for reducing total waste salt volumes or partitioning the salt for alternate wasteform options. Furthermore, two of the more detailed partitioning options discussed herein include halide removal from the salt (dehalogenation), which accounts for more than half of the salt on a molar basis, and RE fission product removal for wasteforms with high-RE loadings. Wasteform properties are compared with emphasis on wasteform volume starting from a given amount of (1) total salt cations or (2) RE cations. Comparisons are also made of wasteform chemical durabilities, with the data available from like testing methods. A main conclusion from this work is the justification of subsequent salt processing after electrorefiner operations for achieving significant wasteform volume reduction.
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