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DOE OSTI2021

Unraveling the Early-Stage Ordering of Krypton Solid Bubbles in Molybdenum: A Multimodal Study

Self-organization of defects such as fission gas bubbles in materials can lead to high inventory capacity for fission gas storage and help mitigate swelling caused by fission gases in nuclear fuel materials under radiation in nuclear reactors. Here, we report the physical mechanism of self-organization of krypton (Kr) gas bubbles in molybdenum (Mo) under ion implantation. The ion fluence and temperature-dependent formation of Kr solid bubble superlattice (SBS) in Mo were investigated by using both synchrotron-based small-angle X-ray scattering (SAXS) and transmission electron microscopy (TEM). Early stage self-organization of gas bubbles is observed at a fluence of 2.5 × 10 16 ions/cm 2 at temperatures of 300–400 °C. The bubble lattice constant increases with increasing implantation temperature from 300 to 400 °C. Both experiments and atomic kinetic Monte Carlo modeling indicate that the Kr solid bubbles are weakly ordered in comparison to previously studied helium (He) gas bubble superlattice (GBS) while the lattice constant are relatively smaller for Kr SBS compared to that of He GBS. The irradiation conditions suggest that spinodal decomposition, which is a form of phase separation, probably precedes gas bubble ordering in Mo. Altogether, our work sheds light on the formation mechanism of noble gas superlattice toward the development of radiation-tolerant materials which are important for the design of advanced nuclear reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Effects of fast and thermal neutron irradiation on Ga-polar and N-polar GaN diodes

Studies of the radiation tolerance and electrical behavior of gallium nitride (GaN) based devices are important for the next generation of high-power and high-voltage electronics that may be subjected to harsh environments such as nuclear reactor and fusion facilities, particle accelerators, and post-denotation environments. Here, in this work, we study the behavior of Ga-polar and N-polar GaN Schottky diodes before and after exposure to fast and thermal + fast neutrons. Temperature-dependent current–voltage (I–V) and circular transmission line method (CTLM) measurements were used to study the electrical characteristics. A strong reduction in reverse leakage current and an increase in differential resistance in forward bias were observed after neutron irradiation. Thermionic emission (TE), Frenkel–Poole (FP) emission, and Fowler–Nordheim (FN) tunneling models were used to explain the forward and reverse I–V characteristics pre- and post-irradiation. The study confirms that Ga-polar and N-polar GaN Schottky diodes exhibit different electrical responses to fast and thermal neutron irradiations. The reverse bias characteristics of N-polar diodes are less affected after the fast neutron irradiation compared to Ga-polar diodes, while in the forward bias region, the electrical behavior after fast and thermal neutron irradiations is similar in Ga-polar and N-polar diodes. The results indicate that the role of orientation should be considered in the design of GaN-based radiation-tolerant electronics.

42 ENGINEERING↗

DOE OSTI2023

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

A Novel Dew Point Meter: Application to the Measurement of the Sulfuric Acid Dew Point for Combustion Flue Gas

Accurate knowledge of acid dew point is essential for industrial and applied combustion applications. Sulfur in the fuel or raw materials is converted to sulfur dioxide (SO2) during combustion, and a portion of the SO2 is oxidized to sulfur trioxide (SO3). The SO3 will react to form H2SO4 vapor when in the presence of water vapor. Even with just trace levels of H2SO4 vapor in the gas phase (1-10 ppm), the dew point can reach 100°C and higher. To avoid acid condensation and the resulting corrosion on heat recovery equipment, plant engineers must ensure that surface temperatures are above the acid dew point, but this decreases the efficiency of thermal energy recovery. Thus, there is a trade-off between minimizing equipment corrosion and maximizing thermal energy recovery, and the acid dew point is a key parameter for this optimization. Commercially available acid dew point meters use electric conductivity sensors. These sensors are known to greatly underestimate the dew point due to their low sensitivity. In addition, no validation testing has been reported for these units and they are often expensive. In this work, we analyze the theory of the sulfuric acid condensation and develop a novel dew point meter based on this analysis. The meter consists of a novel optical instrument that is designed to monitor the slightest appearance of condensation on a hydrophobic window surface as the surface temperature of the window is slowly decreased. In this way, an accurate measurement of the dew point is obtained under a wide range of concentrations. The basis of the instrument is that a collimated beam from a diode laser will generate forward scattered light when the beam encounters surface condensate, and a sophisticated array detector is used to sensitively monitor the onset of light scattering. The measurement procedures are established to rapidly find the acid dew point, while minimizing error. Further, to calibrate the dew point meter we developed a calibration system based on a liquid bubbler that can generate a stable gas flow with a known sulfuric acid dew point. Test results show that the dew point meter can accurately measure acid dew point over a wide range. For H2SO4 vapor concentrations as low as 6 ppm the acid dew point is measured with an error of only ~1°C. To demonstrate the versatility of this instrument, the dew point meter was adapted for use with a high-pressure flow cell to allow for measurements of the dew point of flue gas from pressurized oxy-fuel combustion in a 100 kWth pressurized reactor.

Cheng, Mao↗

DOE OSTI2020

Divertor heat flux challenge and mitigation in SPARC

Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m -2 . This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m -2 ) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Fast Neutron Irradiation of a Multichannel JFET-Based Optical Encoder

Modern electrical components are susceptible to damage from high levels of radiation and extreme temperatures found near reactors in terrestrial nuclear power plants and in aerospace applications. Radiation-hardened electronics are being developed, largely for the aerospace industry, but they sometimes rely on application-specific, small-batch semiconductor fabrication processes. These processes tend to be prohibitively expensive to develop and maintain outside major industrial facilities or governmental agencies. Recently, commercially available, nonradiation-rated junction-gate field-effect transistors (JFETs) were shown to maintain their functionality at gamma doses exceeding 1 MGy, suggesting that nonrated, commercially available electrical components could be used to develop systems that are tolerant to ionizing radiation. However, gamma ray survival is not indicative of neutron dose survival, and few studies characterize JFETs under neutron irradiation. To address this knowledge gap, a JFET-based analog multiplexer and optical pulsewidth modulation (PWM) encoder was developed and irradiated using a 252 Cf source to 1.6×10 13 n/cm 2 . The multiplexed optical encoder (MOE) system maintained functionality throughout testing and showed little evidence of radiation effects. These results indicate that circuitry tolerant to fast neutron damage can be developed using low-cost, nonradiation-rated, commercially available JFETs, which could provide a lower production cost alternative to specialized semiconductor processes when designing and building electronics better able to survive neutron irradiation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

SATS Transient Fission Gas Release Test with Irradiated Fuel under Loss of Coolant Accident Conditions

The transient fission gas release (tFGR) during the temperature ramp associated with a loss-of-coolant accident (LOCA) in light-water reactors (LWRs) is likely a significant contribution to the total pressure in a fuel rod and may cause an unexpected rod burst. The lack of data related to tFGR continues to be a key gap in understanding LWR cladding burst behavior under LOCA conditions. To fully characterize this behavior, tFGR data must be collected from several different systems that can capture all relevant testing conditions. Oak Ridge National Laboratory (ORNL) has developed a system to measure the integral tFGR from irradiated fuel segments. This system was designed to integrate with the existing Severe Accident Test Station (SATS) and to build upon decades of experience capturing fission gas to characterize fuel behavior. The tFGR system consists of a sweep gas system to transport gases from the in-cell SATS apparatus to an out-of-cell fission gas detection system composed of a series of cold traps to capture the off-gas from the heating tests and a gamma spectrometry system to detect and measure 85 Kr. Initial system testing operations were completed during which 85 Kr collection and measurement were verified along with the ability to detect stable inert gases. A tFGR test with a high-burnup fuel specimen was successfully conducted by the in-cell SATS-tFGR system at ORNL. The posttest examination is under way, the result of which will be reported in FY24.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2009

Fission Surface Power Technology Development Testing at NASA's Early Flight Fission Test Facility

Fission surface power (FSP) systems could be used to provide power anytime, anywhere on the surface of the Moon or Mars. FSP systems could be used at polar locations, at locations away from the poles, or in permanently shaded regions, with excellent performance at all sites. A potential reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system (FSPS) is also readily extensible for use on Mars. At Mars the system would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Under the NASA Exploration Technology Development Program (ETDP), NASA and the Department of Energy (DOE) have begun technology development on Fission Surface Power (FSP). The primary customer for this technology is the NASA Constellation Program which is responsible for the development of surface systems to support human exploration on the moon and Mars. The objectives of the FSP technology project are: 1) Develop FSP concepts that meet expected surface power requirements at reasonable cost with added benefits over other options. 2) Establish a hardware-based technical foundation for FSP design concepts and reduce overall development risk. 3) Reduce the cost uncertainties for FSP and establish greater credibility for flight system cost estimates. 4) Generate the key products to allow Agency decision-makers to consider FSP as a viable option for flight development. To be mass efficient, FSP systems must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial systems. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference FSP system uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance on the surface of the moon or Mars. Recent testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference FSP system, and has helped evaluate methods for system integration. In June, 2009, a representative pumped NaK loop (provided by Marshall Space Flight Center) was coupled to a Stirling power converter (provided by Glenn Research Center) and tested at various conditions representative of those that would be seen during actual FSP system operation. In all areas, performance of the integrated system exceeded project goals. High-temperature NaK pump testing has also been performed at the EFF-TF, as has testing of methods for providing long-duration NaK purity.

Houts. Michael G.↗

DOE OSTI2019

Size effects on thermo-mechanical performance of U-10Mo monolithic fuel plates

Monolithic fuel is a fuel form that is considered for the conversion of high performance research reactors. This plate-type fuel consists of a high density U-Mo fuel in monolithic form that is sandwiched between zirconium diffusion barriers, and encapsulated in an aluminum cladding. To date, large number of plates have been irradiated with satisfactory perforamce. The program is now moving into the qualification phase, a predecessor to the timely conversion of the target reactors. It must be shown that the fuel system meets the safety standards and performs well in reactor. The requirement to satisfactory irradiation performance under normal operating conditions is primarily demonstrated by a successful testing. Since each reactor employs distinct fuel plate geometries for various consideration with unique plate design features and attributes, a single “generic” plate geometry capturing all of the extremities is not achievable. Furthermore, testing all these geometric and irradiation parameters on a large size plate is not practical. Therefore, a smaller, “down-scaled” versions of fuel plates, are often employed for experimental purposes. This limitation consequently requires much more cautious performance evaluations, as thermal and mechanical response of a plate with certain geometry may not be representative for a plate with a different geometry. To investigate if plate size has any effects on irradiation performance, the plates with various geometric dimensions were parametrically evaluated. In particular, length and width of the plates were varied between the bounding values. Temperature, deformation, stress values were comparatively evaluated. The results have indicated that effects of geometric ratios and plate size variations in length and width directions are insignificant. However, wider plates could become more prone to a warping-type deformation, if there are nonlinearities.

42 ENGINEERING↗

DOE OSTI2019

Size effects on thermo-mechanical performance of U-10Mo monolithic fuel plates

42 ENGINEERING↗

DOE OSTI2025

Development and application of two-step uncertainty propagation and sensitivity analysis methodology for fast reactor safety analysis

Uncertainty quantification (UQ) in nuclear reactors for transients is directly linked with safety assessment through the cross-sections uncertainties, provided as a covariance matrix, which are propagated through the reactor system to output of interest pertaining to reactor safety, such as peak temperatures in fuel/clad/coolant. Using a two-step approach, uncertainties are first quantified and propagated from basic input variables (such as reaction cross-sections) to intermediate quantities (such as reactivity feedback coefficients) through lattice level calculations. Uncertainties of intermediate quantities (from the first step) are then propagated through the system transient calculations, in the second step, to obtain uncertainties on reactor safety output parameters of interest. The scope of this work consists of Uncertainty Quantification & Propagation of nuclear data uncertainties that are highly correlated through unprotected transient overpower and unprotected loss of flow to assess their impact on core safety parameters. This two-step approach in the presence of covariance renders the sensitivity analysis very challenging. In fact, usually the sensitivity analysis is restricted to each step, which limits its application since the sensitivities between the system output quantities and the basic input variables are difficult to obtain. Here, in this work, we address this issue by proposing a simple, general methodology to combine the sensitivity indices obtained in each step by assuming the model behavior being linear. For the first step Generalized Perturbation theory based indices are used while in the second step the recently studied Johnson indices. The uncertainty quantification and sensitivity methodologies discussed here are demonstrated on a generic LFR design which is based on the 500 MWth demonstration Lead-cooled fast reactor (DLFR) using oxide fuel, developed by Westinghouse Electric Company (WEC).

42 - ENGINEERING↗

NASA NTRS2015

Ultra High Temperature (UHT) SiC Fiber (Phase 2)

Silicon-carbide fiber-reinforced silicon-carbide ceramic matrix composites (SiCSiC CMC) are emerginglightweight re-usable structural materials not only for hot section components in gas turbine engines, but also for controlsurfaces and leading edges of reusable hypersonic vehicles as well as for nuclear propulsion and reactor components. Ithas been shown that when these CMC are employed in engine hot-section components, the higher the upper usetemperature (UUT) of the SiC fiber, the more performance benefits are accrued, such as higher operating temperatures,reduced component cooling air, reduced fuel consumption, and reduced emissions. The first generation of SiCSiC CMC with a temperature capability of 2200-2400F are on the verge of being introduced into the hot-section components ofcommercial and military gas turbine engines.Today the SiC fiber type currently recognized as the worlds best in terms ofthermo-mechanical performance is the Sylramic-iBN fiber. This fiber was previously developed by the PI at NASA GRC using patented processes to improve the high-cost commercial Sylramic fiber, which in turn was derived from anotherlow-cost low-performance commercial fiber. Although the Sylramic-iBN fiber shows state-of-the art creep and rupture resistance for use temperatures above 2550oF, NASA has shown by fundamental creep studies and model developmentthat its microstructure and creep resistance could theoretically be significantly improved to produce an Ultra HighTemperature (UHT) SiC fiber.This Phase II Seedling Fund effort has been focused on the key objective of effectively repeating the similar processes used for producing the Sylramic-iBN fiber using a design of experiments approach to first understand the cause of the less than optimum Sylramic-iBN microstructure and then attempting to develop processconditions that eliminate or minimize these key microstructural issues. In so doing, it is predicted that that theseadvanced process could result in an UHT SiC fiber with 20 times more creep resistance than the Sylramic-iBN fiber,which in turn would allow SiCSiC CMC to operate up to 2700oF and above, thereby further enhancing the performancebenefits of SiCSiC components in aero-propulsion engines. It was also envisioned that the fiber processes developedduring Phase II efforts would not only reduce production costs for the UHT fiber by using low-cost precursor fibers andcombined processes, but also allow the UHT fibers to be directly produced in preforms of the precursor fibers, possibly atthe facilities of the CMC fabricator.

SiC Ceramic Fiber↗

DOE OSTI2025

Root cause analysis of a molten salt pump in FLUSTFA

The primary salt pump installed in the high-temperature FLUoride Salt Test Facility (FLUSTFA) was successfully operated for some time, but later ceased operation. To understand what occurred, a Root Cause Analysis (RCA) was performed. Steps taken to try to get the pump operational include adjusting the shaft position, increasing the heating power of the tape heaters on the pump volute, and manually rotating the pump shaft. While removing the insulation, corrosion was noted on the outside of the pump volute, and decolorization of the insulation and tape heaters was observed. Significant corrosion products were also observed in the pump itself and the piping connected to the pump. The nitrogen cover gas was maintained from before salt was introduced into the loop until the pump was dismounted and continues to be maintained even after the pump was removed. After considering probable scenarios, causes were assigned and corrective actions were developed to prevent those causes. Then, the RCA was presented to an advisory committee for review, the “Review Committee,” consisting of experts in large molten salt systems: Brandon Haugh, David Holcomb, Kevin Robb, and Vicente Rojas. As a result, the advisory committee provided comprehensive feedback, which have been incorporated into a revised RCA. Findings have then been summarized and reported in this publication.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Report on Year-4 of Water NSTF Matrix Testing: Facility Maintenance and Accident Testing

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large-scale test facility has been constructed at Argonne National Laboratory to generate NQA-1 qualified validation data for passive decay heat removal systems in advanced reactors. The Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS) and is intended to study the behavior, bound performance, and ultimately guide design decisions for passive decay heat removal systems for advanced reactors. In addition to the experimental activities detailed in this report, a supportive computational modeling effort is on-going which has been demonstrated to significantly strengthen the experimental program while also improving accuracy of the computer models. Together these create a mutually beneficial relationship integral to meeting the overall program objective of examining the heat removal performance of the RCCS concept. This report serves as a summary of maintenance and experimental activities during the program’s fourth year of water-based operation. A planned six-month maintenance period began in August 2021, during which major inspections, repairs, cleaning, and installation of new instrumentation and data acquisition hardware were conducted. Most significantly, two heaters that faulted during Year-3 were repaired, allowing the facility to resume use of the full heated section area and full range of available electric power. The remainder of the year consisted of eight months of test operations, during which the facility logged 211 hours of active heating across one bake-out (following the maintenance period) and seven matrix tests; five classifieds as Accepted per NQA-1, one as Trending, and one as Failed. Testing began by performing two repeat cases to confirm expected facility response and behavior during both single- and two-phase flow conditions, ensuring no changes were introduced during the maintenance period that might have altered the thermal-hydraulic characteristics of the facility. In continuation of the power parametric series initiated in previous years, a high-power test case was then performed examining heat removal performance at a decay heat load equivalent to 2.4 MWt, full-scale, a level exceeding maximum design targets. Additional testing then introduced various blockage scenarios along the network piping, examining the effects of partial and complete blockages of the flow paths on the system behavior and heat removal performance. A study of static boiling tests directed at understanding the geysering two-phase instability was also conducted. The loop was filled only to the bottom of the tank outlet, creating an open loop configuration that prevents any natural circulation flow from occurring, and the heaters were powered on until the facility reached saturation conditions. Following, a series of quasi-steady-state conditions were introduced by adjusting the inventory level in the adiabatic chimney piping at decreasingly lower elevations above the heated region. A strong correlation of geysering characteristics and loop level was observed, with flow and temperature excursions decreasing in intensity, but increasing in frequency, as the fill were reduced to lower elevations along the chimney piping. Once the level fell very low in the chimney, at points near the top of the heated section, the system reached a stable state of continuous boiling without any occurrence of geysering eruptions. A final significant testing accomplishment this year was successful completion of an “accident scenario” test, whose operating conditions were based on a prototypic decay heat curve provided by Framatome and scaled for the NSTF. This test began by establishing steady-state, single-phase “normal operation” conditions, before simulating an accident trip where the availability of active cooling systems was lost. Loop temperatures gradually increased until reaching saturation and subsequent two-phase boiling flow. Over the course of an extended operational period along the defined decay heat curve, steam boil-off caused gradual but continued depletion of liquid inventory until reaching a critically low level causing flow stagnation and cessation of natural circulation heat removal. At this point, after nearly 72 hours of continuous operation, a cold refill was performed to replenish the system inventory and allow the facility to re-establish closed loop natural circulation flow and return to a safe operational state.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗

DOE OSTI2024

Solid breeder material crystal structure evolution due to Li burn up–Loss of crystal stability

Solid breeder materials for fusion power plants are subjected to extreme conditions of radiation damage, high temperature and designed component burn-up. Here, the loss of lattice stability of Li 2 TiO 3 due to Li burn-up was investigated using first principles-based approaches to aid evaluation of upper limit of component lifetime. The lattice stability is analyzed using calculated phonon dispersion in the Li 2 TiO 3 supercell, while Li burn-up is modeled by the introduction of Li-vacancies. It has been determined that the studied ceramic can be structurally stable up to ~40% of Li atoms burn-up. At 50% burn-up, the negative frequency branch in the phonon dispersion spectrum appears indicating the loss of lattice dynamical stability. Moreover, the structure obtained by introduction of unstable frozen phonons with minima energy amplitude also results in a phonon dispersion with negative branches in the structure obtained. Therefore, the system completely loses stability when approximately half of Li atoms are consumed.

36 MATERIALS SCIENCE↗

DOE OSTI2022

A continuum model for heat and mass transfer in moving-bed reactors for thermochemical energy storage

In this work, a continuum heat and mass transfer model coupling transport phenomena and high-temperature thermochemical reactions is developed for stationary packed-bed and counter-flow moving-bed reactors. After presenting the general modeling framework, we focus on the 2D axisymmetric version of the model for which validation is conducted with experimental results for a packed-bed reactor in the literature for manganese-iron oxide reduction/oxidation and an in-house counter-flow moving-bed reactor for magnesium-manganese oxide reduction up to 1450 °C. Transient simulation results including the local distributions of gas/solid temperatures, oxygen concentration and the extent of reaction, as well as the various energy flow components and energy conversion efficiencies are reported. The results based on the 2D axisymmetric model are also compared with those obtained from a previous 1D model. The comparison shows that capturing the radial variation is critical in reactor modeling and the 2D results demonstrate improved agreement with experiments. Specifically, large temperature variations along the radial direction are observed especially in the reaction zone; this non-uniform radial temperature distribution has a significant effect on the chemical reaction extent due to its strong dependence on temperature; and the overall oxygen concentration at the reactor exit and the predicted system efficiency are slightly lower in the 2D model compared to the 1D model. Finally, the present heat and mass transfer model can provide valuable insights into reactor design, scale-up, and operating conditions selection to maximize system energy storage efficiency.

25 ENERGY STORAGE↗

DOE OSTI2020

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Finally, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗