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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Analysis of Loss-of-Coolant Accidents in the NBSR

This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Study on Neutronics Calculation Method of SCWR with Annular Fuel Assembly

Supercritical Water-Cooled Reactor (SCWR) is one of Gen-IV reactor systems for its advantages over LWRs, such as high thermal efficiency, simplified reactor system and economics. Annular fuel was proposed by MIT to supplant conventional solid fuel and improve power density in LWRs. Annular fuel is cooled from both inside and outside of the fuel. Thus a lower fuel temperature and cladding surface temperature can be achieved with higher power density. SCWRs will benefit from this advantage because high cladding temperature (usually over 650 degree C) is one challenge of SCWRs. Using annular fuel in SCWRs is very promising. This is also because the inner hole of the annual fuel can provide extra neutron moderation. Supercritical water density is too low to provide enough moderation. Different analysis codes have been developed all over the world to carry out SCWR pre-conceptual core design. Yamaji et al. developed SRAC/ASMBURN/COREBN to study Japan version of SCWR. Monti et al. used ERANOS to analyze HPLWR (High Performance Light Water Reactor). WIMS-D4/CITATION was also used to calculate HPLWR by Tashakor et al.. These codes employ a two-step method used in PWRs calculation. Lattice/Assembly calculation is carried out as the first step. In the second step, few group constant of different fuel assembly types is then used in core calculation. MCNP had also been used to study SCWR by Watta et al.. Due to large computation time cost, it's used mostly to analyze fuel assembly or core performance rather than pre-conceptual design, in which much optimization is needed. In this research, a neutronics calculation code for SCWR is developed using two-step method. This code is verified by comparing with Monte-Carlo code. Fuel assembly homogenization method is studied to improve the code accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Diffusional Interaction between U-10 wt.% Zr and Fe at 903K, 923K and 953K (630°C, 650°C, and 680°C)

U-Zr metallic fuels cladded in Fe-alloys are being considered for application in an advanced Sodium-Cooled Fast Reactor (SFR) that can recycle the U-Zr fuels and minimize the long-lived actinide waste. To understand the complex fuel-cladding chemical interaction between the U-Zr metallic fuel with Fe-alloys, a systematic multicomponent diffusion study was carried out using solid-to-solid diffusion couples. The U-10 wt.% Zr vs. pure Fe diffusion couples were assembled and annealed at temperatures, 903, 923 and 953K for 96 hours. Development of microstructure, phase constituents, and compositions developed during the thermal anneals were examined by scanning electron microscopy, transmission electron microscopy and X-ray energy dispersive spectroscopy. Complex microstructure consisting of several layers that include phases such as U 6 Fe, UFe 2 , ZrFe 2 , α-U, β-U, Zr-precipitates, Χ, ε and λ were observed. Multi-phase layers were grouped based on phase constituents and microstructure, and the layer thicknesses were measured to calculate the growth constant and activation energy. The local average compositions through the interaction layer were systematically determined, and employed to construct semi-quantitative diffusion paths on isothermal U-Zr-Fe ternary phase diagrams at respective temperatures. The diffusion paths were examined to qualitatively estimate the diffusional behavior of individual components and their interactions. Furthermore, selected area diffraction analyses were carried out to determine, for the first time, the exact crystal structure and composition of Χ, ε and λ-phases. The Χ, ε and λ-phases were identified as Pnma(62) Fe(Zr,U), I4/mcm(140) Fe(Zr,U) 2 , and I4/mcm(140) U 3 (Zr,Fe), respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Steam oxidation of chromium corrosion barrier coatings for sic-based accident tolerant fuel cladding

SiC-based materials are currently being considered for accident tolerant fuel (ATF) cladding, however, there are concerns regarding hydrothermal corrosion rates and hermeticity. These issues may be addressed with the utilization of a Cr corrosion barrier coating (CBC) but the interaction of Cr with SiC under loss of coolant accident (LOCA) conditions needs to be investigated. In the present work, Cr coatings were deposited onto chemically vapor deposited (CVD) SiC with either high power impulse magnetron sputtering (HiPIMS), cathodic arc (CA), or a combination of both. Annealing and steam thermogravimetric analysis (TGA) was used to study reaction product formation with and without oxidation. Additionally, after 1200 °C annealing, formation of a Cr x C y carbide layer at the ambient interface and a Cr x Si y C z silicide layer at the SiC interface was observed. After 1200 °C TGA exposure, similar carbide and silicide layers were observed but with a Cr 2 O 3 outer reaction layer and a Si and O rich interfacial layer forming between the carbide and silicide layers. Initial mass gain of the coatings during the 1200 °C TGA exposure was parabolic and were similar to reported rates of chromia formation. Mass gain behavior and observed transitions in chromia microstructure indicate that metallic Cr may be consumed within the first ~1 h at 1200 °C and that subsequent chromia formation occurs at the expense of the underlying Cr carbide layer. Overall, the results show no detrimental impact of Cr coatings on the steam oxidation resistance of SiC up to 4 h at 1200 °C.

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

Predicting the flow stress of Zircaloy-4 under in-reactor accident conditions - 2016-0075

During loss of coolant accident (LOCA) and reactivity-initiated accident (RIA), nuclear fuel rods experience high heating rates that change the microstructure and properties of zirconium cladding materials. The aim was to determine how different fast heating rates affect the kinetics of transformation and whether its consequences on the material yield stresses in the dual-phase region could be predicted. Phase fraction and texture of cold-rolled Zircaloy-4 were thus measured at fast heating rates similar to those experienced during LOCA and RIA by in-situ high-energy synchrotron x-ray diffraction; the yield stress was also measured at different heating rates to the same dual-phase temperature by electrical resistance method. Crystal plasticity finite element model was used to simulate the flow stresses with the inputs of the measured phase fraction and texture. The model is calibrated using the flow stress at the heating rate of 10 deg. Cs{sup -1} and used to predict the flow stress at the heating rate of 50 deg. Cs{sup -1} at the same temperature. A significant shift of the β transus temperature was identified for fast heating rates in the range of 10-100 deg. Cs{sup -1}. There is great similarity in texture at different heating rates. In the dual-phase temperature region, the a texture is almost identical to the starting texture and the β texture is nearly random. Yield stresses of cold-rolled Zircaloy-4 decrease at faster heating rate at the same temperature of 920 deg. C. A good agreement was found between the experimental and predicted flow stress at different heating rates, indicating this methodology has potential to be used to predict flow behavior in a fast transient regime. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO 2 Interactions

This status report provides the background and current status of a series of irradiation capsules that were designed and are being built to test the interactions between candidate FeCrAl cladding for enhanced accident tolerant applications and prototypical enriched commercial UO 2 fuel in a neutron radiation environment. These capsules will test the degree, if any, of fuel cladding chemical interactions (FCCI) between FeCrAl and UO 2 . The capsules are to be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to burn-ups of 10, 30, and 50 GWd/MT with a nominal target temperature at the interfaces between the pellets and clad of 350°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Geo-polymer solidification technology approved by Czech / Slovak nuclear authority to immobilise NPP resins and sludge waste - 15555

This paper presents the features and performance record of the geo-polymer solidification technology, SIAL{sup R}, licensed for use by both the Slovakian (UJD SR) and Czech Nuclear (SUJB) regulators. An overview of its performance at Bohunice in Slovakia, where it has been used to successfully treat resins and sludge wasteforms will be provided along with references to other plants where it has been used. More recently, geo-polymers have been noted as an immobilization technology which shows potential for resins produced by treatment of radioactive contaminated water resulting from the clean-up of the Fukushima Daiichi Plant. The Nuclear Power Plant A-1 located in Jaslovske Bohunice was completed in 1972 and operated for 5 years until two accidents in 1976 and 1977. After the second accident in February 1977 (INES level 4) the Nuclear Power Plant was permanently shut down for decommissioning. Fuel assemblies and fuel cladding damaged in the accidents led to significant Sr-90, Cs-137 and transuranic contamination. As a consequence of the long-term corrosion of the barrier's material, highly contaminated sludge accumulated which could not be effectively immobilized using conventional methods such as cementation or bitumen treatment due to adverse physical-chemical properties and high specific activity of the waste (Cs-137). This challenge led to AMEC developing and licensing the geo-polymer, SIAL{sup R}. Today, this geo-polymer is one of the most proven and widely used for on-site solidification of radioactive materials such as sludge, resins, sorbents and organic liquids. There are numerous benefits to using this geo-polymer for waste immobilisation. It can be used at room temperature and can incorporate up to approximately four times as much waste (sludge, resin) as a Portland cement matrix equivalent depending on the waste being treated. In addition, it is characterized by excellent mechanical and physical properties, compared with the earlier generation techniques. This includes higher mechanical strength, lower leachability, low volatility, posing a low fire risk and excellent physical stability in the presence of frost and water (no distortion or cracking). Dedicated facilities to receive and encapsulate waste from different locations can be created but this geo-polymer can also provide efficient and practical on-site treatment of radioactive waste streams. This is achievable because the equipment used to encapsulate waste using SIAL{sup R} is also modular, flexible and versatile. At Bohunice, the facility for encapsulation was designed to occupy existing areas of the plant in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles (ROVs) and for the waste to be immediately treated on site without the need for transportation saving both time and resource. To achieve this, some of the rooms and areas of the building were assessed and modified to ensure that the waste package flow was suitable whilst ensuring that appropriate safety measures were in place such as adequate shielding to reduce or entirely eliminate dose uptake of workers. To date, SIAL{sup R} has been used to successfully immobilize approximately 500 tons of waste which includes sludge and resin from Bohunice, Slovakia and approximately 250 m{sup 3} of spent ion exchange resins from tanks on site at the Dukovany nuclear plant, Czech Republic. It now has a track record of over 15 years which includes on-going research and development. The International missions WANO and OSART evaluated the SIAL{sup R} matrix technology at NPP Dukovany as an example of good practice. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Overview of recent studies related to lead-alloy-cooled fast reactors

The recent progress of the studies related to lead alloy-cooled reactors (LFR) and the accelerator driven system (ADS) is summarized. The compatibility of materials with lead alloys has been clarified under steady and transient temperature conditions. Higher Cr content, Si and Al addition and Al-Fe alloy-coating improved the corrosion resistance of steels. The Al-Fe alloy-coated steel was not corroded even under high temperature transient conditions. The ceramics of SiC and Si{sub 3}N{sub 4} are expected to be used as cladding material for high temperature LFR. For the analytical consideration of mass transport in lead alloy circuit, the diffusion coefficient of Ni was measured using the capillary methods. A new bubble visualization method in LBE using gamma-ray radiography was developed. The thermal interaction of lead-bismuth eutectic (LBE) and lead droplets with sub-cooled water, and the behaviors of droplet fragmentation were investigated, and the visualization of volatile liquids in high temperature liquids was achieved.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Step-loaded creep testing of Zircaloy-4 cladding at higher temperatures in the α-phase

A refined understanding of zirconium-based cladding thermomechanical performance during rapid transients is essential for enhancing the safety and operation of light-water reactors. Traditional models for zirconium alloys under accident conditions generally assume that creep dominates fuel cladding performance. Here, these historic models have largely remained unchanged and serve as the basis for safety criteria development. As the U.S. nuclear industry pursues higher burnup levels, the increased release of fission gases during transients raises the risk of cladding failure in the low-temperature hcp α-phase, making the fidelity of these models of greater importance. Creep testing was conducted from 550–700°C with 25–120 MPa applied hoop stresses to investigate Zircaloy-4 deformation at accident-relevant temperatures in the α-phase. Step-loading was employed to capture creep behavior across a wide stress range from a single sample. The stress-strain rate data at higher temperatures (650 and 700°C) were well-described by isotropic versions of the Erbacher and Kaddour models, while the lower temperature data (550 and 600°C) were underpredicted by both anisotropic and isotropic model variants. Greater strain rates during the initial loading step at 650 and 700°C were attributed to recrystallization and grain growth of sub-micron crystallites. Yet, texture analysis revealed the basal split texture remained after testing. These observations produced results suggesting Zircaloy-4 claddings experience higher creep rates across the α-phase than previously thought, possibly related to dynamic anisotropy due to temperature dependent activation of deformation mechanisms, effects of biaxial loading, and variation in material condition between the current testing used in previous model development.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Microstructural Characteristics of U-Zr Ingot Slug and Atomized Powder for SFR metal Fuel

Rod-type fuel slugs and spherical powder of U- 10wt.%Zr alloy was also fabricated and characterized in microstructure by injection casting, gravity casting and centrifugal atomization process. U grains became finer from ∼30μm in conventionally ingot rod to ∼2μm in atomized powder. Laminar structure became so fine from ∼0.2μm in conventionally cast ingot rod to 0.05∼0.1μm in atomized powder. Metallic fuels such as U-Zr and U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR), related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's. Metallic fuel slugs of U-Pu-Zr alloys for SFR have conventionally been fabricated by an injection casting method. U-Pu-Zr alloy melts are injected and cast into quartz tube molds during injection casting. Metallic fuel slugs are collected by fracturing quartz tubes from the molds. Demolded quartz tubes are treated with radioactive wastes. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. The metallic fuel for SFR should be remotely fabricated under a radiation shielded environment such as a hot cell, because it contains highly radioactive MA such as Np, Am and Cm. The addition of MA in the fuel composition introduces new challenges in the manufacturing of a metallic fuel slug. Am has a high vapor pressure under a vacuum atmosphere at above a melting temperature of U-Zr or U-Pu-Zr. Used graphite crucibles and quartz molds cannot be recycled easily because of difficult decontamination of radioactive residues on their surfaces. It is necessary to minimize the vaporization of Am and volume of radioactive wastes. Atomized metallic fuel does not need these quartz molds during fabrication, as uranium alloys are prepared as a particle form. It is possible to reduce the formation of radioactive wastes by applying the fabrication process of a particulate metallic fuel. Metallic fuel having a fine microstructure, which originates from a rapid solidification effect during atomization process, results in a higher FG release rate during irradiation. In the case of applying a new forming method such as vibrocompaction in Zr sheath, the sheath can act as a reliable FCCI (fuel/cladding chemical interaction) barrier and a re-constraint for external elongation. In addition, atomized fuel powder, dispersed in a Na liquid metal having a high thermal conductivity, may have a lower temperature distribution during irradiation. In order to develop an efficient fabrication process of metallic fuel for SFR, U-Zr ingot rods were fabricated in a lab-scale with various casting methods. Rod-type U-Zr fuel slugs were fabricated by gravity casting and injection casting. Spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. And then, the microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscopy. Atomized U-10Zr powder generally showed spherical morphology of below 200 μm in particle size. U grains became finer from ∼30 μm in conventionally ingot rod to ∼2 μm in atomized powder. Laminar structure became so fine from ∼0.2 μm in conventionally cast ingot rod to 0.05∼0.1 μm in atomized powder. Atomized particulate metallic fuel for SFR has a fine microstructure, resulting in a higher fission gas release rate during irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

New Strategies for Licensing the Storage and Transportation of High Burn-up Spent Nuclear Fuel in the United States - 12546

An alternative approach may be needed to the licensing of high-burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates. Packaging strategies and regulations should be developed to reduce the potential for requiring fuel to be repackaged unnecessarily. This would lessen the chance of accidents and mishaps during loading and unloading of casks, and decrease dose to workers. A packaging approach that shifts the safety basis from reliance upon the fuel condition to reliance upon an inner canister could eliminate or lessen the need for repackaging. In addition, the condition of canisters can be more readily monitored and inspected than the condition of fuel cladding. Canisters can also be repaired and/or replaced when deemed necessary. In contrast, once a fuel assembly is loaded into a canister and placed in a storage overpack, there is little opportunity to monitor its condition or take mitigating measures if cladding degradation is suspected or proven to occur. (authors)

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

Materials Challenges in Nuclear Energy

Nuclear power currently provides about 13% of the worldwide electrical power, and has emerged as a reliable baseload source of electricity. A number of materials challenges must be successfully resolved for nuclear energy to continue to make further improvements in reliability, safety and economics. The operating environment for materials in current and proposed future nuclear energy systems is summarized, along with a description of materials used for the main operating components. Materials challenges associated with power uprates and extensions of the operating lifetimes of reactors are described. The three major materials challenges for the current and next generation of water-cooled fission reactors are centered on two structural materials aging degradation issues (corrosion and stress corrosion cracking of structural materials and neutron-induced embrittlement of reactor pressure vessels), along with improved fuel system reliability and accident tolerance issues. The major corrosion and stress corrosion cracking degradation mechanisms for light water reactors are reviewed. The materials degradation issues for the Zr alloy clad UO2 fuel system currently utilized in the majority of commercial nuclear power plants is discussed for normal and off-normal operating conditions. Looking to proposed future (Generation IV) fission and fusion energy systems, there are 5 key bulk radiation degradation effects (low temperature radiation hardening and embrittlement, radiation-induced and modified solute segregation and phase stability, irradiation creep, void swelling, and high temperature helium embrittlement) and a multitude of corrosion and stress corrosion cracking effects (including irradiation-assisted phenomena) that can have a major impact on the performance of structural materials.

nuclear fuels↗
DOE OSTI2026

Documentation, assessment, and improvements of fission gas modeling capabilities in BISON

Fission gas behavior critically affects nuclear fuel performance, influencing thermal conductivity, microstructure, swelling, and fuel-cladding interactions. The Simple Integrated Fission Gas Release and Swelling (Sifgrs) model, implemented within the BISON fuel performance code, provides advanced capabilities for modeling these phenomena. This manuscript reviews and assesses Sifgrs’ modular structure and submodels describing intragranular, intergranular, and release mechanisms. The modular implementation enables rapid improvements as our understanding advances and new fuel forms emerge. After an overview of capabilities, we present a comprehensive validation suite, comparing predictions to experimental data from tests under normal and transient conditions, including reproduction of the Vitanza curve. Results demonstrate Sifgrs’ accuracy, robustness, and applicability. Current limitations and ongoing efforts are discussed, including mechanistic, multiscale modeling, uncertainty quantification, and validation within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, supporting the safe and efficient use of nuclear fuels.

BISON↗
DOE OSTI2021

Grid-to-rod fretting wear study of SiC/SiC composite accident-tolerant fuel claddings using an autoclave fretting bench test

Grid-to-rod-fretting (GTRF) in pressurized water reactors (PWRs) is known to cause wear and surface damage on the fuel claddings, potentially leading to radioactive leakage. One of the accident-tolerant fuel (ATF) concepts is to use advanced cladding materials that could withstand higher temperatures. Here, we investigated the wear behavior of candidate silicon carbide (SiC)-based composite claddings with different levels of surface finish in fretting against a commercial ZIRLO alloy grid using a unique bench-scale autoclave GTRF rig. The experiments mimicked the environment in an industrial full-assembly PWR simulator. Fretting tests were conducted with a realistic load (~0.5 N) in deionized water under a pressure of 20–23 bar at 204 °C for 100 h. While the SiC/SiC composite claddings showed significantly higher wear resistance than the commercial ZIRLO alloy cladding as expected, the smoother versions experienced surprisingly higher wear than the much softer counterface, ZIRLO grid. The wear mechanism of the SiC/SiC cladding was attributed to the SiC wear debris that was trapped at the fretting interface causing both 3-body and 2-body (embedded into the grid surface) abrasion of the cladding. Rougher SiC/SiC claddings had less material loss but caused more wear on the ZIRLO grid. Finally, pre-oxidized ZIRLO grid showed better compatibility with the SiC/SiC cladding to protect both the cladding and grid as a result of reduced wear debris trapping.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Development of high performance ODS alloys

This project aims to capitalize on insights developed from recent high-dose self-ion irradiation experiments in order to develop and test the next generation of optimized ODS alloys needed to meet the nuclear community's need for high strength, radiation-tolerant cladding and core components, especially with enhanced resistance to void swelling. Two of these insights are that ferrite grains swell earlier than tempered martensite grains, and oxide dispersions currently produced only in ferrite grains require a high level of uniformity and stability to be successful. An additional insight is that ODS particle stability is dependent on as-yet unidentified compositional combinations of dispersoid and alloy matrix, such as dispersoids are stable in MA957 to doses greater than 200 dpa but dissolve in MA956 at doses less than 200 dpa. These findings focus attention on candidate next-generation alloys which address these concerns. Collaboration with two Japanese groups provides this project with two sets of first-round candidate alloys that have already undergone extensive development and testing for unirradiated properties, but have not yet been evaluated for their irradiation performance. The first set of candidate alloys are dual phase (ferrite + martensite) ODS alloys with oxide particles uniformly distributed in both ferrite and martensite phases. The second set of candidate alloys are ODS alloys containing non-standard dispersoid compositions with controllable oxide particle sizes, phases and interfaces.

36 MATERIALS SCIENCE↗