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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Enrichment Penalty for Molybdenum Cladded Accident Tolerant Fuel

A tri-layered fuel cladding has been proposed as an enhanced accident tolerant fuel (EATF) design. The EATF cladding has an inner layer of zirconium alloy, a layer of molybdenum, and an outer layer of zirconium alloy. Due to its high melting point, high temperature creep resistance, and improved resistance to steam corrosion, the molybdenum layer will better tolerate a loss of active coolant accident than the standard single layer zircaloy cladding currently in use. The addition of the molybdenum layer to the cladding, however, has a deleterious effect on the fuel management due to the high thermal neutron capture cross section of Mo{sup 95}. This negative reactivity effect must be counterpoised by increasing the fuel enrichment in order to maintain a constant cycle length. This work presents a prima facie quantification of the enrichment increase required to maintain cycle length for a standard PWR 3 x 18 month cycle. The quantification of the required enrichment increase is necessary in order for a more thorough financial analysis to be performed for the prospective new fuel design. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A preliminary study on irradiation growth behavior modelling of N36 zirconium alloy tube

As a new kind of Zr-Sn-Nb series alloy, N36 has not been explored in irradiation growth behavior before. In this paper, four empirical correlations including power, logarithmic, exponential and hyperbolic tangent law were used to build irradiation growth model of N36 zirconium alloy preliminarily. Before that, lengths of the fuel rods with N36 claddings from pool-side measurement were analyzed to get irradiation growth strains, which were then used to fit the four empirical correlations mentioned above. Distribution of fast fluence along axial direction of tubes was considered in all irradiation growth models. The fitting results showed that the hyperbolic tangent law fit irradiation growth strains the best, while power law did the worst. In order to improve the precision of the hyperbolic tangent model, several improvements were done. Verification and validation of the improved hyperbolic tangent irradiation growth model suggested that the size of second phase may have an influence on irradiation growth behavior of N36 tubes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Damage progression and failure of SiC/SiC composite tubes under hard-contact radial expansion

The response of silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composite cladding to mechanical interaction with fissile fuel is a knowledge gap that must be overcome to design and assess SiC-based cladding systems for advanced nuclear applications. This study developed the relevant mechanical testing capability and identified the failure behavior and the critical microstructural features and processing defects. Sections of SiC composite tube were subjected to a modified expansion-due-to-compression (EDC) test in an X-ray computed tomography microscope: a polyurethane plug pressed surrogate Al 2 O 3 into the inner walls of the SiC/SiC composite tubes to achieve hard contact. A pure EDC test with just a polyurethane plug was also performed as a reference. Through the use of displacement fields, digital volume correlation revealed inhomogeneous deformation fields in the tubes, even for pure EDC, which was related to the inherent defects in the structure. Deep learning–aided segmentation and systematic data analysis revealed that the presence of inhomogeneous deformation applied by the hard contact was exaggerated by the presence of inner surface imperfections left behind from the matrix densification process. In conclusion, the findings provide insights into the applications, highlighting the necessity for improvements in inner surface roughness and the incorporation of localized contacts in pellet–cladding mechanical interaction computational models.

Composites↗
DOE OSTI2016

An innovative accident tolerant LWR fuel rod design based on uranium-molybdenum metal alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The paper is organized as follows. First, an overview of the design features of the U-Mo fuel concept is summarized, including the key attributes and technical challenges to be addressed through the research and development program. Second, recent results are provided on the triple extrusion of a depleted U-Mo fuel rodlet with a niobium liner and SS-316 cladding. A status is presented next on the high temperature water corrosion testing of short samples with pre-drilled cladding defects. Lastly, an assessment of the enhanced accident tolerance features is discussed that highlights the key performance attributes of the U-Mo fuel rod design. Scoping calculations have identified that the reduced stored energy of the U-Mo fuel at the initiation of a LOCA event may reduce core temperatures below 500-600 Celsius degrees during the early phase of the LOCA event. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, anticipated operational occurrences (AOOs) and postulated accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Nuclear analyses of supercritical water cooled reactor with carbon nano-tube cladding

The authors have confirmed the feasibility of the dual layer clad comprised of iron and carbon nano-tube to problems of Super Critical Water Reactor cores. Continuous energy Monte Carlo method was applied. The difference between JENDL-3.3 and ENDF-6 was confirmed. Depletion was carried out. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal-hydraulic performance of the TREAT Multi-SERTTA for reactivity initiated accident experiments

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) program. The Multi-SERTTA is comprised of 4 isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of fresh UO{sub 2} fuel in zircaloy cladding. The results show that for energy depositions less than 1380 J/g-UO{sub 2}, the vessel environment remains sub-supercritical. Two extremes of gap conductance show significant effect on event timing and peak cladding temperature but little effect on end state conditions. Inclusion of oxidation has increasing impact for increasing energy depositions. The effect of oxidation is non-negligible for planned TREAT experiments in the range of explored energy depositions. Pulse width has little impact on thermal hydraulic conditions under the assumptions used in this analysis. The model used in this work has been coupled to the fuel performance code, BISON, to accomplish best estimate predictive capability for Multi-SERTTA experiments. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Identification of the threshold security applied for fuel under reactivity-initiated conditions by self-organizing maps and neural networks - 14386

In recent times, we have 439 reactors distributed in 31 countries that give about 377 GW, covering around 12% of world capacity of energy. Currently, it accounts for an estimated 40 serious accidents reported in nuclear power plants. In this article, turns out the behavior of fuel rod in light-water reactors under accident conditions. Besides, we investigate deeply about reactivity-initiated accidents, by introducing Nordheim-Fuchs model as state equations. The early research showed that pre-irradiated fuels and afterward subjected to a rapid pulse of energy. It adopts the key items necessary to check the threshold of failure. These items are few limits as burning the fuel; the value applied to the peak of energy, pulse width, oxidation layer, and cladding. The kinetic equations of nuclear reactors modeled as a non-linear system. In this way, it can apply the paradigm of artificial intelligence using the self-organized maps. Where the goal is to find the cluster of failure for irradiated fuels until 75 GWd/MTU. The second model based on statistical is the logistic regression theory. Hence, the tests carried out and compared against the results found by neural networks for classification of clusters. It proofs that both approaches present capacity to find the flaws accurately of about 80% of cases. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Monte Carlo simulation of a quantum noise limited Čerenkov detector based on air-spaced light guiding taper for megavoltage x-ray imaging

Purpose: Electronic Portal Imaging Devices (EPIDs) have been widely used in radiation therapy and are still needed on linear accelerators (Linacs) equipped with kilovoltage cone beam CT (kV-CBCT) or MRI systems. Our aim is to develop a new high quantum efficiency (QE) Čerenkov Portal Imaging Device (CPID) that is quantum noise limited at dose levels corresponding to a single Linac pulse. Methods: Recently a new concept of CPID for MV x-ray imaging in radiation therapy was introduced. It relies on Čerenkov effect for x-ray detection. The proposed design consisted of a matrix of optical fibers aligned with the incident x-rays and coupled to an active matrix flat panel imager (AMFPI) for image readout. A weakness of such design is that too few Čerenkov light photons reach the AMFPI for each incident x-ray and an AMFPI with an avalanche gain is required in order to overcome the readout noise for portal imaging application. In this work the authors propose to replace the optical fibers in the CPID with light guides without a cladding layer that are suspended in air. The air between the light guides takes on the role of the cladding layer found in a regular optical fiber. Since air has a significantly lower refractive index (∼1 versus 1.38 in a typical cladding layer), a much superior light collection efficiency is achieved. Results: A Monte Carlo simulation of the new design has been conducted to investigate its feasibility. Detector quantities such as quantum efficiency (QE), spatial resolution (MTF), and frequency dependent detective quantum efficiency (DQE) have been evaluated. The detector signal and the quantum noise have been compared to the readout noise. Conclusions: Our studies show that the modified new CPID has a QE and DQE more than an order of magnitude greater than that of current clinical systems and yet a spatial resolution similar to that of current low-QE flat-panel based EPIDs. Furthermore it was demonstrated that the new CPID does not require an avalanche gain in the AMFPI and is quantum noise limited at dose levels corresponding to a single Linac pulse.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Phased development of accident tolerant fuel

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1-3 corresponds to the 'proof-of-concept' stage (Phase 1), TRL 4-6 to 'proof-of-principle' (Phase 2), and TRL 7-9 to 'proof-of-performance' (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress $\sigma_\theta$ in the tested ring specimen. A scaling factor called $\chi$-factor is defined to correlate the ring load F ring into hoop stress $\sigma_\theta$, and written as $\sigma_\theta = \chi \frac{F_{ring}} {tl}$. The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.

expansion plug test↗
DOE OSTI2017

Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviors of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite materials made with added lanthanide and uranium oxides. These analyses show the corrosion behaviors of the alloy/ceramic composite materials are very similar to the corrosion behaviors of multi-phase alloy waste forms, and that the presence of oxide inclusions does not impact the corrosion behaviors of the alloy phases. Mixing with metallic waste streams is beneficial to lanthanide and uranium oxides in that they react with Zr in the fuel waste to form highly durable zirconates. The measured corrosion behaviors suggest properly formulated composite materials would be suitable waste forms for combined metallic and oxide waste streams generated during electrometallurgical reprocessing of spent nuclear fuel. Electrochemical methods are suitable for evaluating the durability and modeling long-term behavior of composite waste forms: the degradation model developed for metallic waste forms can be applied to the alloy phases formed in the composite and an affinity-based mineral dissolution model can be applied to the ceramic phases.

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

Development of the SharkFin Distribution for Fuel Lifetime Estimates in Severe Accident Codes

Historically, MELCOR has employed models at various degrees of fidelity for evaluating fuel mechanical response to the effects of cladding oxidation, material interactions (i.e., intermetallic formation), Zircaloy melting, fuel swelling, and other processes that occur at very high temperatures. In lieu of detailed models, a simple temperature-based criterion is used to define the threshold beyond which normal ('intact') fuel rod geometry can no longer be maintained, and the core materials at a particular location collapse into particulate debris. The temperature-based criterion reflects physico-chemical processes that affect fuel rod integrity. Using a composite of information, such as from the PHEBUS studies, four data points were log-linearly interpolated to create the rod collapse criteria used in the SOARCA study. This time-at-temperature (TaT) criterion was introduced in an attempt to avoid non-physical cliff-edge effects that are observed during a calculation when fuel temperatures are predicted to hover just below a failure temperature for extended periods. However, the code still models local rod collapse due to high temperature as a discrete and near-instantaneous (over one time step) event at each core cell. This paper introduces the SharkFin statistical model for TaT fuel failure relationships whose median curve has recently been adopted as the default fuel failure criteria in MELCOR. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC

Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Interdiffusion Behavior of U{sub 3}Si{sub 2} and FeCrAl via Diffusion Couple Studies

Uranium Silicide (U{sub 3}Si{sub 2}) is in use as a dispersion plate fuel in test reactors, but has not been fully tested in light water reactor (LWR) conditions. U{sub 3}Si{sub 2} pellets have been fabricated for use in commercial LWRs as part of an accident tolerant fuel system U{sub 3}Si{sub 2} has a higher uranium loading density as well as a higher thermal conductivity relative to UO{sub 2}. FeCrAl alloy is a candidate cladding material to replace the industry standard, Zircaloy. FeCrAl has outstanding high-temperature mechanical strength and oxidation resistance under accident conditions. The U{sub 3}Si{sub 2}-FeCrAl system is expected to provide more than double the amount of time from a Scram until failure criteria for the reactor is met, as well as producing roughly half the amount of H{sub 2} in the event of a loss of coolant accident (LOCA). Uranium and iron have well know eutectics that may affect/enhance diffusion of the U{sub 3}Si{sub 2}/FeCrAl fuel system. The interaction of iron with molten uranium has been investigated In the U-Fe system, UFe{sub 2}, which melts congruently at 1235 deg. C, forms a eutectic with iron that melts at 1080 deg. C and has a composition of U-77at% Fe. UFe{sub 2} also forms a eutectic with uranium that melts at 725 deg. C and has a composition of U-34at% Fe. The U-34at% Fe eutectic extended the temperature range down to 740 deg. C. Regular operating temperatures for U{sub 3}Si{sub 2} fuel is expected to be around 300 - 400 deg. C. The eutectics of iron and uranium would nominally only be of concern in accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Characterization of SiC-SiC accident tolerant fuel cladding after stress application

In this work, SiC-SiC tubes with representative LWR (water cooled reactor) cladding dimensions were subjected to cycles of out-of-pile thermal stress followed by quenching. Samples were held at elevated temperatures and dropped into a water bath to replicate the high thermal stresses generated during reflooding. The samples were then characterized to measure changes in strength, permeability, and structure. Results were evaluated both for quenching into room temperature and boiling water. The results show that SiC-SiC composite tubes quenched from 1000 Celsius degrees in air into a boiling water bath exhibited no increase in He leak rate through the cylinder wall up to a pressure difference of 34.5 kPa compared to the as-fabricated state. SiC-SiC composite tubes quenched from 300-400 C. degrees into room temperature water showed an increase in He leak rate compared to the as-fabricated state. We attribute this to the more severe thermal gradient generated through the wall thickness due to higher heat flux characteristic of nucleate boiling compared to film boiling. A persistent vapor film forms around the samples quenched into boiling water, whereas no vapor film forms for the samples quenched into room temperature water. Another important result is that not loss in strength was noted compared to unquenched samples and that the stresses generated by the quenching were not enough to cause fiber fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Interdiffusion and Reaction between Zr and Al Alloys from 425 degrees to 625 degrees C

Zirconium has recently garnered attention for use as a diffusion barrier between U–Mo nuclear fuels and Al cladding alloys. Interdiffusion and reactions between Zr and Al, Al-2 wt.% Si, Al-5 wt.% Si or AA6061 were investigated using solid-to-solid diffusion couples annealed in the temperature range of 425 degrees to 625 degrees C. In the binary Al and Zr system, the Al3Zr and Al2Zr phases were identified, and the activation energy for the growth of the Al3Zr phase was determined to be 347 kJ/mol. Negligible diffusional interactions were observed for diffusion couples between Zr vs. Al-2 wt.% Si, Al-5 wt.% Si and AA6061 annealed at or below 475 degrees C. In diffusion couples with the binary Al–Si alloys at 560 degrees C, a significant variation in the development of the phase constituents was observed including the thick t1 (Al5SiZr2) with Si content up to 12 at.%, and thin layers of (Si,Al)2Zr, (Al,Si)3Zr, Al3SiZr2 and Al2Zr phases. The use of AA6061 as a terminal alloy resulted in the development of both T1 (Al5SiZr2) and (Al,Si)3Zr phases with a very thin layer of (Al,Si)2Zr. At 560 degrees C, with increasing Si content in the Al–Si alloy, an increase in the overall rate of diffusional interaction was observed; however, the diffusional interaction of Zr in contact with multicomponent AA6061 with 0.4–0.8 wt.% Si was most rapid.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Corrosion and interdiffusion studies of U{sub 3}Si{sub 2}

U{sub 3}Si{sub 2} is a candidate fissile phase for use in accident tolerant fuel systems in light-water reactors because of its higher fissile density, higher thermal conductivity when compared to uranium dioxide. It is also expected to have acceptable performance in contact with high temperature water (e.g., a leaker event from breached cladding) and in regards to interaction with different cladding materials. In high temperature water, U{sub 3}Si{sub 2} is expected to form a protective SiO{sub 2} coating. However, the resistance of U{sub 3}Si{sub 2} to corrosion under light water reactor service conditions has been previously unknown. In this study, the corrosion tests of sintered U{sub 3}Si{sub 2} discs were conducted in pressurized water at 300 C. degrees in an autoclave. The exposure times to corrosion were 6 hours and 24 hours. Detailed composition and microstructural analysis were carried out using a combination of techniques including X-ray diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM) equipped with energy dispersive X-ray spectroscopy (EDS). Only slight weight change of U{sub 3}Si{sub 2} discs can be detected in corrosion tests up to 24 hours. A layered structure of corrosion products formed on U{sub 3}Si{sub 2}, with a glassy silicon oxide layer at the top followed by a nanocrystalline UO{sub 2} layer, a Si-rich oxide layer and then a mixture oxide layer of U and Si. A diffusion couple of U{sub 3}Si{sub 2} in contact with Zircaloy-4 has been tested at 800 C. degrees up to 100 hours. The resulting microstructure and interdiffusion have been studied with SEM/EDS. Some limited interdiffusion has been observed in these tests as well as a layer containing Zr and Si forms between the U{sub 3}Si{sub 2} disc and the Zircaloy-4 disc. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Establishment of technical basis to implement accident tolerant fuels and component to existing LWRs

Fuel cladding, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of Light Water Reactors (LWRs). Fundamental data have been mainly accumulated from the viewpoint of material technology. In order to create the technical basis for the practical use of the accident tolerant fuels (ATFs) and the other components in LWRs, a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan has been initiated. Main candidate materials of the ATFs and components in Japan are silicon carbide (SiC), stainless steel including oxide-dispersion strengthened (ODS) FeCrAl alloys, and tri-structural-isotropic (TRISO) fuel. First of all, current status of development of the candidates was investigated using information released from inside and outside Japan. Based on the results of the investigation, technology readiness of the candidates was evaluated and the attribute guide was created to clarify technical subjects for the practical use as the core components of the existing LWRs. For that purpose, a technology readiness table was reviewed in terms of fuel rod design, production and quality management, core and plant design, safety evaluation for design basis accident and severe accident, safety evaluation for transport and storage, establishment of treatment and disposal technologies, and development of standards and criteria, with 9 readiness levels. Considering the reviewed technology readiness table and the additionally evaluated influence of implementation of the new materials, a map of remaining subjects for research/development (Attribute guide) was prepared. Finally, a research plan was drafted for design and fabrication towards implementation of ATFs and components to the existing LWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗