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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. Further, a large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2020

Innovative accident tolerant fuel concept enabled through direct manufacturing technology

Nuclear energy is one of the largest sources of carbon-free electricity in the world. Some countries are looking at new ways to support and revitalize the nuclear sector since the Fukushima disaster. The accident tolerant fuel program is geared toward improving the safety of nuclear energy by investigating materials that can replace or modify the current uranium-dioxide nuclear fuel and zirconium-based cladding. This research program is being supported by all major nuclear countries since 2011. The practical limitations on allowable uranium enrichment has taken the focus away from the most promising fuels with high radioactivity retention such as tristructural isotropic particle fuel. To overcome such enrichment limitation, a new fuel concept is proposed using advanced ceramic direct manufacturing with laser-induced chemical vapor deposition. The fuel-as-fiber concept is an accident-tolerant fuel design that features high thermal conductivity, strong capability of radioactivity retention and most importantly requires reasonable enrichment levels with uranium nitride as the fuel. In this work, the initial fabrication of uranium-based fuel with laser-induced chemical vapor deposition technology is demonstrated. Then an advanced multi-physics guided modeling approach based on finite element analysis codes and informed by the manufacturing capabilities is developed to accelerate the advancement of fuel-as-fiber concept for use in current light-water reactor technology. The detailed thermomechanical analysis showed promising results for viability of the innovative fuel-as-fiber concept. The predicted stresses in the fuel structural materials were similar to the case of tristructural isotropic particle fuel experience base that has shown excellent reliability in retention of fuel radioactivity at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Comparative analysis of temperature dependent properties of commercial nuclear fuel pellet and surrogates undergoing cracking: A review

A variety of normal and accident scenarios can generate thermal stresses enough to cause cracking in light water reactor (LWR) fuel. To better understand this behavior, cracking experiments have been carried out to induce thermal gradients in fuel pellets via induction heating and direct resistance heating. In this study, Ceria (CeO 2 ) and yttria stabilized zirconia (YSZ) pellets were chosen as a surrogate fuel material for uranium dioxide (UO 2 ). Induction heating was done using copper coils and molybdenum susceptors which heated the surrogates to a threshold temperature that is sufficiently high for the fuel material to conduct current. Thereafter, direct resistance heating was achieved by passing current through the specimen using a DC power supply to introduce volumetric heating to replicate reactor operating conditions. Simultaneous real-time dual imaging of the ceria pellet surface has been designed using optical and infra-red camera system to capture images of cracks and full-field temperature gradients on pellet. It was observed that YSZ pellets need a much higher voltage-current for volumetric heating and do not exhibit the cracking pattern as ceria and UO 2 , hence YSZ was ruled out for further consideration. Ceria was found to be an appropriate surrogate for UO 2 . Furthermore, the experimental set up and test conditions demand the need to primarily understand the thermo-physical, mechanical and optical properties of UO 2 and the surrogates. Thus, it is essential to have an in-depth knowledge about the various temperature dependent properties of UO 2 and CeO 2 for establishing a comparative analysis between the surrogate and UO 2 for conducting experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this work, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. Consequently, while the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

42 ENGINEERING↗
DOE OSTI2024

A mesoscale 3D model of irradiated concrete informed via a 2.5 U-Net semantic segmentation

The concrete biological shield in light-water reactors is exposed to neutron and gamma irradiation, which deteriorates the concrete’s mechanical properties in the long term. To assess the irradiation-induced damage, predictive mechanical models are developed and used in parallel with the characterization of irradiated concrete samples. Realistic 3D simulation domains can drastically improve a model’s prediction. In this work, we utilized x-ray computed tomography (XCT) data of a concrete specimen to reconstruct its 3D microstructure. The XCT data shows low contrast between the concrete’s aggregates and cement paste, resulting in poor image segmentation when using traditional unsupervised techniques. To address this issue, we developed and trained a 2.5D U-Net model on only 24 pre-labeled XCT layers to segment 651 layers of the XCT data. The overall F1-score of the model is approximately 96%. Then, we created a 3D finite element (FE) mesh based on the stack of segmented images. The FE model contains radiation-induced expansion, damage, and creep. The constitutive equations are adapted to each phase (aggregates and cement paste). Here, we simulated the effects of neutron irradiation in the concrete specimen as well as the specimen’s mechanical response to uniaxial compression. Finally, model validation was performed using experimental data on similar concrete specimens in the literature.

2.5D U-Net↗
DOE OSTI2020

Oxidation behavior of Fe-Cr-2Si alloys in high temperature steam

FeCrSi alloys have been proposed as alternatives to FeCrAl alloys for accident-tolerant fuel cladding in light water reactors. Experiments on the oxidation resistance of three Fe-Cr-2Si alloys in high-temperature steam, with 12, 16, and 20 wt. % Cr, show that ≥16 % Cr is sufficient to form a thin, passivating, binary oxide layer. Additionally, at 1200 °C, only Cr 2 O 3 and SiO 2 form on Fe 16 Cr 2 Si and Fe 20 Cr 2 Si, while additional ternary Fe oxides form on Fe 12 Cr 2 Si. Decohesion and spalling of oxide were observed on Fe-Cr-Si alloys exposed to steam at 900 °C and above. Fe 20 Cr 2 Si exhibits the best oxidation resistance of the three in high-temperature steam.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Effect of grain refinement on high temperature steam oxidation of an FeCrAl alloy

FeCrAl alloys are promising candidates to replace Zr alloys as fuel cladding materials in nuclear light-water reactors. Grain refinement has been indicated to improve irradiation resistance. To enhance corrosion resistance as well, the effects of grain refinement on steam corrosion behavior were investigated in this work. Samples of Kanthal D alloy (Fe-21Cr-5Al) with two different grain sizes (coarse-grained and ultrafine-grained) were exposed to steam at 1200 °C for 2 hrs. In conclusion, results indicate improved steam corrosion resistance in ultrafine-grained Kanthal D with formation of a thinner protective Al oxide layer and the presence of a thin underlying Cr oxide layer.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Assessment of BISON capabilities for component-level prediction of tritium transport in fusion and fission applications

Here we assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution. . The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Elucidating the Impact of Flow on Material-Sensitive Critical Heat Flux and Boiling Heat Transfer Coefficients: An Experimental Study with Various Materials

In this work, steady-state internal flow boiling critical heat flux (CHF) experiments were carried out at 10°C inlet subcooling and atmospheric pressure (84 kPa at an altitude of about 5300 feet) with mass flux ranging from 200 to 2,000 kg/m 2 -s on seven tube materials, including the accident-tolerant fuel cladding candidate FeCrAl (C26M) alloy. At a relatively low flow rate, the sensitivity of CHF to different materials was appreciable (the average difference among the tested materials was 19% at 200 kg/m 2 -s). As the absolute CHF differences stayed constant for higher flow rates, the increased flow rate led to reduced relative CHF differences among the tested materials (average relative difference of 15% at 1,000 kg/m 2 -s). FeCrAl alloy claddings that were oxidized under simulated Light Water Reactor (LWR) conditions yielded CHF comparable to as-received FeCrAl alloy cladding, with less than 1.8% difference. Experimental CHF results imply marginal material sensitivity impacts on the departure from nucleate boiling ratio in steady-state nuclear reactor cores. Yet, for accident progression analysis, material sensitivity may hold importance, as boiling behavior variation at low mass flux is appreciable. Macroscopic material parameters (surface wettability, surface roughness, and thermal properties) do not provide a suitable explanation for observed material-sensitive CHFs. Yet, surface roughness and thermal effusivity revealed power and linear relations, respectively, to the nucleate boiling heat transfer coefficients at low mass flux. At high mass flux, material sensitivity on nucleate boiling heat transfer significantly decreased due to the increased shear stress on boiling surface overwhelming material-related boiling implications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A Eulerian three-fluid flow framework for the study of fuel dispersal behavior under loss-of-coolant accident conditions

As the nuclear industry explores new fuel designs to accommodate increased burnup, studying fuel behavior during loss-of-coolant accidents is essential for ensuring the safe operation of light-water reactors. Here, this study presents a computational model to simulate the intricate three-phase flow dynamics of fuel dispersal following cladding breach, where solid fuel particles and carrier fission gases interact with the surrounding liquid or gas medium. The model utilizes a Eulerian three-fluid framework to depict the phase interaction within interpenetrating continua, treating fragmented fuel as granular material. Closure relationships for bulk-flow properties of the solid phase and detailed interfacial exchange terms in high-temperature, high-speed gas-solid-liquid flow scenarios are proposed. The model was validated using adiabatic experimental results in the literature, focusing on characterizing fuel dispersal behavior during loss-of-coolant accidents conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched probability density functions reported in experimental studies. Key highlights of this study include the theoretical description of a coupled three-phase flow with phase change, interpretation of the high-pressure boundary conditions and insights into the transient behavior of gas-solid-liquid phase dynamics during a simulated high-speed dispersal event.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Bridging microscale to macroscale mechanical property measurements of FeCrAl alloys by crystal plasticity modeling

FeCrAl alloys are candidates for accident tolerant fuel cladding of light water reactors. In this work, a microstructure- and temperature-dependent crystal plasticity model is employed to bridge microscale to macroscale mechanical property measurements of FeCrAl alloys. With the visco-plastic self-consistent (VPSC) polycrystal plasticity framework, a mechanism-based single crystal plasticity (MSCP) model adopts the Arrhenius type rate equation to describe the dependence of the critical resolved shear stress for dislocation slips on their temperature-dependent intrinsic frictional resistance and the microstructure-dependent irradiation hardening. The intrinsic frictional resistance associated with {110}<111> and {112}<111> slip systems were measured by in-situ micromechanical testing on unirradiated/irradiated samples at 25-500 °C. The irradiation hardening is estimated by the Bacon-Kocks-Scattergood (BKS) model with density and size of radiation-induced defects measured from microstructural characterization. Several features associated with thermo-mechanical behavior of unirradiated/irradiated polycrystalline FeCrAl alloys are captured. High density of deformation-induced dislocations and radiation-induced defects results in obvious hardening at room temperature, which is weakened at high temperature, and facilitates damage evolution during deformation. Moreover, both high temperature and radiation-induced defects, which facilitate dislocation multiplication, trigger large hardening rate. Finally, the proposed method together with application of accelerator-based ion irradiation technique is a surrogate approach to simulate neutron damage, improving the efficiency associated with evaluation of mechanical properties of FeCrAl alloys exposed to temperature, stress and radiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Three-dimensional bubble reconstruction in high burnup UO 2

In light water reactor (LWR) UO 2 fuels, the evolution of volatile fission products is one of the critical areas of fuel behavior that is yet to be fully understood. In UO 2 irradiated to high burnups, it is well known that most released fission gases come from the central region of the fuel as opposed to the highly porous high burnup structure (HBS) on the periphery of the pellets. However, fuels with and without interconnected bubble networks at the fuel center showed high to moderate release fractions, which conceals the mechanisms responsible for the gas release in the latter scenario. In this work, focused ion beam tomography was used to investigate the three-dimensional bubble structure in an irradiated LWR UO 2 fuel pellet with high degree of fission gas retention so that the degree of bubble interconnection could be assessed. Six radial locations with different burnups and temperatures were serially sectioned and imaged to reconstruct the three-dimensional bubble structure. As expected, the highest porosity was observed at the periphery of the fuel (HBS). Here, the porosity then decreased towards the pellet center, except for the centermost location. This location had a slightly higher porosity than its adjacent mid-radial location, which was attributed to the temperature difference between the two locations. This study provides a first-time volumetric evaluation of the porosity at different radial locations on a UO 2 fuel pellet. During this investigation, no significant bubble interconnection was noted at any of the six radial locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Radiation response of FeCrAl-coated Zircaloy-4

Coating the surface of Zircaloy-4 light water reactor fuel cladding tube is of considerable interest for enhancing its accident tolerance. In this study, thermal annealing of FeCrAl-coated Zircaloy-4 at 725 °C for 500 h, was used to induce interfacial reactions between Zircaloy-4 and FeCrAl. The interface zones were then irradiated by 3.5 MeV Zr ions at 400 °C, up to 50, 100, 150 peak dpa values. Transmission electron microscopy (TEM) and scanning TEM were used to characterize microstructural and composition changes before and after ion irradiation. Three interfacial phases were identified: FeZr 3 , (Fe,Cr) 2 Zr, and ZrC. The widest intermetallic layer, FeZr3, had large grains. The narrower phases, (Fe,Cr)2Zr and ZrC, contained small grains and were often mixed. The unexpected observation of the ZrC phase was attributed to the presence of very small impurity level concentrations of carbon in the powder material. No void swelling was observed in any of the phases, including the FeCrAl coating and Zircaloy-4 substrate. (Fe,Cr) 2 Zr, however, fully amorphized after irradiation, even at the lowest dpa.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Chemical vapor deposited (CVD) SiC and nuclear-grade SiC/SiC composite materials were exposed to light water reactor (LWR) conditions at the MIT Nuclear Reactor (MITR) here. Three sets of samples were exposed within flowing PWR-chemistry-coolant under either 1) neutron and ionizing irradiation with resulting radiolysis products, 2) radiolysis products without neutron flux, 3) loop coolant. Post-irradiation examination (PIE) exhibited an increase in corrosion from radiolysis products (particularly H 2 O 2 ), and potentially a further increase from radiation damage. For CVD SiC, regardless of exposure location, the in-situ observed corrosion rate was less than 1 µm per year, potentially within acceptable limits for LWR application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding

As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at Oak Ridge National Laboratory. A second generation of these alloys bearing additional alloying elements over the previously tested model alloys were tested in boiling water reactor (BWR) conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe–10Cr–6Al–2Mo, Fe–13Cr–5Al–2Mo, Fe–13Cr–6Al–2Mo, Fe–13Cr–7Al–2Mo, and Fe–13Cr–5Al–2Mo–1Nb were tested in normal water chemistry (NWC) and hydrogen water chemistry (HWC) for 9 months in continuously refreshing autoclaves. Commercial FeCrAl alloy APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Fe–13Cr–7Al–2Mo performed worst with an average mass loss of 1.3 mg/cm2 over 9 months. This mass loss represents an estimated thickness loss of approximately 60 μm over 6 years. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm2 or mass gains up to 0.05 mg/cm2. Based on the results of this testing, the 2nd generation FeCrAl alloys tested exhibit low wall thickness loss and are suitable in terms of corrosion resistance for use as LWR cladding in BWR-HWC and BWR-NWC under normal operating conditions.

Raiman, Stephen↗
DOE OSTI2020

Hydrothermal corrosion behavior of CVD SiC in high temperature water

Here, the hydrothermal corrosion of polished and as-cut high purity chemical vapor deposited (CVD) SiC was studied in a constantly refreshing water loop. Light water reactor (LWR) conditions were simulated at 288, 320, and 350 °C with dissolved gas concentrations between 0.15 and 3 ppm H 2 or between 1 and 4 ppm O 2 . In hydrogenated water, the rate of material loss was low, calculated to be ~1.3 μm of recession after 5 years of service in 320 °C water. Moreover, there was no observed localized attack at any temperature. In oxygenated conditions, the corrosion rate was higher, with a calculated material loss >10 μm after 5 years of service in 1 ppm O 2 , 320 °C water. Mass loss significantly increased when grain fallout became significant (as early as 200h with 4 ppm O 2 at 350 °C or after 1000–2000h with 2 ppm O 2 at 288 °C). Grain fallout more than doubled the corrosion rate and a steady state corrosion rate in the grain fallout regime was not observed but expected to eventually occur once large grains begin to be removed. Polished specimens had lower mass loss than unpolished coupons. A kinetic analysis of the data in this work suggests that the corrosion rates are controlled by a single activation step in both oxygenated and deoxygenated conditions, with the reaction order with respect to oxygen being 1. A resulting reaction rate equation to predict corrosion of SiC (in mg/cm 2 s) in high purity water from 288 to 350 °C and up to 4 ppm O 2 was constructed: Rate = $\frac{0.1458}{1+SA}$ T (1.09(1–10 -3 T)[O 2 ]e– $\frac{1.275x10^4}{T}$ + 7.91x10 -6 e– $\frac{7.39x10^3}{T}$).

36 MATERIALS SCIENCE↗
DOE OSTI2020

Performance of U 3 Si 2 in an LWR following a cladding breach during normal operation

High-density fuels have been proposed as a possible replacement for uranium-dioxide as a fuel for light water reactors (LWRs) due to their increased loading of fissionable material. The objectives of this proposal are (1) to increase reactor cycle length and reactor power, and (2) to offset any neutronic penalty associated with advanced cladding systems. Of the high-density fuels under consideration, there is particular interest in triuranium disilicide (U 3 Si 2 ) due to its increased metal density and favorable thermal properties as compared to UO 2 . However, there are concerns regarding the chemical compatibility of U 3 Si 2 with water and steam as used for LWR coolant.This paper summarizes research on fuel-coolant chemical compatibility for UO 2 LWR fuel during a cladding breach and highlights that generally, because of its chemical inertness, UO 2 -coolant reactions are of little consequence to reactor operation. However, the volumetric expansion associated with the reaction of UO 2 and oxygen is a concern for possible conditions encountered during air ingress of dry storage. These same concerns arise for U 3 Si 2 , which exhibits greater volumetric expansion than UO 2 when exposed to water or steam. These reactions ultimately result in increased fuel volume that the cladding must accommodate, as well as additional heat generated as the fuel reacts.The BISON fuel performance code was used to perform a comparative analysis on the behavior of UO 2 and U 3 Si 2 under normal operation. Silicide fuel simulations were then extended to demonstrate how varying thermodynamic and chemical kinetics influence fuel expansion and subsequent cladding performance during a cladding breach. These simulations were further extended to a 3D subsection of a fuel rod to demonstrate the characteristics of the resulting cladding crack.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗