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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Advancements in FeCrAl Alloys for Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors

Advanced FeCrAl alloys are being considered as an enhanced accident tolerant fuel (ATF) cladding for light water reactor (LWR) applications as they can exhibit slower reaction kinetics with high temperature steam and have a reduced production of hydrogen during accident scenarios compared both to historic 304L cladding and Zircaloy. Both of these attributes are key towards producing a fuel-form with enhanced accident tolerance over the current Zr-based-UO{sub 2} fuel form in deployment. A multi-year, multi-laboratory effort has been established to rapidly advance FeCrAl alloys as an ATF concept. This effort has focused on optimizing the microstructure and composition of these alloys in an attempt to provide performance on-par with or better than the current Zr-based cladding in both normal operating conditions and under accident scenarios. To date, a significant database supporting key performance factors in both scenarios has been developed. Based on this database, a compositional and microstructural window has been identified providing the insight for an optimized, advanced nuclear-grade FeCrAl alloy for LWRs. This work will provide a comprehensive overview of the systematic effort to advance FeCrAl alloys as an enhanced ATF cladding in commercial LWRs. Efforts undertaken by this multi-year, multi-laboratory program have shown the viability of using FeCrAl alloys as an ATF cladding in LWRs. Most importantly, it has been shown the oxidation resistance, decreased hydrogen production, and reduced heat generation of FeCrAl alloys strongly contribute to its ATF capability. Furthermore, it has been shown that the normal operational performance of a FeCrAl cladding can rival that of Zr-based alloys. Finally, it was found that optimizing the performance of FeCrAl cladding materials requires balancing the starting composition and microstructure in order to satisfy key performance indicators such as oxidation resistance and radiation tolerance. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Tritium Content in and Release from Pressurized-Water-Reactor Fuel Cladding

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels; >30 yrs). 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. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritiated water (THO or T{sub 2}O). In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy'R cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste attributed to Zircaloy'R clad fuel by recovering the zirconium from the cladding for reuse. Treatment options for Zircaloy cladding include recycling to recover the significant value of the zirconium and eliminate costs associated with the disposal of transuranic-contaminated Zircaloy. If LWR fuel is reprocessed and the fuel is dissolved without tritium pretreatment, the tritium in the cladding will remain bound to the cladding. If tritium pretreatment is included in the flow sheet, some portion of the tritium in the cladding may be released along with the tritium from the fuel matrix. To design and operate reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the UNF must be understood and quantified. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0-96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. The present study was undertaken to understand how tritium pretreatment at standard tritium pretreatment conditions (oxidation in air at 480 to 600 deg. C) affects the tritium content in the Zircaloy cladding and the extent to which the tritium content could be reduced with modest increases in the tritium pretreatment temperature. The data indicate that <20% of the tritium in PWR cladding will be removed by pretreatment at 500-700 deg. C for 24 h. Essentially 100% of the tritium will likely be released if the heating time at 700 deg. C is extended to 96 h. Small amounts of tritium are expected to be released if the cladding is heated at ≤ 600 deg. C for up to 96 h. The results of this study indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the pretreatment system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in any subsequent processing of the cladding to recover/recycle the zirconium. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Advanced Fuels Campaign FY 2014 Accomplishments Report

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of a “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. AFC uses a “goal-oriented, science-based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. The modeling and simulation activities for fuel performance are carried out under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which is closely coordinated with AFC. In this report, the word “fuel” is used generically to include fuels, targets, and their associated cladding materials. R&D of light water reactor (LWR) fuels with enhanced accident tolerance is also conducted by AFC. These fuel systems are designed to achieve significantly higher fuel and plant performance to allow operation to significantly higher burnup, and to provide enhanced safety during design basis and beyond design basis accident conditions. The overarching goal is to develop advanced nuclear fuels and materials that are robust, have high performance capability, and are more tolerant to accident conditions than traditional fuel systems. AFC management and integration activities included continued support for international collaborations, primarily with France, Japan, the European Union, Republic of Korea, and China, as well as various working group and expert group activities in the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD-NEA) and the International Atomic Energy Agency (IAEA). Three industry-led Funding Opportunity Announcements (FOAs) and two university-led Integrated Research Projects (IRPs), funded in 2013, made significant progress in fuels and materials development. All are closely integrated with AFC and Accident Tolerant Fuels (ATF) research. Accomplishments made during fiscal year (FY) 2014 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the lead technical contact is provided for each section.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Texture Evolution in U-10Mo Nuclear Fuel Foils during Plasma Spray Coating with Zr

A uranium-molybdenum alloy clad in 6061 aluminum has the potential to lead to a wide application of low-enriched uranium fuels, replacing highly enriched uranium for research reactors. A Zr coating acts as a diffusion barrier between the fuel and the aluminum cladding. In this study, U-10Mo (mass %) was coated with Zr using a plasma spray technique recognized as a fast and economical coating method. Neutron time-of-flight diffraction was used to study the microstructure evolution by quantifying the phase fractions of involved phases as well as the texture evolution of U-10Mo and Zr during plasma spray coating with Zr. Quantitative texture analysis revealed that the texture was drastically changed for high coating temperatures, likely due to selective grain growth. Furthermore, the Zr coating showed a preferential orientation, which could be correlated with the initial texture of the uncoated U-10Mo. This could be explained by the epitaxial growth of the Zr on the U-10Mo substrate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

UFD Proof-of-Concept Demonstration

The BISON Spent Nuclear Fuel modeling capability has been demonstrated through simulations of the complete nuclear fuel lifecycle including in-reactor irradiation, spent fuel pool cooling, vacuum drying and finally dry cask storage. The importance of hydrogen in the cladding and the formation of hydrides at various stages of the fuel lifecycle have been a focus of the initial development work. In addition, the integration in BISON of a hydride damage model has improved the prediction of the cladding mechanical behavior. The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-17IN0201012 on UFD Proof-of-Concept Demonstration for the INL Engineering-Scale Fuel Performance (BISON) Workpackage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Competing effects of electronic and nuclear energy loss on microstructural evolution in ionic-covalent materials

Ever increasing energy needs have raised the demands for advanced fuels and cladding materials that withstand the extreme radiation environments with improved accident tolerance over a long period of time. Ceria (CeO2) is a well known ionic conductor that is isostructural with urania and plutonia-based nuclear fuels. In the context of nuclear fuels, immobilization and transmutation of actinides, CeO2 is a model system for radiation effect studies. Covalent silicon carbide (SiC) is a candidate for use as structural material in fusion, cladding material for fission reactors, and an inert matrix for the transmutation of plutonium and other radioactive actinides. Understanding microstructural change of these ionic-covalent materials to irradiation is important for advanced nuclear energy systems. While displacements from nuclear energy loss may be the primary contribution to damage accumulation in a crystalline matrix and a driving force for the grain boundary evolution in nanostructured materials, local non-equilibrium disorder and excitation through electronic energy loss may, however, produce additional damage or anneal pre-existing defect. At intermediate transit energies where electronic and nuclear energy losses are both significant, synergistic, additive or competitive processes may evolve that affect the dynamic response of materials to irradiation. The response of crystalline and nanostructured CeO2 and SiC to ion irradiation are studied under different nuclear and electronic stopping powers to describe some general material response in this transit energy regime. Although fast radiation-induced grain growth in CeO2 is evident with no phase transformation, different fluence and dose dependence on the growth rate is observed under Si and Au irradiations. While grain shrinkage and amorphization are observed in the nano-engineered 3C SiC with a high-density of stacking faults embedded in nanosize columnar grains, significantly enhanced radiation resistance is attributed to stacking faults that promote efficient point defect annihilation. Moreover, competing effects of electronic and nuclear energy loss on the damage accumulation and annihilation are observed in crystalline 4H-SiC. Systematic experiments and simulation effort are needed to understand the competitive or synergistic effects.

Zhang, Yanwen↗
DOE OSTI2017

Containment Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodology for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

NEAMS Workbench and Bison Fuel Performance Remote Application Configuration

The mission of the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program is to develop, apply, deploy, and support state-of-the-art predictive modeling and simulation tools for the design and analysis of current and future nuclear energy systems. This is accomplished using computing architectures that range from laptops to leadership class facilities. The NEAMS Workbench [1,2,3] initiative will facilitate the transition from conventional computer codes to high-fidelity ones by providing a common analysis environment and interface for model creation, review, execution, output review, and visualization for integrated codes. The BISON fuel performance code [4] is a finite element method (FEM)-based software being developed by multiple organizations led by the Idaho National Laboratory for the US Department of Energy Office of Nuclear Energy. BISON can simulate a variety of fuel and cladding compositions and geometries in varying dimensionalities, depending upon the problem type and complexity, including one-dimensional (1D), 1.5D, 2D axially symmetric (2D-rz), 2D radial-azimuthal (2D-rθ), or 3D models. The NEAMS Workbench allows execution of the BISON fuel performance code on local and remote compute resources. In this report, the detailed instructions of configuring the NEAMS Workbench for remote execution of the BISON fuel performance code on a portable batch system (PBS)-scheduled compute resource, as well the transfer of associated Bison files, is illustrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Integrated analytical method to demonstrate 10 CFR 50.46c compliance

Large break loss-of-coolant accident (LOCA) evaluation models (EMs) have historically been focused on fresh fuel at or near beginning-of-life in water cooled reactors. However, these evaluation models did not account for fuel thermal conductivity degradation (TCD). With the explicit modeling of fuel TCD, it is possible that the limiting LOCA conditions could occur for higher burnup fuel. NRC-sponsored research conducted by ANL as part of the 10 CFR 50.46c rulemaking identified mechanisms for cladding embrittlement that exist in higher burnup fuel which were not previously well understood. Additionally, the explicit consideration of fuel pellet thermal conductivity degradation results in more severe fuel rod / assembly considerations for higher burnup fuel in the large break LOCA scenario. In aggregate, these embrittlement mechanisms and phenomena necessitate the consideration of all fuel in the core which undergoes a postulated LOCA, as the limiting fuel cannot be determined a priori for each of the analysis acceptance criteria. The Full Spectrum LOCA EM accounts for all assemblies in the core at the time of a postulated LOCA by defining representative hot assemblies which combine the limiting elements of the fuel assemblies which they represent. By doing so, it is ensured that the LOCA analysis result represents the more limiting assembly in the core for each criterion. Changes in the LOCA boundary conditions (fuel rod initialization, axial power distributions, etc.) that occur during a cycle of operation are also appropriately accounted for in the FSLOCA EM by integrating fuel rod design and nuclear design methods into the LOCA EM. As such, the FSLOCA EM is well-positioned to show compliance with the 10 CFR 50.46c rulemaking as the rule package is nearing completion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Current status of post-irradiation examination for the ATF-1 irradiation

A series of irradiations, referred to as the ATF series, is currently underway to evaluate the feasibility of different Accident Tolerant Fuels for commercial light water reactors. For initial screening, concepts consisting of an uranium bearing phase and cladding were irradiated in stainless steel capsules. These tests provide screening data among concepts to ensure viability of a fuel form going forward and can provide some initial data on fuel performance particularly information about fuel-cladding interaction and fuel behavior at operating temperatures and fission densities. Screening concepts are under irradiation or planned for irradiation at the Advanced Test Reactor of Idaho National Laboratory for a number of industrial partners (AREVA, GE, Westinghouse) and national laboratories (Los Alamos National Laboratory, Oak Ridge National Laboratory). Baseline PIE will be performed on all irradiated rodlets and will consist of visual examinations, neutron radiography, gamma spectrometry, dimensional inspections, fission gas analysis, optical microscopy, and chemical analysis. Additional evaluations may also be performed. The first capsules have arrived at the hot cell for initial post-irradiation examination. Results from initial examinations are presented along with example results from equivalent exams on similar fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Refinements of a treat model for transient analysis

In order to prepare for the TREAT reactor being taken off standby, the Gateway for Advanced Modeling and Simulation Lab has used a set of KENO models to simulate certain TREAT temperature- limited transients using the Transient Reactor excursion simulator (T-ReX) code, a hybrid deterministic-stochastic code. These transients are from the M8CAL experiment series, a set of experiments done to calibrate the reactor for the planned M8 series of fuel tests. Recently, the model has been updated in media definitions and geometry. Models were examined and calculations were performed to low uncertainty values to find the k{sub eff} and reactivity insertions. Alterations to modeled values of shutdown and initial transient rod insertion depth were made to create models which had experimental values of reactivity insertion, pre-transient k{sub eff}, or both. It was found that two of the altered media inputs, fuel and Zircaloy-3 cladding, had significant effect on the k{sub eff}. Surprisingly, the Zircaloy-3 composition had a larger effect on k{sub eff}. The reactivity insertions in this model's 2855, 2856 and 2857 temperature-limited transients were found to be above experimental values by a factor of 5.5% to 7%, and that pre-transient k{sub eff} was too high, at 1.002. It was found that increasing shutdown rod insertion by 2 to 2.5 cm decreased pre-transient k{sub eff} to 1.0, and that increasing shutdown rod insertion by 3 to 5 cm and decreasing initial transient rod insertion by 1 to 2 cm gave perfect pre-transient k{sub eff} and total reactivity insertion. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Analysis of the absorbed hydrogen in cladding tubes applied in the QUENCH-LOCA tests

A shortened loss of coolant scenario was simulated experimentally at fuel rod bundle scale in the QUENCH-LO test. All pressurized rods burst at temperatures between 790 and 855 deg. C. The hydrogen distribution in the cladding of chosen fuel rod simulators was studied by neutron radiography and tomography. It was shown that the time between burst and quenching as well as the temperatures occurring during this period, and hence the degree of inner oxidation, are important parameters for the hydrogen uptake. No hydrogen uptake was measured for the specimen with 63 s or less between burst and quenching. For longer time intervals, hydrogen enriched bands were formed. These bended bands are oriented non-symmetrically to the tube axis. Micro-structure investigations give hints that the hydrogen keeps in solution in the a-zirconium lattice. The hydrogen enriched bands influence the mechanical properties of the rods, characterized by a higher hardness. Brittle fracture occurs in these bands in the tensile test. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Community Data Contribution to M.E.T.A. with ATF-relevant Hydrided Zr cladding (Coated and Uncoated)

Since the aftermath of the Fukushima Daiichi loss-of-coolant accident, accident-tolerant fuel (ATF) claddings have been developed to improve the coping times in such events. However, the mechanical performance of ATF cladding is crucial in ensuring that it does not negatively impact the mechanical integrity during all other stages of the nuclear fuel cycle, and the validity of the existing safe operating margins must be verified. However, due to the cladding’s tube geometry and textured anisotropy, determination of apparent mechanical properties under certain deformation paths is challenging. In the uniaxial hoop direction, for instance, the measured mechanical stresses include frictional forces caused by loading mandrels or varying deformation paths in the sample during traditional ring tensile testing. This experimental difficulty is exacerbated by the specimen size. However, addressing these challenges enables irradiation separate-effects investigations in which the materials can be inserted in reactors like the High Flux Isotope Reactor, and reducing the material consumption of commercially irradiated material allows for further post-irradiation examinations. Despite the advantages of reduced-scale mechanical testing, any drawbacks from new specimen geometries must be evaluated, and uncertainties from specimen preparation, setup, and analysis methodologies must be understood.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO 2 ) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗