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DOE OSTI2013

Land and Water Use, CO2 Emissions, and Worker Radiological Exposure Factors for the Nuclear Fuel Cycle

The Department of Energy Office of Nuclear Energy’s Fuel Cycle Technologies program is preparing to evaluate several proposed nuclear fuel cycle options to help guide and prioritize Fuel Cycle Technology research and development. Metrics are being developed to assess performance against nine evaluation criteria that will be used to assess relevant impacts resulting from all phases of the fuel cycle. This report focuses on four specific environmental metrics. • land use • water use • CO2 emissions • radiological Dose to workers Impacts associated with the processes in the front-end of the nuclear fuel cycle, mining through enrichment and deconversion of DUF6 are summarized from FCRD-FCO-2012-000124, Revision 1. Impact estimates are developed within this report for the remaining phases of the nuclear fuel cycle. These phases include fuel fabrication, reactor construction and operations, fuel reprocessing, and storage, transport, and disposal of associated used fuel and radioactive wastes. Impact estimates for each of the phases of the nuclear fuel cycle are given as impact factors normalized per unit process throughput or output. These impact factors can then be re-scaled against the appropriate mass flows to provide estimates for a wide range of potential fuel cycles. A companion report, FCRD-FCO-2013-000213, applies the impact factors to estimate and provide a comparative evaluation of 40 fuel cycles under consideration relative to these four environmental metrics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

ATF material model development and validation for priority fuel concepts

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the Bison fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on the focus was on U 3 Si 2 fuel and FeCrAl cladding as part of a high impact problem through the Nuclear Energy and Advanced Modeling Simulation (NEAMS) program. Then, developments for Cr 2 O 3 -doped UO 2 fuel, and SiC-SiC and Cr-coated zirconium-based claddings began based upon industry interests. In late fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors (CASL) took over further ATF work in Bison in support of the Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC identified their list of priority fuel and cladding concepts, which included Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and Cr-coated zirconium-based and FeCrAl claddings. In particular, the NRC suggested that reports similar in form to NUREG/CR-7024 [1] that was developed for traditional LWR materials UO 2 and zirconium-based claddings (i.e., Zircaloy-4, M5®, ZIRLO) be created for the priority ATF concepts. The approach to ATF capability development in Bison since the beginning has been two-fold: (1) empirical correlations and (2) multiscale model development. Both approaches have uncertainty inherent to them. Uncertainty in empirical correlations is bounded by the experimental data upon which with the correlation was developed. Models developed through a multiscale approach have uncertainty associated with the lower length scale calculations and input parameters that must be propagated to the engineering scale model in Bison. In this report, the recommended models, their range of applicability (e.g., temperature, burnup), and associated uncertainty for the NRC priority fuel concepts Cr 2 O 3 -doped UO 2 and U 3 Si 2 are presented in a manner similar to the aforementioned NUREG. In addition, the Cr 2 O 3 -doped UO 2 models are validated to the Halden Reactor tess IFA-677.1 rods 1 and 5 and IFA-716.1 rod 1. For U 3 Si 2 the models are validated to the recent post-irradiation examination (PIE) data ATF-13 and ATF-15 rods that were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Finally, an uncertainty quantification (UQ) and sensitivity analysis (SA) is completed on a single rod for each fuel concept that takes into account the defined uncertainty in select models. The results of the UQ and SA analyses indicate that given the large uncertainty in some of the input ATF models that the uncertainty on the Bison predictions of fuel performance metrics of interest bound the available experimental data. Correlation coefficients are also reported that identified the uncertain inputs with the strongest correlation (positive or negative) with the outputs of interest. Fuel performance metrics investigated include fuel elongation, rod internal pressure, fuel centerline temperature, and fission gas release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Preliminary Evaluation of Loading DOE Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility

This evaluation looks at the equipment and operations necessary to load United States Department of Energy (DOE) -owned Spent Nuclear Fuel (SNF) into DOE SNF Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The ATR fuel ready for loading would come from fuel storage canisters in the CPP-603 facility, while the Peach Bottom fuel ready for loading could come from either fuel storage canisters in the CPP-603 facility and or containers in the CPP-749 Underground Fuel Storage Facility – Generation I Vaults.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Research Needs for Uranium-Zirconium- Based Metallic Fuels

The purpose of this report is to summarize the needs and present recommendations related to the future direction for U-Zr-based metallic fuel research (including binary U-Zr and ternary U-Pu-Zr alloys). These needs and recommendations were determined by subject matter experts from various institutions during a two-day workshop held at the University of Florida in November, 2019. During open-floor discussions, the highest priority gaps in our understanding of U-Zr-based fuels were down-selected, and near- and long-term needs that directly impact the implementation of these metallic fuels were identified. The identified near-term needs include investigation of the following phenomena: i) swelling and fission gas release, ii) fuel-cladding chemical interaction, iii) phase evolution/constituent redistribution, and iv) thermal properties of the fuels. The long-term needs are: i) investigation of fuel creep and plasticity and ii) fission product (lanthanide) transport. In addition, there was general agreement that all institutions and subject matter experts would benefit from an open-source metallic fuels database with thermophysical property and microstructural data, along with fuel operation/irradiation history, which should be regularly updated with vetted information from new experimental and computational investigations and used to advance metallic fuels research and development. Finally, we recommend that metallic fuel research should be ongoing and that fuel qualification and fuel optimization should be equally prioritized; research combining experiments with modeling and simulation has the largest potential impact.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

AGR-5/6/7 Fuel Specification

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) program is developing and qualifying fuel for a first of a kind high temperature gas-cooled reactor (HTGR) and associated systems to generate electrical power and produce hydrogen or process heat. One of the critical technology development areas of the program is nuclear fuel development and qualification. Fuel qualification is needed to support licensing, operation, and near term deployment of an HTGR reactor.5 Toward this end, the Advanced Gas Reactor (AGR) program includes manufacturing fuels for a series of irradiation experiments in the Advanced Test Reactor (ATR) at INL.5 The fifth, sixth, and seventh irradiation experiments, called “AGR 5/6/7,” will use fuel manufactured entirely in engineering scale equipment to demonstrate the performance of fuels fabricated in a production environment. This fuel will be composited from multiple low enriched uranium carbide/oxide (LEUCO) kernel forming runs, multiple tristructural isotropic (TRISO) coating runs, and multiple overcoating runs. The fuel will be composited into lots as kernels, TRISO particles, and compacts. Previous AGR fuel specifications1,2,8,9 provide the basis for most of the following fuel specifications, which are specific to the AGR 5/6/7 experiment, and may not be immediately applicable to fuel produced for an HTGR reactor. Fuel specifications will be influenced by the final reactor design.

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DOE OSTI2019

Evaluation of Aluminum-clad Spent Nuclear Fuel during Drying and Dry Storage

The DOE SNF Working Group ASNF Subgroup and the U.S. Nuclear Waste Technical Review Board have identified knowledge gaps regarding the management of aluminum-clad spent nuclear fuel (ASNF).[1, 2] The objectives of this research are to observe, sample and analyze the nature of aluminum surface corrosion and corrosion layer chemistry relevant to the ASNF performance over the duration of the dry storage segment of the fuel life cycle. The primary goal is to build an understanding of ASNF behavior, from reactor service and wet storage through drying and dry storage, with parallel testing of surrogate materials. The premise is that residual water affiliated with the aluminum oxide from corrosion during reactor service and wet storage, continues to influence ASNF performance during dry storage, and that even after successful drying the surface conditions of ASNF have yet to be satisfactorily characterized. The initial scope collected and analyzed specimens from irradiated non-fuel components of aluminum-clad Advanced Test Reactor (ATR) fuel elements that are routinely removed at the end of reactor service life. Techniques have been developed for remote material sampling from ATR elements from vented dry storage at the Irradiated Fuel Storage Facility (IFSF). These techniques involve 1) nipping off a small corner of the aluminum near where the end box had been removed and 2) scraping the exterior (non-fueled) plate surface to collect an oxide sample. This second effort targets ATR elements expected to have seen the longest residence in storage, both in the CPP-603 pool (with poor water quality) and in dry storage at IFSF, within the fuel handling scheduled to consolidate ATR at IFSF and receive the remaining inventory at CPP-666. Follow on work has been proposed to nip corners and collect scrapings immediately following ATR fuel drying, before placement in dry storage. Sample analysis will use multiple techniques to understand the chemical composition and morphology of the surface oxides. A simultaneous effort is progressing to develop surrogate materials to enable rigorous controlled experiments. Consistently produced, these surrogate materials can then be used to demonstrate limitations inherent to remote sampling and to validate performance models under relevant constraints. The combined efforts will substantiate ASNF performance and fuel management decisions regarding drying and packaging over a more complete range of life cycle conditions. [1] DOE Spent Nuclear Fuel Working Group, Aluminum-Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage, DOE/ID RPT#: 1575, prepared by Aluminum-Clad SNF Sub Working Group, June 2017. [2] U.S. Nuclear Waste Technical Review Board, Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel, A Report to the United States Congress and the Secretary of Energy, December 2017, p. 143.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

GNF fuel technology update

In 2015, Global Nuclear Fuel (GNF) launched the GNF3 fuel design by inserting lead-use assemblies into two boiling water reactors (BWRs) in the U.S. A lead use assembly is a nuclear fuel assembly which is inserted in a reactor core to confirm its performance. A third set of GNF3 lead-use assemblies are scheduled to be inserted into another reactor in the spring of 2017. GNF3 is a robust next-generation 10*10 fuel design that will improve both fuel cycle and fuel reliability. In addition, GNF3 will be delivered in a fully assembled condition - forgoing the need to insert fuel channels over the bundles at the site. The process of shipping BWR fuel in a channeled configuration is complex and requires an understanding of the potential high-cycle and low-cycle fatigue conditions that can occur during shipping, as characterized by accelerometer data from ship tests. The key component in evaluating channeled fuel shipping is the spacer, and the GNF3 spacer fatigue performance was found to be approximately twice that of GNF2. While fuel reliability has traditionally been related to the avoidance of fuel leakers, for the past 15 years channel - control blade interference has been a close second for concern. To eliminate this problem, GNF3 will come standard with NSF (a zirconium alloy) channels that are effectively resistant to fluence gradient-induced bow and shadow corrosion-induced bow. In addition to the development of GNF3, GNF is evaluating the use of advancing 3D metal printing or additive manufacturing for manufacturing fuel bundle components. Additive manufacture offers the hope of more flexibility in design and shorter manufacturing development cycles. However, since there is little experience with this technology in the nuclear industry, GNF has been evaluating the use of additive manufacture to fabricate small scale components that are easily retrieved from the reactor for testing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Synchrotron Microdiffraction Study of Heavy-Ion-Irradiated Uranium-Molybdenum Dispersion Fuel

There has been an ongoing effort by the Department of Energy Office of Material Management and Minimization to convert research and test reactors from high-enriched to low-enriched uranium fuels. Uranium-molybdenum alloy fuel dispersed in an aluminum matrix (UMo-Al dispersion fuel) has emerged as the leading candidate fuel for converting the remaining high-power research reactors in Europe. One of the primary barriers to using this fuel type is the deleterious reaction of UMo with Al. Depending on the operating temperature of the reactor, this interaction layer can be amorphous under irradiation and allows for the accumulation of fission gas products in large bubbles, leading to unacceptable swelling of the fuel. Therefore, detailed characterization of the phases that form within the fuel and in the interaction layer is critical to understanding, and controlling, UMo-Al dispersion fuel behavior under irradiation. Furthermore, the phase stability of the UMo fuel is of great importance, as the alpha-UMo phase has undesirable swelling compared to the metastable gamma-UMo phase. Heavy ion irradiation was used to reproduce the interaction layer that develops in-reactor. Samples of both the fuel and interaction layer formed under ion irradiation were examined using the high-energy x-ray microdiffraction technique at the Advanced Photon Source (APS) at Argonne National Laboratory to examine the microstructure of the interaction layer. The high-resolution phase information obtained in this experiment provides a valuable basis for understanding the evolution of UMo-Al dispersion fuel under irradiation and as input for computational studies to improve predictive capabilities. Phase information for in-pile fuels can currently be determined from TEM samples (limited sample volume, limiting statistics), or by extrapolation from SEMEDS analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

CAMPUS: Fully Coupled Multiphysics Modeling and Simulation of LWR and FBR Nuclear Fuel Behavior

Nuclear fuels operate in an environment that involves complex multiphysics phenomena, and the multiphysics behavior is often tightly coupled. The prediction of fuel behavior under irradiation involves very complex material behavior that is not seen in any other engineering discipline. Understanding and predicting the evolving material properties and phenomenal behaviors of nuclear fuels are major challenges for fuel design and behavior prediction. The multiphysics behavior is often a tightly coupled thermal, mechanical and chemical phenomenon associated with a large number of control parameters and uncertainties, which requires an advanced fuel performance modeling and simulation capability to improve the existing reactors performance, to address environmental and reactor safety concerns. Nuclear fuel performance modeling and simulation capability is necessary as a result of the high cost and difficulty in experimental measurements. So, a fuel performance code called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed to predict both light water reactor and fast reactor fuel behavior. An effective approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with light water reactor and fast reactor fuels performance analysis. The modelling and simulations approach is of particular importance since it demonstrates highly interrelated and nonlinear fuel performance models can be developed and implemented into a multiphysics platform, and thus more efforts can be put into physics model development and improvement. The modelling results can be used to identify the important phenomena and parameters, to direct and design experiments and to assist in post analysis of expensive nuclear experiments for light water reactor and fast reactor nuclear fuel performance analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Transition to a Thorium-based Fuel Cycle - Choices, Challenges, and Options

The Charter for the Evaluation and Screening (E and S) of Nuclear Fuel Cycle Options requested by the US Department of Energy calls for the identification of promising fuel cycle options to help guide the future directions of research and development in this area. The final report for this study, which was produced by various national laboratory and industry experts, collected the expansive set of fuel cycle options into 40 evaluation groups (EGs). In support of this study, detailed steady-state physics and mass flow data for each of these groups were compiled into a set of Fuel Cycle Data Packages (FCDPs). The data in the FCDPs were then used to quantify the nine criteria detailed in the study to identify the most promising EGs. In addition to supporting the E and S study, data from these FCDPs can be used to benchmark fuel cycle simulation tools for equilibrium calculations. Once benchmarked, their dynamic modeling capabilities can support the continuing effort to determine the most effective approaches for transitioning to proposed advanced future fuel cycles. While steady-state conditions can be calculated directly in some cases, fuel cycle simulation tools allow for the transition scenarios to be simultaneously optimized on multiple parameters such as the fissile material availability, the amount of desirable material in storage, and the time required for transition. In addition to fuel cycle simulations where material availability is considered and is often a constraint to achieving transition, determining the technology needs to enable a transition to take place is also a key consideration (e.g., what technology is needed, when it is needed, and whether it is needed only during the transition and not for the equilibrium fuel cycle). This paper evaluates the material and technology issues and challenges associated with a transition from a uranium-based fuel cycle (such as that operating in the US today, represented by EG01) to a thorium-based fuel cycle, recognizing that thorium requires fissile material to start up. Based on the findings of the E and S report, the two most promising thorium fuel cycle options are considered in this report: (1) EG26, a thermal spectrum molten salt reactor (MSR), and (2) EG28, represented by a sodium fast reactor (SFR, fast spectrum) and a resource-renewable boiling water reactor (RBWR, intermediate spectrum). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Recrystallization on Gas Bubble Swelling in UMo Fuels

The development of low-enriched uranium (LEU) fuels such as UMo dispersion fuels and monolithic UMo fuels is part of a global effort on nuclear nonproliferation. Although UMo fuels have the advantage of high density, excellent irradiation performance and good thermal conductivity, the volumetric swelling in irradiated metallic fuels is known to be an important design parameter because it affects not only the thermal conductivity of the fuels but also the mechanical integrity of fuel structures. Experiments also show that recrystallization dramatically affects the gas bubble swelling kinetics. A series of materials processes such as casting, mechanical rolling and thermal treatment are used in fuel fabrication. Resulting microstructure, such as grain size, porosity, and second phase, strongly depends on the material processes and process parameters. Therefore, a fundamental understanding how fabricated microstructures and irradiation conditions affect the nucleation of gas bubbles, recrystallization and gas bubble swelling kinetics is essential for optimizing material processes to obtain desired microstructure and fuel performance. Great effort has been made in the development of modeling capability for predicting gas bubble swelling kinetics and fission gas release kinetics in nuclear fuels. However, these models are based on the assumptions that the inter- and intra-granular gas bubble sizes are independent of position and only depend on time or fission density. Therefore, they are applicable to a material with uniform microstructure such as equiaxed grains with homogeneous defect distributions, homogeneous thermodynamic and kinetic properties of defects, and uniform radiation conditions. Very recently, we developed a three dimension (3D) microstructure dependent gas bubble swelling model for investigating the effect of heterogeneous microstructures and inhomogeneous thermodynamic properties of defects on gas bubble swelling in polycrystalline UMo fuels. However, the model doesn't take into account the effect of recrystallization on swelling kinetics. In this work, the spatially dependent rate theory is used to describe the evolution of interstitials and interstitial loop density. Recrystallization is described by a phase-field model with a concept of 'phase transition' when the local interstitial loop density is larger than a critical value. The intra- and inter- granular gas bubble evolution is described by the spatially dependent Booth model. A model integrating all the microstructure evolution enables one to study the effect of heterogeneous microstructures, recrystallization, and the metallic fuel thermodynamic and kinetic properties on gas bubble swelling kinetics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Optimization of a dry, mixed nuclear fuel storage array for nuclear criticality safety - 14339

A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different types of research and test reactor fuels. Significant neutronic interaction occurs between fuel canisters within this large storage arrangement of 636 fuel storage canisters. Nuclear criticality safety is maintained primarily by moderation control and through control of neutron interaction and leakage provided by the credited geometry. Over the nearly forty years of operation the collected fuels are now arranged in a configuration that does not minimize the system neutron multiplication factor (k{sub eff}). This creates a concern since the entire remaining facility storage capacity cannot be utilized without exceeding the system minimum subcritical state (k{sub safe}) for ensuring nuclear criticality safety. This work determines a more optimal arrangement for the stored fuels with a goal to minimize the system k{sub eff}, but also with a minimum of potential fuel canister relocation movements. The ultimate goal is to establish a solution for this multiple-objective optimization problem by finding the Pareto Front for maximum k{sub eff} reduction versus the number of canister relocations. The process first required additional analysis of the array by Monte-Carlo neutron transport codes (specifically, MCNP) to gain a greater understanding of the interaction characteristics of the disparate fuel types. Some stored fuels are composed of primarily metallic and oxide components that provide very little inherent moderation, while other fuel types contain graphite or zirconium-hydride which can contribute a significant amount of moderation. These factors cause a complex relationship within a fuel arrangement with respect to thermal utilization and axial leakage probability. The new MCNP analysis was utilized to develop a simplified spreadsheet-based empirical model which simulates the array k{sub eff} result. This empirical model was benchmarked against additional MCNP calculations for various arrangements and then employed in the optimization analysis. A Tabu metaheuristic and stochastic approximation optimization was applied to the empirical model with the optimization function to minimize the array k{sub eff}. The optimization algorithm interrogated the empirical model to determine the optimal rearrangement for up to sixty canister movements. The Pareto Front was found, which shows that a limited number of canister relocation movements could allow the existing thirty-one empty array positions to be filled with common research reactor fuels while also yielding a reduction in the overall array k{sub eff} relative to the current arrangement. This will impart an improvement in the facility utilization while also offering an enhancement in the safety margin. The reduction in array reactivity has been confirmed by a final set of MCNP calculations with the proposed new arrangements. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Advanced Reactors Spent Fuel & Waste Science and Technology Program

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP ) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes three primary tasks the Spent Fuel & Waste Science and Technology (SFWST) Program is executing (or collaborating on) related to the back end of the nuclear fuel cycle (BENFC) for potential future advanced reactors: 1. Advanced Reactors Spent Fuel and Waste Streams: Strategies for the BENFC This set of activities define a high-level strategy for how to systematically approach, identify, and close research and development (R&D) gaps/potential issues associated with long-term management and final disposition of AR SNF and other possible AR waste streams. This task involves summarizing advanced reactor concepts, their likely SNF and other waste forms, and identifying previous experience with similar materials, for example from DOE-managed SNF, with closely related characteristics to the potential future AR SNF. Technical R&D gaps between the breadth of detailed understanding for safe storage, transportation and disposal of the existing light water reactor SNF fuel cycle (e.g., see NASEM, 2022) and potential future fuel cycles based on advanced reactors would then be identified. 2. Characterization and Packaging Options of Advanced Reactor SNF These activities evaluate characteristics and packaging options for advanced reactor spent fuel forms. The fuel forms are categorized into three types: (1) tri-structural isotropic (TRISO), (2) metallic, and (3) fuel salt. Emphasis is given to TRISO and metallic SNF and additional waste streams from such AR as driven by the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations1. Preliminary information for the spent-fuel salt discharged from molten-salt reactors (MSRs) is examined to provide a baseline for future efforts. All calculations and assumptions in this work are based on publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. 3. Back-End Management of Advanced Reactors (BEMAR) The DOE NE-8 organization has defined an Integrated Project Team to evaluate the Back End Management of Advanced Reactors (BEMAR) (includes DOE staff from a range of organizations (e.g., NE-81, NE-82, OCED) and national laboratory technical staff within the DOE NE-81 and NE-82 programs). This BEMAR group works directly with advance reactors developers to assess for the DOE the technical feasibility of storage, transportation, and disposal of AR SNF based on the characteristics provided by the developers to DOE (much of which is proprietary). The BEMAR is also tasked to develop rough-order-of-magnitude cost estimates to compare the waste management system for individual advanced reactors to existing light-water reactor management practices. To accomplish this, the BEMAR group is implementing a Systems Engineering approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

BISON-FIPD integration enhanced low-burnup SFR metallic fuel swelling model evaluation framework

Experiments indicated that metallic fuel in sodium-cooled fast reactors (SFRs) rapidly swells radially and axially at low burnup. Despite that, prior studies have been focused on describing high burnup axial fuel elongation. With recent conventional and non-conventional metallic fuel concepts being considered for license applications, understanding multidimensional fuel swelling at a wide range of burnup levels is important to fuel analysis and qualification activities. Here, we report the development and demonstration efforts of a low-burnup SFR metallic fuel swelling model evaluation framework using the BISON advanced fuel performance code. The framework leverages the Integral Fast Reactor (IFR) program X423 experiment data through the ongoing integration project to enable standardized and automated use of legacy metallic fuel irradiation data maintained in the SFR fuel irradiation and physics database (FIPD) for BISON metallic fuel model verification and validation. In conclusion, the performance of the framework was demonstrated using the two representative metallic fuel swelling model sets implemented in BISON, with a series of insights about future advanced swelling model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗