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Uncertainty Quantification and Management for Multi-scale Nuclear Materials Modeling
Understanding and improving microstructural mechanical stability in metals and alloys is central to the development of high strength and high ductility materials for cladding and cores structures in advanced fast reactors. Design and enhancement of radiation-induced damage tolerant alloys are facilitated by better understanding the connection of various unit processes to collective responses in a multiscale model chain, including: dislocation nucleation, absorption and desorption at interfaces; vacancy production, radiation-induced segregation of Cr and Ni at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels; investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S); time evolution of swelling (cluster growth) phenomena of irradiated materials; and energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip.
Insight into the interfacial microstructure and chemistry of hot isostatically pressed AA6061-AA6061 bonds for U-10Mo fuel cladding application
Here, in the present study, hot isostatic pressing (HIP) bonding was carried out using AA6061-AA6061 plates for U-10Mo monolithic fuel cladding application. HIP was performed at 560°C with an applied isostatic pressure of 103 MPa for 90 min. Interfacial microstructure characterizations were performed using scanning electron microscopy (SEM) combined with electron backscatter diffraction (EBSD) to quantify precipitate fractions and study recrystallization, grain growth, and orientation relationships across the interface. EBSD results confirmed that no interfacial grain growth occurred during the HIP process. Transmission electron microscopy (TEM) was also employed to quantify composition and identify phases of various precipitates and an oxide layer that formed at the interface during HIP bonding. Microstructural analyses revealed the formation of a large fraction of Mg 2 Si precipitate and a ~50 nm thick Mg-oxide layer at the interface. Further TEM analysis confirmed that the oxide layer was not continuous, and Mg 2 Al 2 O 5 oxide particles with an average size of 18 nm were present in the oxide layer. Finite element modeling analysis was performed to study the stress distribution at the interface during HIP processing.
Sister Rod Destructive Test Results (FY19)
This report documents work performed under the Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-19ORO010201026 within work package SF-19OR01020102 and is an update to the work reported in M2SF-19OR010201028 and M3SF-19OR010201027. The document incorporates comments received with respect to the previous reports and incorporates additional test results. As a part of the US Department of Energy (DOE) Office of Nuclear Energy (NE) High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zircaloy-4 (LT Zirc-4), ZIRLO ® , and M5 ® . The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and consolidation activities. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans. The DEs are performed using a phased approach, and the Phase 1 DEs being performed at ORNL include: • full-length rod heat treatments (FHT) of 3 selected sister rods to examine the effects of temperatures reached during dry storage preparation, • rod internal pressure and void volume measurements of the 3 FHT rods and 3 corresponding baseline rods, as well as 2 additional rods selected for depressurization/gas transmission tests and fatigue lifetime tests, • fission gas sampling and analysis, • depressurization and gas transmission tests, • rough segmentation of the selected rods for mechanical tests and rod characterization, • fuel sampling and burnup analysis, • metallography, • cladding total hydrogen measurements, • mechanical testing, including fatigue lifetime (CIRFT), four-point bend, axial tension, microhardness, ring compression, and burst tests. Table S-1 provides a summary of the DE status. The mechanical testing will be performed using fueled segments and is expected to complement previous and current mechanical test results using defueled cladding segments.
Fuel cladding post-quench LOCA embrittlement: mechanical test relevance
During a loss-of-coolant-accident (LOCA), the fuel cladding is submitted to a high temperature increase in a steam environment, along with an internal pressure increase which would eventually lead to ballooning and burst of the cladding. Steam could then penetrate through the burst opening and lead to inner surface oxidation and secondary-hydriding of the cladding. Post-quench cladding rupture can then occur either in two-sided oxidized area with wall thinning due to ballooning, or in single-sided oxidized area with large hydrogen content. Common tests involved in the evaluation of the post-quench ductility are Ring Compression Tests (RCT), Expansion Due to Compression Test (EDC) or Axial Tensile Tests (AT). Questions arise about these mechanical tests, as for instance, the influence on sample failure of the oxide layer thickness partitioning between inner and outer diameter. A systematic study was carried out to compare several mechanical tests on both single-sided and two-sided Zircaloy-4 oxidized samples. Five different mechanical tests were compared, including RCT: C-shaped ring Compression Test (CCT), Axial Tensile test (AT), Ring Tensile Test with well-defined machined gage section (RTT) and plain Ring Tensile Test (pRTT). All samples were taken from cladding tubes which had been oxidized in steam at 1200 Celsius degrees and then water-quenched. Mechanical tests were performed at room temperature and two to three samples were tested for each oxidation conditions and for every sample geometry to assess the reproducibility. Metallographic examinations were carried out to measure the ZrO{sub 2} and αZr(O) layer thicknesses and the ex-β/α(O) phase fraction. Mean hydrogen content measurements were also performed systematically on each sample. Results indicate that the different mechanical tests have not the same sensitivity to the oxide layer partitioning: RTT and pRTT load at failure seems to depend mainly on the cumulated oxide layer thickness rather than on its partitioning, whereas this parameter has an impact on CCT and AT test results. For all tensile tests, stresses at failure were derived from load-displacement curves. The failure of brittle samples was assumed to be governed by linear elastic fracture mechanics. Using conventional stress intensity factor solution, a value of the material fracture toughness controlling the sample failure was determined from experimental results and discussed. (authors)
The use of gamma spectrometry in the determination of fission products migration in irradiated fuel
Non destructive examinations realized in hot cells of LECA STAR facility give main data on irradiated fuel rods and pins. Among those examinations, gamma spectrometry allows access to fuel, inside the cladding, thanks to gamma rays of fission products such as {sup 137}Cs, {sup 154}Eu coming from pellets... From those gamma scannings we can detect the position and the length of the fuel column and of its pellets, calculate local burnup.. In the database of our lab we have already such gamma scannings on hundreds of rods or pins with different fuels, claddings and irradiations conditions (under nominal or non nominal). We have detected that in specific cases, an unusual shape of the {sup 137}Cs scanning (distribution quite different from those of {sup 154}Eu) can be explained by the migration of this isotope, moving to the cold sides of the pellet. This phenomenon is mainly associated with an increasing of the pellet's temperature. Based on our observations, we have developed a quantitative approach of the changes on the {sup 137}Cs scannings through the calculation of appropriate 'indicators'. Those calculations allow us to be able to localize, quantify and compare the {sup 137}Cs migration all along rods or pins. Those migration results are so quickly and easily achievable from gamma measurements and can then be easily correlated to other observations realized with destructive examinations, puncturing and calculations, already realized or to come. Because such a migration is the result of temperature increasing in the pellets, our indicators can be directly associated with this local temperature. In perspective, thermically activated phenomena such as geometrical changes in the shape of the pellets, fission gas release... can also indirectly be deduced from our indicators. (authors)
An Overview of Material Models in the BISON Fuel Performance Code
BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. BISON is designed to be a multi-fuels code. Although primarily applied to Light Water Reactor (LWR) fuel to date, the code has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models which are often complex and nonlinear. Nearly all of the material models currently available in BISON are empirical, using correlations from existing fuel performance codes or material property libraries. Although these models perform adequately under many conditions, they have well-understood limitations. One obvious weakness is that the models are only valid in the domain of the associated experimental data, and extrapolating beyond this domain can result in substantial errors. A further weakness is that these models are often correlated in terms of imprecise quantities, such as burnup. Materials can, for example, reach a given burnup under very different circumstances, leading to very different material behavior. To overcome these weaknesses, efforts are underway with BISON to develop models based on a multi-scale approach, where material behavior is described with a series of interdependent microstructure based models. This paper provides an overview of the material models available in BISON. To keep the scope manageable, only LWR fuel will be considered. The first section focuses on UO{sub 2} fuel, identifying the models generally required and briefly describing those that have been implemented in the code. The second section repeats this approach for Zircaloy cladding. A final section provides examples of areas where multi-scale material modeling has been applied to BISON. BISON is a modern 2D and 3D nuclear fuel performance code that is applicable to multiple fuel forms. A wide variety of complex and nonlinear material models are required to simulate fuel behavior. Focusing on UO{sub 2} fuel and Zircaloy cladding (LWR fuel), the set of empirical models currently implemented in BISON have been reviewed. To overcome some of the deficiencies associated with empirical models, multi-scale material models are being developed. Two multi-scale models that have been used in BISON were reviewed. (authors)
Assessment of PCMI Simulation Using the Multidimensional Multiphysics BISON Fuel Performance Code
Since 2008, the Idaho National Laboratory (INL) has been developing a next-generation nuclear fuel performance code called BISON. BISON is built using INL’s Multiphysics Object-Oriented Simulation Environment, or MOOSE. MOOSE is a massively parallel, finite element-based framework to solve systems of coupled non-linear partial differential equations using the Jacobian-FreeNewton Krylov (JFNK) method. MOOSE supports the use of complex two- and three-dimensional meshes and uses implicit time integration, which is important for the widely varied time scales in nuclear fuel simulation. MOOSE’s object-oriented architecture minimizes the programming required to add new physics models. BISON has been applied to various nuclear fuel problems to assess the accuracy of its 2D and 3D capabilities. The benchmark results used in this assessment range from simulation results from other fuel performance codes to measurements from well-known and documented reactor experiments. An example of a well-documented experiment used in this assessment is the Third Risø Fission Gas Project, referred to as “Bump Test GE7”, which was performed on rod ZX115. This experiment was chosen because it allows for an evaluation of several aspects of the code, including fully coupled thermo-mechanics, contact, and several nonlinear material models. Bump Test GE7 consists of a base-irradiation period of a full-length rod in the Quad-Cities-1 BWR for nearly 7 years to a burnup of 4.17% FIMA. The base irradiation test is followed by a “bump test” of a sub-section of the original rod. The bump test takes place in the test reactor DR3 at Risø in a water-cooled HP1 rig under BWR conditions where the power level is increased by about 50% over base irradiation levels in the span of several hours. During base irradiation, the axial power profile is flat. During the bump test, the axial power profile changes so that the bottom half of the rod is at approximately 50% higher power than at the base irradiation level, while the power at the top of the rod is at about 20% of the base irradiation power level. 2D BISON simulations of the Bump Test GE7 were run using both discrete and smeared pellet geometry. Comparisons between these calculations and experimental measurements are presented for clad diameter and elongation after the base irradiation and clad profile along the length of the test section after the bump test. Preliminary comparisons between calculations and measurements are favorable, supporting the use of BISON as an accurate multiphysics fuel simulation tool.
In-pile Thermal Conductivity Characterization with Time Resolved Raman
The project is designed to achieve three objectives: (1) Develop a novel time resolved Raman technology for direct measurement of fuel and cladding thermal conductivity. (2) Validate and improve the technology development by measuring ceramic materials germane to the nuclear industry. (3) Conduct instrumentation development to integrate optical fiber into our sensing system for eventual in-pile measurement. We have developed three new techniques: time-domain differential Raman (TD-Raman), frequency-resolved Raman (FR-Raman), and energy transport state-resolved Raman (ET-Raman). The TD-Raman varies the laser heating time and does simultaneous Raman thermal probing, the FR-Raman probes the material’s thermal response under periodical laser heating of different frequencies, and the ET-Raman probes the thermal response under steady and pulsed laser heating. The measurement capacity of these techniques have been fully assessed and verified by measuring micro/nanoscale materials. All these techniques do not need the data of laser absorption and absolute material temperature rise, yet still be able to measure the thermal conductivity and thermal diffusivity with unprecedented accuracy. It is expected they will have broad applications for in-pile thermal characterization of nuclear materials based on pure optical heating and sensing.
Behavior of contaminants and simulant fission products in the advanced chlorination process
The University of Tennessee, Knoxville and Oak Ridge National Laboratory have completed proof-of principle testing on an advanced sulfur-based zirconium and aluminum chlorination process for the recovery of Zr or Al from spent fuel claddings and purification of the products. This process could be used to recover hafnium-free nuclear-grade zirconium for reuse or disposal as low-level waste. Testing showed that even non-optimized reaction protocols could produce zirconium product streams from simulated irradiated fuel with purities >90%. The chlorination process was tested on representative fission product species that would be found in Advanced Test Reactor fuel (ATR). First, fission product generation was determined through modeling and simulation of the ATR. The generation of fission products was used to selected contaminants and potentially volatile fission product simulants for laboratory testing. The goal is to identify which species may react with the sulfur-chlorine reagents and be transported into the advanced chlorination off-gas.
High-burnup Experiments in Reactivity Initiated Accidents (HERA)
HERA is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). During RIAs the fuel cladding can be breached by one of two mechanisms; pellet-cladding mechanical interaction or high temperature balloon and rupture. Pellet-cladding mechanical interaction (PCMI) failure occurs in the early phase of an RIA transient where the fuel pellet expands rapidly into the cladding prior to any meaningful heat transfer from the pellet to the cladding. This results in high stresses in the cladding resulting in crack propagation through the base Zircaloy metal and axially along the cladding tube. Failure by PCMI is a brittle failure mode and occurs at low cladding strains, approximately 2% hoop strain. If the cladding survives the PCMI phase of the transient, it may still fail due to ballooning and rupture induced by extended exposure to high temperature. Thermal transport from the fuel pin to the coolant is significantly reduced if a transition to film boiling on the cladding surface occurs. This result in both a rapid increase in fuel temperature, which drives fission gas release, and a spike in cladding temperature, which results in loss of mechanical strength. If the internal pressure of the fuel rod is greater than the system pressure the cladding can inelastically deform and eventually rupture. Failure by swelling and rupture occurs at larger strains, generally greater than 5% hoop strain. While PCMI failures are generally seen as more limiting, the dominance of one failure mode over the other can be affected by both the cladding material conditions, and the transient evolution.
High-burnup Experiments in Reactivity Initiated Accidents (HERA)
HERA is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). During RIAs the fuel cladding can be breached by one of two mechanisms; pellet-cladding mechanical interaction or high temperature balloon and rupture. Pellet-cladding mechanical interaction (PCMI) failure occurs in the early phase of an RIA transient where the fuel pellet expands rapidly into the cladding prior to any meaningful heat transfer from the pellet to the cladding. This results in high stresses in the cladding resulting in crack propagation through the base Zircaloy metal and axially along the cladding tube. Failure by PCMI is a brittle failure mode and occurs at low cladding strains, approximately 2% hoop strain. If the cladding survives the PCMI phase of the transient, it may still fail due to ballooning and rupture induced by extended exposure to high temperature. Thermal transport from the fuel pin to the coolant is significantly reduced if a transition to film boiling on the cladding surface occurs. This result in both a rapid increase in fuel temperature, which drives fission gas release, and a spike in cladding temperature, which results in loss of mechanical strength. If the internal pressure of the fuel rod is greater than the system pressure the cladding can inelastically deform and eventually rupture. Failure by swelling and rupture occurs at larger strains, generally greater than 5% hoop strain. While PCMI failures are generally seen as more limiting, the dominance of one failure mode over the other can be affected by both the cladding material conditions, and the transient evolution.
limiting rods selection in large break LOCA analysis with the RELAP5-3D code
The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, compared to the current practice in which only one hot rod needs to be analyzed. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods under LOCA conditions. In this work, the best-estimate plus uncertainty analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. With the new analysis method, the limiting transient case and the limiting rods can be easily identified in response to the proposed new rulemaking. (authors)
Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3
The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with known or suspected PCI failures to determine how well they can be estimated.
Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3
The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.
Corrosion test of LWR cladding tubes in high temperature seawater
In the case of cooling system failure in spent fuel pools (SFP) due to station black out and/or loss of coolant accidents, a pool water level must be kept high enough to avoid overheat of spent fuels. Seawater is one of the candidate coolant materials in such emergencies. In the present work, we have studied the corrosion behavior of zirconium alloys in seawater. Tube samples of Zircaloy-2 and -4 claddings were immersed in artificial seawater (3, 10 and 20 wt.%) at high temperature (80 - 85 Celsius degrees). Boric acid was added to the seawater at a corrosion test of Zircaloy-4. Two types of specimens were prepared according to the simulation of circumferential tensile stress. One was a plugged tube specimen, which was pressurized to around fuel rod pressure in a spent fuel pool by argon gas. Another specimen was of C-ring shape, where large circumferential tensile stress was partially induced at the surface. A notch was longitudinally grooved to increase the local tensile stress on some of the C-ring specimens. The specimen surface was observed by a scanning electron microscopy (SEM) after the corrosion test. After the 200 day long corrosion test for the plugged tube specimens without boric acid, Mg(OH){sub 2} deposits were observed on the surface of Zircaloy-2 and -4 specimens whereas Mg(OH){sub 2} was not observed with boric acid because of probably a lower pH. Neither pitting corrosion nor stress corrosion cracking (SCC) was observed on the surface of all plugged tube specimens. Only in the C-ring specimens in 20 wt.% seawater without boric acid and in 3 and 10 wt.% seawater with boric acid, small cracks were observed at the notch. The results of the present work show that Zircaloy-2 and -4 have a high tolerance towards corrosion and SCC in high temperature seawater.
Development of an Electrochemical Impedance Instrument for Monitoring Corrosion Formation (Oxide and Hydride) on Cladding Materials
The objective of this work package is to develop electrochemical impedance spectroscopic (EIS) techniques for the in pile measurement of changes in cladding materials in coupling with model simulation and material characterization. This work package will involve the initial development of electrochemical sensing technologies for measuring spatial and time resolved changes in cladding chemistry. Specific attention will be paid to monitoring changes in monitor cladding corrosion including oxide and hydride formation and deformation. The planned work during this period was focused on cladding material zirconium alloy corrosion, oxide and hydride development, and to characterize samples using EIS under controlled gas atmosphere, atom probe tomography (APT), Raman Spectroscopy, and Synchrotron X-ray diffraction (XRD). The corrosion, including oxide and hydride of zirconium alloy samples, was tested in this study. The samples were treated at a gas environment containing 3% H2 (vol%) balanced with N2 gas at various temperatures (100 – 500°C) and different times (up to 60 hours) and were characterized by in situ EIS. After treatment, various characterization techniques were applied to measure the samples, including TGA, Ramen, XRD, and APT. This study found that impedance spectra can be affected significantly with (furnace/environment) temperature. Higher temperature resulted in higher overall impedance, and longer exposure time exerted similar effects. Proof of concept characterization of oxidized samples, including TGA, Raman, XRD, and APT, were accomplished on samples exposed to 700 oC gas environment containing 20% O2 and 80% N2. TGA provides observable breakaway and bulk oxide growth measurements. High-temperature, in-situ Raman provides observation of oxidation stress at the surface of the oxide. Additionally, a transition from tetragonal to monoclinic zirconia can be seen early in the oxidation experiment. Specifically, tetragonal-rich zirconia is seen at the metal/oxide interface, where a transition to monoclinic-rich zirconia is seen in the bulk of the oxide. It has also been confirmed that highly compressive residual stresses are seen near the metal/oxide interface. It is concluded that the transition from tetragonal to monoclinic zirconia is induced by massive shifts in stress as the oxide grows. APT has supplied evident differences in the distribution of zirconium and oxygen at the metal/oxide interface. On the pre-breakaway sample, the metastable Zr(O) region at the boundary is relatively thin, interfaced by Zr-rich (i.e., metal) and O-rich (i.e., ZrO2) on either side. In contrast, it has been seen on a post-breakaway sample that the Zr(O) metastable region is much greater in thickness, with evidence of four “zones” showing slightly contrasting composition between 49-54% Zr. In addition, it can be seen from the Zr-2.65Nb post-breakaway sample that secondary element concentrations are considerably different than bulk composition measurements show. Specifically, Nb concentration is much lower than expected, whereas Fe concentration is much higher than expected. These differences could play a role in each element’s contribution to oxide growth mechanisms, as well as causation for breakaway. Future work is focused on analyzing EIS spectra of hydride and its correlation with characterization features, finite element (FE) modeling of cladding materials, electrochemical sensor development, targeting cladding oxides/hydrides under thermodynamic equilibria, during transients and under irradiation.
Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation
Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.