346 results on '"Terrani, A"'
Search Results
2. Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix
- Author
-
Petrie, Christian M., primary, Linton, Kory D., additional, Vasudevamurthy, Gokul, additional, Schappel, Danny, additional, Seibert, Rachel L., additional, Carpenter, David, additional, Nelson, Andrew T., additional, and Terrani, Kurt A., additional
- Published
- 2023
- Full Text
- View/download PDF
3. Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment
- Author
-
Doyle, Peter, primary, Stuckert, Juri, additional, Grosse, Mirco, additional, Steinbrück, Martin, additional, Nelson, Andrew T., additional, Harp, Jason, additional, and Terrani, Kurt, additional
- Published
- 2023
- Full Text
- View/download PDF
4. A correlation-based approach for evaluating mechanical properties of nuclear fuel cladding tubes
- Author
-
Gussev, M.N., primary, Garrison, B., additional, Massey, C., additional, Coq, A. Le, additional, Linton, K., additional, and Terrani, K.A., additional
- Published
- 2023
- Full Text
- View/download PDF
5. Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment
- Author
-
Peter Doyle, Juri Stuckert, Mirco Grosse, Martin Steinbrück, Andrew T. Nelson, Jason Harp, and Kurt Terrani
- Subjects
Nuclear and High Energy Physics ,Nuclear Energy and Engineering ,General Materials Science - Published
- 2023
- Full Text
- View/download PDF
6. Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix
- Author
-
Christian M. Petrie, Kory D. Linton, Gokul Vasudevamurthy, Danny Schappel, Rachel L. Seibert, David Carpenter, Andrew T. Nelson, and Kurt A. Terrani
- Subjects
Nuclear and High Energy Physics ,Nuclear Energy and Engineering ,General Materials Science - Published
- 2023
- Full Text
- View/download PDF
7. Extensive nanoprecipitate morphology transformation in a nanostructured ferritic alloy due to extreme thermomechanical processing
- Author
-
Kurt A. Terrani, David T. Hoelzer, Steven J. Zinkle, Kinga A. Unocic, Baptiste Gault, Caleb P. Massey, Philip D. Edmondson, and Yury N. Osetskiy
- Subjects
010302 applied physics ,Materials science ,Morphology (linguistics) ,Polymers and Plastics ,Precipitation (chemistry) ,Metals and Alloys ,02 engineering and technology ,Atom probe ,021001 nanoscience & nanotechnology ,01 natural sciences ,Electronic, Optical and Magnetic Materials ,law.invention ,Chemical engineering ,law ,0103 physical sciences ,Volume fraction ,Ceramics and Composites ,Thermomechanical processing ,Dislocation ,Elongation ,0210 nano-technology ,Dissolution - Abstract
Nano-oxide precipitates in a modern nanostructured ferritic alloy were investigated after extreme thermomechanical processing into a thin-walled tube geometry. It was found that the morphology of the precipitates changed from spherical to rod-shaped, with some increasing to aspect ratios of up to 9, despite the precipitate volume fraction (0.3%) and number density (> 1023 m−3) of precipitates remaining unchanged. High-resolution electron microscopy showed that the precipitates likely remained coherent with the Fe-matrix, while atom probe tomography confirmed that the precipitate compositions remained unaffected by the transformation. The morphological change was attributed to the shearable nature of the (Y,Ti,O)-rich precipitates, indicating they should be considered as “soft” obstacles to dislocation motion. The elongation was most pronounced in larger (>5 nm) precipitates, which may be caused by preferential dissolution of the smallest (1–3 nm) precipitates followed by the competition between re-precipitation and solute diffusion to larger precipitates during recovery heat treatments.
- Published
- 2020
- Full Text
- View/download PDF
8. A correlation-based approach for evaluating mechanical properties of nuclear fuel cladding tubes
- Author
-
M.N. Gussev, B. Garrison, C. Massey, A. Le Coq, K. Linton, and K.A. Terrani
- Subjects
Nuclear and High Energy Physics ,Nuclear Energy and Engineering ,General Materials Science - Published
- 2023
- Full Text
- View/download PDF
9. Characterization of radiation damage in 3D printed SiC
- Author
-
Lach, Timothy G., primary, Le Coq, Annabelle G., additional, Linton, Kory D., additional, Terrani, Kurt A., additional, and Byun, Thak Sang, additional
- Published
- 2022
- Full Text
- View/download PDF
10. In-situ X-ray Computed Tomography Analysis of Fracture Mechanisms in Ultrasonic Additively Manufactured Al-6061 Alloy
- Author
-
Cakmak, Ercan, primary, Gussev, Maxim N., additional, Watkins, Thomas R., additional, Arregui-Mena, David J., additional, and Terrani, Kurt A., additional
- Published
- 2021
- Full Text
- View/download PDF
11. Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors
- Author
-
Kurt A. Terrani, Jeffrey J. Powers, Nathan M George, Ryan Sweet, G. Ivan Maldonado, Brian D. Wirth, and Andrew Worrall
- Subjects
Hybrid fuel ,Materials science ,020209 energy ,Alloy ,Zirconium alloy ,02 engineering and technology ,engineering.material ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,Bundle ,Boiling ,0103 physical sciences ,Pellet ,0202 electrical engineering, electronic engineering, information engineering ,engineering ,Boiling water reactor ,Composite material ,Cycle length - Abstract
The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory.
- Published
- 2019
- Full Text
- View/download PDF
12. Stability of a model Fe-14Cr nanostructured ferritic alloy after long-term thermal creep
- Author
-
Steven J. Zinkle, Anoop Kini, Kurt A. Terrani, Caleb P. Massey, David T. Hoelzer, Philip D. Edmondson, and Baptiste Gault
- Subjects
010302 applied physics ,Work (thermodynamics) ,Materials science ,Mechanical Engineering ,Metallurgy ,Metals and Alloys ,02 engineering and technology ,Atom probe ,Nuclear reactor ,021001 nanoscience & nanotechnology ,Condensed Matter Physics ,Microstructure ,01 natural sciences ,Grain size ,law.invention ,Creep ,Mechanics of Materials ,law ,0103 physical sciences ,General Materials Science ,0210 nano-technology ,Porosity ,Microvoid coalescence - Abstract
The successful development of advanced materials such as nanostructured ferritic alloys (NFA) for next generation nuclear reactor concepts and ultra-supercritical steam power plants requires information on long term thermal creep. To support this initiative, this work examines the NFA MA957 (Fe-14Cr-1.0Ti-0.3Mo + 0.3Y2O3 in wt%) crept to 61,251 h at 825 °C and 70 MPa, the longest creep test available to date for this material. Using atom probe tomography and electron microscopy, it is shown that the grain size and nanoprecipitate size/composition are unaffected following this 7 yr creep test, although significant porosity is noted throughout the microstructure attributed to microvoid coalescence and growth.
- Published
- 2019
- Full Text
- View/download PDF
13. Elastic moduli reduction in SiC-SiC tubular specimen after high heat flux neutron irradiation measured by resonant ultrasound spectroscopy
- Author
-
Takaaki Koyanagi, Yutai Katoh, Christian P. Deck, Christian M. Petrie, Gyanender Singh, Kurt A. Terrani, and José David Arregui-Mena
- Subjects
Resonant ultrasound spectroscopy ,Nuclear and High Energy Physics ,Materials science ,Composite number ,02 engineering and technology ,021001 nanoscience & nanotechnology ,Orthotropic material ,01 natural sciences ,010305 fluids & plasmas ,Stress (mechanics) ,Shear (sheet metal) ,Nuclear Energy and Engineering ,Heat flux ,0103 physical sciences ,General Materials Science ,Irradiation ,Composite material ,0210 nano-technology ,Elastic modulus - Abstract
The initial results of a post-irradiation examination study conducted on a SiC-SiC tubular specimen irradiated under a high radial heat flux are presented herein. The elastic properties of the specimen were evaluated before and after the irradiation using the resonant ultrasound spectroscopy (RUS) technique. The composite tubular specimen was considered as an orthotropic elastic with nine elastic constants (Young's moduli, shear moduli and Poisson's ratios—three components of each) for representing its full elastic deformation behavior. All the elastic moduli decreased after irradiation; the reduction was as high as 35% in one of the moduli. The significant decrease in the moduli indicates the presence of microcracks. The results from a computational study show significant stress development in the specimen due to irradiation, primarily caused by differential swelling across the thickness of the specimen. The evaluated stresses exceed the proportional limit stress of the material, indicating the likelihood of matrix microcracking, and thus corroborating the results obtained from RUS. X-ray Computed Tomography (XCT) study confirmed the presence of cracks in the irradiated specimen. These cracks occurred at the inner region of the specimen and propagated in axial and hoop directions. These XCT results are in agreement with the RUS results and stress distribution results from the computational study.
- Published
- 2019
- Full Text
- View/download PDF
14. Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration
- Author
-
Baptiste Gault, Rachel Seibert, Anoop Kini, Philip D. Edmondson, Kurt A. Terrani, Caleb P. Massey, David T. Hoelzer, and Steven J. Zinkle
- Subjects
Nuclear and High Energy Physics ,education.field_of_study ,Materials science ,Population ,Metallurgy ,Alloy ,02 engineering and technology ,Atom probe ,engineering.material ,021001 nanoscience & nanotechnology ,Microstructure ,01 natural sciences ,Grain size ,010305 fluids & plasmas ,law.invention ,Nanoclusters ,Solid solution strengthening ,Nuclear Energy and Engineering ,law ,0103 physical sciences ,engineering ,General Materials Science ,Grain boundary ,0210 nano-technology ,education - Abstract
Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC ( O ak Ridge F ast R eactor A dvanced Fuel C ladding) with solute additions of Mo, Ti, and Nb has been developed to provide the desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850 °C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 1023 m-3) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. The ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).
- Published
- 2019
- Full Text
- View/download PDF
15. Investigation of sol-gel feedstock additions and process variables on the density and microstructure of UN microspheres
- Author
-
Kurt A. Terrani, Rachel Seibert, Rodney D. Hunt, Chinthaka M. Silva, Tyler J. Gerczak, Grant W. Helmreich, Jacob W. McMurray, and T.J. Reif
- Subjects
Nuclear and High Energy Physics ,Materials science ,Pellets ,Sintering ,Raw material ,Microstructure ,Chemical state ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,Chemical engineering ,chemistry ,visual_art ,visual_art.visual_art_medium ,Silicon carbide ,General Materials Science ,Ceramic ,Sol-gel - Abstract
The kernel of fully ceramic microencapsulated (FCM) fuel requires a material with high fissile density. For that reason, among others, UN is the appropriate chemical state for low enriched U fuel. The UN kernels are spheres ∼800 μm in diameter made using a sol-gel process. The effect of additives on the chemistry and density of the UN microspheres are investigated in this work. Gadolinium nitrate hexahydrate, Gd2O3, B, SiO2 and SiC were incorporated into the sol-gel broth in varying concentrations. It was found that Gd can serve as both a sintering aid and burnable poison when added to the sol-gel broth as a Gd(NO3)36H2O. However, even with the increased theoretical density of the UN microspheres, the U density was still too low for the FCM design that replaces UO2 pellets in a commercial light water reactor. Silicon carbide and B were also successfully added but produced a lower density final product. Other sol-gel processing variables were investigated. The pour density of the sol-gel feedstock was found to influence the final converted UN kernel density.
- Published
- 2019
- Full Text
- View/download PDF
16. Fabrication of UO2-Mo composite fuel with enhanced thermal conductivity from sol-gel feedstock
- Author
-
Sarah Finkeldei, James O. Kiggans, Rodney D. Hunt, Andrew T. Nelson, and Kurt A. Terrani
- Subjects
Nuclear and High Energy Physics ,Materials science ,Fabrication ,Uranium dioxide ,Composite number ,chemistry.chemical_element ,02 engineering and technology ,Raw material ,021001 nanoscience & nanotechnology ,01 natural sciences ,010305 fluids & plasmas ,chemistry.chemical_compound ,Thermal conductivity ,Nuclear Energy and Engineering ,chemistry ,Chemical engineering ,Molybdenum ,0103 physical sciences ,Metal powder ,General Materials Science ,0210 nano-technology ,Sol-gel - Abstract
Fabrication of monolithic UO2 and UO2-Mo composites from Mo metal powder feedstock and UO3 spheres derived from internal gelation was explored. The addition of 10 vol% Mo as a secondary phase to UO2 led to an increase of up to 30% in thermal conductivity compared to syntheses in which only pure UO2 was used. Using coarse and fine Mo metal powder feedstock led to different microstructures, influencing the thermal conductivity of the resulting composites.
- Published
- 2019
- Full Text
- View/download PDF
17. An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading
- Author
-
M. Nedim Cinbiz, Kurt A. Terrani, Kory Linton, and Maxim N. Gussev
- Subjects
Digital image correlation ,Materials science ,Nuclear fuel ,020209 energy ,Acoustics ,Internal pressure ,Video camera ,02 engineering and technology ,Tube specimen ,Cladding (fiber optics) ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Single camera ,Nuclear Energy and Engineering ,law ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Telecentric lens - Abstract
An advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.
- Published
- 2019
- Full Text
- View/download PDF
18. Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel
- Author
-
Soon K. Lee, Kurt A. Terrani, Maolong Liu, Nicholas R. Brown, Heng Ban, Edward D. Blandford, Colby Jensen, and Youho Lee
- Subjects
Fluid Flow and Transfer Processes ,Cladding (metalworking) ,Materials science ,Atmospheric pressure ,Critical heat flux ,Mechanical Engineering ,Mass flow ,Zirconium alloy ,02 engineering and technology ,021001 nanoscience & nanotechnology ,Condensed Matter Physics ,01 natural sciences ,010305 fluids & plasmas ,Boiling ,0103 physical sciences ,Composite material ,0210 nano-technology ,Inconel ,Nucleate boiling - Abstract
Steady and transient (with a heating rate of 685 °C/s) internal-flow CHF (Critical Heat Flux) experiments were conducted under atmospheric pressure at a fixed inlet temperature (40 °C or 60 °C) and mass flow (300 kg/m2 s) on Fe-13Cr-6Al alloy, Inconel 600 and Zircaloy-4 tube samples. Multiple experiments were repeated on the same specimen to investigate the effect of surface characteristic changes (i.e., roughness, wettability, and oxide scale morphology) on the occurrence of CHF. Despite notable changes of wettability, roughness, and oxide layer characteristics on samples that had already been subjected to CHF, measured flow CHF remained unchanged throughout repeated experiments for tested materials. This demonstrates that the surface effects on flow CHF are limited in the test conditions. In the steady-state flow boiling condition, Fe-13Cr-6Al alloy demonstrated a 22% and 14% increase in CHF compared to Zircaloy-4 and Inconel 600, respectively. Compared to the 2006 Groeneveld CHF lookup table, Fe-13Cr-6Al alloy gives a 13% increase in the tested flow boiling condition. Material properties are considered primarily responsible for the observed CHF differences among the tested materials. The surface thermal economy parameter ( ρ c p 3 / 2 k ) is proposed as an explanation for the observed CHF differences; this parameter is related to material’s ability to avoid an irreversible dry spot formation. The apparent disagreement of Zircaloy-4 CHF with both the look up table predictions and Inconel 600 shows the limitation of departure of nucleate boiling (DNB) evaluations that do not consider cladding materials. The transient Fe-13Cr-6Al CHF is 39% and 23% higher than the lookup table prediction and the steady-state condition experimental result, respectively.
- Published
- 2019
- Full Text
- View/download PDF
19. Fully Ceramic Microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Sensitivity of reactor behavior during design basis accidents to fuel properties and the potential impact of the SiC defect annealing process
- Author
-
Yutai Katoh, Nicholas R. Brown, Gerhard Strydom, Kurt A. Terrani, Takaaki Koyanagi, and Cihang Lu
- Subjects
Nuclear and High Energy Physics ,Potential impact ,Materials science ,Annealing (metallurgy) ,020209 energy ,Mechanical Engineering ,Nuclear engineering ,Control rod ,02 engineering and technology ,01 natural sciences ,010305 fluids & plasmas ,Thermal conductivity ,Nuclear Energy and Engineering ,visual_art ,0103 physical sciences ,Thermal ,0202 electrical engineering, electronic engineering, information engineering ,visual_art.visual_art_medium ,General Materials Science ,Ceramic ,Irradiation ,Wigner effect ,Safety, Risk, Reliability and Quality ,Waste Management and Disposal - Abstract
We conducted reactor performance calculations to assess the potential design basis accident performance of HTGR fuel designs. Three Fully Ceramic Microencapsulated (FCM) fueled HTGR designs were developed in a previous work (Lu et al., 2018). The maximum fuel temperature in the cores fueled by these three FCM fuels was predicted to be higher than that in the reference 350-MWt mHTGR core in both normal operating conditions and during representative design basis accidents (Lu and Brown, 2019). To better understand the potential safety margins in mHTGR design basis accidents, we performed thermal-hydraulics sensitivity studies to investigate how maximum fuel temperature varies considering various parameters, e.g. thermal properties, within the ranges corresponding to the differences between the FCM-fueled prismatic mHTGR cores and the reference core with conventional fuel compacts. We found that the difference in the steady-state axial power distribution contributed the most to the difference in the maximum fuel temperature, in both normal operation and design basis accidents. Experimental data suggested that the annealing process of irradiation defects in SiC would be rapid at mHTGR relevant fuel temperatures. The bounding potential impact of the SiC annealing on the maximum fuel temperature was analyzed considering both the thermal conductivity recovery and the Wigner energy release due to the annealing of SiC. We found that the defect annealing process in SiC would at most increase the peak maximum fuel temperature of an FCM-fueled core by 40 K in loss of forced cooling accidents and by 10 K in a control rod withdrawal accident. Additional experiments on the SiC defect annealing kinetics and Wigner energy release in more relevant conditions are needed.
- Published
- 2019
- Full Text
- View/download PDF
20. Post irradiation examination of nanoprecipitate stability and α′ precipitation in an oxide dispersion strengthened Fe-12Cr-5Al alloy
- Author
-
Sebastien N. Dryepondt, Kevin G. Field, Caleb P. Massey, Steven J. Zinkle, David T. Hoelzer, Philip D. Edmondson, and Kurt A. Terrani
- Subjects
Materials science ,Alloy ,Analytical chemistry ,Oxide ,02 engineering and technology ,Atom probe ,engineering.material ,01 natural sciences ,law.invention ,chemistry.chemical_compound ,law ,0103 physical sciences ,General Materials Science ,Irradiation ,Dissolution ,010302 applied physics ,Number density ,Precipitation (chemistry) ,Mechanical Engineering ,Metals and Alloys ,021001 nanoscience & nanotechnology ,Condensed Matter Physics ,chemistry ,Mechanics of Materials ,engineering ,Post Irradiation Examination ,0210 nano-technology - Abstract
Oxide dispersion strengthened (ODS) FeCrAl alloy (Fe-12Cr-5Al wt%) was neutron irradiated to 1.8 displacements per atom (dpa) at 215, 357, and 557 °C to investigate (Y,Al,O) nano-oxide stability and Cr-rich α′ precipitate formation using atom probe tomography. The nano-oxide sizes remain unchanged after irradiation, but the compositions are dependent on irradiation temperature due to competition between ballistic dissolution and precipitate reformation. Cr-rich clusters observed after 357 °C irradiation have lower Cr contents (61 at.%) than equilibrium α′, with a lower number density than wrought FeCrAl alloys. This suggests a suppression of α′ precipitation due to the high sink strength of ODS FeCrAl.
- Published
- 2019
- Full Text
- View/download PDF
21. Production and characterization of TRISO fuel particles with multilayered SiC
- Author
-
Mehdi Balooch, Rachel Seibert, Brian C. Jolly, Kurt A. Terrani, and Daniel Schappel
- Subjects
Nuclear and High Energy Physics ,Materials science ,Composite number ,Fracture mechanics ,02 engineering and technology ,engineering.material ,Nanoindentation ,021001 nanoscience & nanotechnology ,01 natural sciences ,010305 fluids & plasmas ,Stress (mechanics) ,Cracking ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Coating ,0103 physical sciences ,Silicon carbide ,engineering ,General Materials Science ,Composite material ,0210 nano-technology ,Layer (electronics) - Abstract
Three distinct composite architectures of silicon carbide (SiC) and pyrocarbon (PyC) were incorporated into the SiC coating layer of tristructural-isotropic (TRISO) nuclear fuel particles. The composite architectures are meant to increase the resistance of SiC coating layer to cracking and fission product attack during operation and accident scenarios. All composite layers were produced using the existing fluidized bed chemical vapor deposition apparatus that is used for production of TRISO fuel particles without modifications. Detailed characterization of the composite microstructure was carried out via optical and electron microscopy. Nano-indentation examination confirms that mechanical properties of the SiC phase was not affected in the composite architectures, however, the resistance to crack propagation in this coating layer was greatly increased in all cases when compared to the reference monolithic coating layer. The stress required to debond the SiC-inner PyC interface in the reference TRISO particles was determined to be ∼1 GPa using micropillar compression technique. The high strength may explain the ease of crack propagation from the inner PyC to SiC in the reference design. In the composite architectures, the means of crack deflection were effectively incorporated at this interface. Finite element analysis of stress evolution in the fuel particles during normal operation with the reference and composite SiC coating layer architectures did not show any significant differences between the variants.
- Published
- 2019
- Full Text
- View/download PDF
22. Stored energy release in neutron irradiated silicon carbide
- Author
-
Yutai Katoh, Kurt A. Terrani, Lance Lewis Snead, and Takaaki Koyanagi
- Subjects
Nuclear and High Energy Physics ,Materials science ,Analytical chemistry ,02 engineering and technology ,021001 nanoscience & nanotechnology ,Microstructure ,01 natural sciences ,Fluence ,010305 fluids & plasmas ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Nuclear graphite ,0103 physical sciences ,Silicon carbide ,General Materials Science ,Irradiation ,Graphite ,Crystallite ,0210 nano-technology ,Single crystal - Abstract
The purpose of this investigation is to experimentally quantify the stored energy release upon thermal annealing of previously irradiated high-purity silicon carbide (SiC.) Samples of highly-faulted polycrystalline CVD β-SiC and single crystal 6H SiC were irradiated in a mixed spectrum fission reactor near 60 °C in a fluence range from 5 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.05–20 dpa, in order to quantify the stored energy release and correlate the release to the observed microscopic swelling, lattice dilation, and microstructure as observed through TEM. Within the fluence of this study the crystalline material was observed to swell to a remarkable extent, achieving 8.13% dilation, and then cross a threshold dose for amorphization at approximately 1 × 1025 n/m2 (E > 0.1 MeV) Once amorphized the material attains an as-amorphized swelling of 11.7% at this irradiation condition. Coincident with the extraordinary swelling obtained for the crystalline SiC, an equally impressive stored energy release of greater than 2500 J/g at the critical threshold for amorphization is inferred. As expected, following amorphization the stored energy in the structure diminishes, measured to be approximately 590 J/g. Generally, the findings of stored energy are consistent with existing theory, though the amount of stored energy given the large observed crystalline strain is remarkable. The overall conclusion of this work finds comparable stored energy in SiC to that of nuclear graphite, and similar to graphite, a stored energy release in excess of its specific heat in some irradiation conditions.
- Published
- 2019
- Full Text
- View/download PDF
23. Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents
- Author
-
Nicholas R. Brown, Gyanender Singh, M. Nedim Cinbiz, Kurt A. Terrani, Takaaki Koyanagi, and Yutai Katoh
- Subjects
Nuclear and High Energy Physics ,Materials science ,Nuclear fuel ,Composite number ,02 engineering and technology ,Strain rate ,021001 nanoscience & nanotechnology ,Cladding (fiber optics) ,01 natural sciences ,Thermal expansion ,010305 fluids & plasmas ,law.invention ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Magazine ,law ,0103 physical sciences ,Pellet ,Silicon carbide ,General Materials Science ,Composite material ,0210 nano-technology - Abstract
The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms. The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. The failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.
- Published
- 2019
- Full Text
- View/download PDF
24. Deformation analysis of SiC-SiC channel box for BWR applications
- Author
-
Jacob P. Gorton, Brian D. Wirth, Gyanender Singh, Daniel Schappel, Kurt A. Terrani, Yutai Katoh, and Nicholas R. Brown
- Subjects
Nuclear and High Energy Physics ,Neutron transport ,Materials science ,02 engineering and technology ,Mechanics ,Nuclear reactor ,021001 nanoscience & nanotechnology ,Cladding (fiber optics) ,01 natural sciences ,Finite element method ,010305 fluids & plasmas ,law.invention ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Neutron flux ,law ,0103 physical sciences ,Silicon carbide ,General Materials Science ,Neutron ,Light-water reactor ,0210 nano-technology - Abstract
Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.
- Published
- 2019
- Full Text
- View/download PDF
25. Delaying Reproductive Aging by Ovarian Tissue Cryopreservation and Transplantation: Is it Prime Time?
- Author
-
Oktay, Kutluk H., primary, Marin, Loris, additional, Petrikovsky, Boris, additional, Terrani, Michael, additional, and Babayev, Samir N., additional
- Published
- 2021
- Full Text
- View/download PDF
26. Irradiation stability and thermomechanical properties of 3D-printed SiC
- Author
-
Terrani, K.A., primary, Lach, T., additional, Wang, H., additional, Coq, A. Le, additional, Linton, K., additional, Petrie, C., additional, Koyanagi, T., additional, and Byun, T.S., additional
- Published
- 2021
- Full Text
- View/download PDF
27. Embedded sensors in additively manufactured silicon carbide
- Author
-
Petrie, Christian M., primary, Schrell, Adrian M., additional, Leonard, Donovan N., additional, Yang, Ying, additional, Jolly, Brian C., additional, and Terrani, Kurt A., additional
- Published
- 2021
- Full Text
- View/download PDF
28. Influence of neutron irradiation on Al-6061 alloy produced via ultrasonic additive manufacturing
- Author
-
Gussev, M.N., primary, Sridharan, N., additional, Babu, S.S., additional, and Terrani, K.A., additional
- Published
- 2021
- Full Text
- View/download PDF
29. Corrigendum to “Integral LOCA fragmentation test on high-burnup fuel” [Nucl. Eng. Design (2020) 367 110811, ISN 0029-5493]
- Author
-
Capps, Nathan, primary, Yan, Yong, additional, Raftery, Alicia, additional, Burns, Zachary, additional, Smith, Tyler, additional, Terrani, Kurt, additional, Yueh, Ken, additional, Bales, Michelle, additional, and Linton, Kory, additional
- Published
- 2021
- Full Text
- View/download PDF
30. Thermomechanical properties and microstructures of yttrium hydride
- Author
-
Hu, Xunxiang, primary and Terrani, Kurt A., additional
- Published
- 2021
- Full Text
- View/download PDF
31. Deuterium permeation and retention in 316L Stainless Steel Manufactured by Laser Powder Bed Fusion
- Author
-
Hu, Xunxiang, primary, Lach, Timothy G., additional, and Terrani, Kurt A., additional
- Published
- 2021
- Full Text
- View/download PDF
32. Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation
- Author
-
Byun, T.S., primary, Garrison, B.E., additional, McAlister, M.R., additional, Chen, X., additional, Gussev, M.N., additional, Lach, T.G., additional, Coq, A. Le, additional, Linton, K., additional, Joslin, C.B., additional, Carver, J.K., additional, List, F.A., additional, Dehoff, R.R., additional, and Terrani, K.A., additional
- Published
- 2021
- Full Text
- View/download PDF
33. Microstructure and mechanical properties of high Mn-containing ferritic-martensitic alloys exposed to cyclical thermal treatment
- Author
-
Zhong, Weicheng, primary, Yang, Ying, additional, Field, Kevin G., additional, Sridharan, Niyanth, additional, Terrani, Kurt, additional, and Tan, Lizhen, additional
- Published
- 2021
- Full Text
- View/download PDF
34. A Critical Review of High Burnup Fuel Fragmentation, Relocation, and Dispersal under Loss-Of-Coolant Accident Conditions
- Author
-
Capps, Nathan, primary, Jensen, Colby, additional, Cappia, Fabiola, additional, Harp, Jason, additional, Terrani, Kurt, additional, Woolstenhulme, Nicolas, additional, and Wachs, Daniel, additional
- Published
- 2021
- Full Text
- View/download PDF
35. Architecture and properties of TCR fuel form
- Author
-
Terrani, K.A., primary, Jolly, B.C., additional, Trammell, M.P., additional, Vasudevamurthy, G., additional, Schappel, D., additional, Ade, B., additional, Helmreich, G.W., additional, Wang, H., additional, Rossy, A. Marquiz, additional, Betzler, B.R., additional, and Nelson, A.T., additional
- Published
- 2021
- Full Text
- View/download PDF
36. Characterization of radiation damage in 3D printed SiC
- Author
-
Timothy G. Lach, Annabelle G. Le Coq, Kory D. Linton, Kurt A. Terrani, and Thak Sang Byun
- Subjects
Nuclear and High Energy Physics ,Nuclear Energy and Engineering ,General Materials Science - Published
- 2022
- Full Text
- View/download PDF
37. Influence of mechanical alloying and extrusion conditions on the microstructure and tensile properties of Low-Cr ODS FeCrAl alloys
- Author
-
Sebastien N. Dryepondt, Caleb P. Massey, Kurt A. Terrani, Steven J. Zinkle, and Philip D. Edmondson
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Precipitation (chemistry) ,Metallurgy ,02 engineering and technology ,021001 nanoscience & nanotechnology ,Microstructure ,01 natural sciences ,Nuclear Energy and Engineering ,0103 physical sciences ,Ultimate tensile strength ,General Materials Science ,Extrusion ,0210 nano-technology ,Strengthening mechanisms of materials - Published
- 2018
- Full Text
- View/download PDF
38. Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys
- Author
-
Nicholas R. Brown, Youho Lee, Amir F. Ali, Colby Jensen, Edward D. Blandford, Jacob P. Gorton, and Kurt A. Terrani
- Subjects
Nuclear and High Energy Physics ,Critical heat flux ,020209 energy ,Mechanical Engineering ,Zirconium alloy ,Pressurized water reactor ,02 engineering and technology ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Nuclear Energy and Engineering ,Heat flux ,law ,Boiling ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Surface roughness ,General Materials Science ,Light-water reactor ,Composite material ,Safety, Risk, Reliability and Quality ,Waste Management and Disposal ,Nucleate boiling - Abstract
Surface wettability analysis, including measurements of static (θ) advance (θ A ), and receding (θ r ) contact angles, and surface roughness, Ra, are effective parameters used in the literature to predict changes in the pool boiling Critical Heat Flux (CHF). The CHF is an important aspect of the thermal hydraulic performance that needs to be investigated for new Accident Tolerant Fuel (ATF) cladding materials, such as iron-chromium-aluminum alloys (FeCrAl). Surface wettability of FeCrAl samples oxidized under different simulated Light Water Reactor (LWR) water chemistry conditions was measured. These measurements were compared to as-machined and oxidized Zircaloy-4 (Zirc-4), the reference cladding material in LWRs, under the same conditions. Theoretical models were used to predict the pool boiling CHF using surface wettability measurements. Pool boiling experiments were conducted using the same samples to measure the CHF. The measured and predicted CHF data were compared for model validation. The obtained results showed no significant difference in the measured static contact angle and hence the predicted CHF between the as-machined samples of FeCrAl, 310 SS, and Zirc-4. The contact angles (static, advance, and receding angles) for corroded FeCrAl samples under different LWR water chemistry conditions are lower, and the measured surface roughness values are higher than Zirc-4 corroded under the same conditions. Existing models in the literature predicted higher pool boiling CHF of corroded FeCrAl compared to 310 stainless steel (SS), and Zirc-4. The measured pool boiling CHF for oxidized FeCrAl samples was higher than Zirc-4 samples. The measured and predicted CHF values are in good agreement. The CHF and the Departure from Nucleate Boiling Ratio (DNBR) were calculated using the Consortium for Advanced Simulation of Light Water Reactors (CASL) subchannel code COBRA-TF (CTF) for a one-eighth model of a 17 × 17 Pressurized Water Reactor (PWR) fuel assembly. The inlet conditions are consistent with typical PWR values except for the power, which was set 50% higher than is typical to represent possible accident conditions more accurately. The calculated distributions for the CHF and DNBR for oxidized FeCrAl showed significantly higher values throughout the fuel assembly octant compared to as machined Zirc-4 and as machined FeCrAl. These preliminary results show that oxidized FeCrAl may be able to withstand the proposed accident conditions without leading to a boiling crisis.
- Published
- 2018
- Full Text
- View/download PDF
39. Restructuring in high burnup UO2 studied using modern electron microscopy
- Author
-
Kurt A. Terrani, Philip D. Edmondson, Charles A. Baldwin, Tyler J. Gerczak, and Chad M. Parish
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Structure formation ,Materials science ,Nuclear engineering ,02 engineering and technology ,021001 nanoscience & nanotechnology ,Microstructure ,01 natural sciences ,Grain size ,Nuclear Energy and Engineering ,Creep ,0103 physical sciences ,Pellet ,General Materials Science ,Grain boundary ,Light-water reactor ,0210 nano-technology ,Burnup - Abstract
Modern electron microscopy techniques were used to conduct a thorough study of an irradiated urania fuel pellet microstructure to attempt at an understanding of high burnup structure formation in this material. The fuel was irradiated at low power to high burnups in a light water reactor, proving ideal for this purpose. Examination of grain size and orientation with strict spatial selectivity across the fuel pellet radius allowed for capturing the progression of the restructuring process, from its onset to full completion. Based on this information, the polygonization mechanism was shown to be responsible for restructuring, involving formation of low-angle grain boundaries with their initiation occurring at the original high-angle grain boundaries of the as-fabricated pellet and at the gas bubble-matrix interfaces. The low-angle character of boundaries between the subdivided grains disappeared in the fully developed high burnup structure, likely due to creep deformation in the pellet.
- Published
- 2018
- Full Text
- View/download PDF
40. Thermal expansion behavior of δ-zirconium hydrides: Comparison of δ hydride powder and platelets
- Author
-
Kurt A. Terrani, Xunxiang Hu, and M. Nedim Cinbiz
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Zirconium ,Materials science ,Hydride ,Precipitation (chemistry) ,Analytical chemistry ,chemistry.chemical_element ,02 engineering and technology ,Zirconium hydride ,021001 nanoscience & nanotechnology ,01 natural sciences ,Synchrotron ,Thermal expansion ,law.invention ,Lattice constant ,Nuclear Energy and Engineering ,chemistry ,law ,0103 physical sciences ,General Materials Science ,0210 nano-technology ,Dissolution - Abstract
The thermal expansion coefficient of δ hydrides and the evolution of the d-spacing of δ hydride platelets in Zircaloy-4 during heat treatments were investigated by conducting synchrotron x-ray diffraction experiments. Identical experiments enabled a direct comparison of the d-spacing of δ zirconium hydride platelets with the d-spacing of the powder δ hydrides. By analyzing the experimental data of this study and the data available in the literature, the thermal expansion coefficient of pure δ hydrides was determined as 14.1 10−6 °C−1. A direct comparison of the d-spacings of δ hydride platelets in CWSR Zry-4 sheet and the powder δ hydride samples showed an evolution of temperature-dependent strains within the δ hydride precipitates in which the strain components normal to the platelet edges exceed that normal to the platelet face during cooling/precipitation but not during heating/dissolution above approx. 200 °C.
- Published
- 2018
- Full Text
- View/download PDF
41. Young's modulus evaluation of high burnup structure in UO2 with nanometer resolution
- Author
-
Quinlan B. Smith, Kurt A. Terrani, Mehdi Balooch, and Joseph R. Burns
- Subjects
Cladding (metalworking) ,Nuclear and High Energy Physics ,Materials science ,Modulus ,Young's modulus ,02 engineering and technology ,Nanoindentation ,021001 nanoscience & nanotechnology ,Microstructure ,01 natural sciences ,010305 fluids & plasmas ,symbols.namesake ,Nuclear Energy and Engineering ,0103 physical sciences ,symbols ,General Materials Science ,Composite material ,0210 nano-technology ,Single crystal ,Elastic modulus ,Burnup - Abstract
The mechanical properties of a well-characterized irradiated light water reactor fuel pin with an average burnup of 72 MWd/kgU were investigated by utilizing nano-indentation technique. Young's modulus and hardness are reported as functions of radial positions covering from mid radial position in the urania pellet to high burnup structure, fuel-cladding chemical interaction zone and zirconium-based cladding. By probing microstructure of the high burnup zone, the mechanical properties of sub-grains and the interaction between them were examined at the nanometer scale. Force-displacement curves from each of these individual sub-grains suggest they resemble single crystal UO2. However, these restructured grains are weakly bonded to their neighbors.
- Published
- 2018
- Full Text
- View/download PDF
42. Modeling the performance of TRISO-based fully ceramic matrix (FCM) fuel in an LWR environment using BISON
- Author
-
Jeffrey J. Powers, Lance Lewis Snead, Daniel Schappel, Brian D. Wirth, and Kurt A. Terrani
- Subjects
Nuclear and High Energy Physics ,Materials science ,02 engineering and technology ,Ceramic matrix composite ,01 natural sciences ,010305 fluids & plasmas ,chemistry.chemical_compound ,0103 physical sciences ,Pellet ,Silicon carbide ,General Materials Science ,Pyrolytic carbon ,Ceramic ,Composite material ,Safety, Risk, Reliability and Quality ,Waste Management and Disposal ,Mechanical Engineering ,Fuel type ,021001 nanoscience & nanotechnology ,Coolant ,Nuclear Energy and Engineering ,chemistry ,Particle packing ,visual_art ,visual_art.visual_art_medium ,0210 nano-technology - Abstract
Fully ceramic microencapsulated (FCM) fuel is a proposed fuel type for improved accident performance in LWRs (Light Water Reactors) that involves TRISO (TRistructural-ISOtropic) particles embedded in a nano-powder sintered silicon carbide (SiC) matrix. The TRISO particles contain a spherical fuel kernel ranging from 500 to 800 µm in diameter. The kernel and buffer layer are then coated with three layers, each of which is 30–40 µm thick, composed of dense inner pyrolytic carbon (IPyC), chemically vapor deposited silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. These TRISO particles are then embedded in a fully dense sintered SiC matrix with an expected particle packing fraction of about 35–40% by volume. As is the case for gas reactor applications, the release of radioactivity into the coolant is dependent on the integrity of the silicon carbide layer of the TRISO particles, in addition to the SiC matrix. In this work, we report on fuel performance modeling of TRISO-bearing FCM fuel using the BISON code to simulate the thermo-mechanical behavior of this fuel in a prototypic LWR environment. This paper considers the effects of embedding a TRISO particle in the SiC pellet matrix and includes a discussion of the irradiation-induced dimensional change in the pyrolytic carbon (PyC) layers of the TRISO particle. Additionally, methods were developed to simulate a FCM pellet containing a large number of discrete and independent particles. Future work will report on developing an interface debonding model, a fracture model, and a radionuclide transport model.
- Published
- 2018
- Full Text
- View/download PDF
43. Reactor physics phenomena in additively manufactured control elements for the High Flux Isotope Reactor
- Author
-
David Chandler, Joseph R. Burns, Bojan Petrovic, and Kurt A. Terrani
- Subjects
Physics ,Neutron transport ,Fabrication ,Nuclear engineering ,Absorption cross section ,02 engineering and technology ,Oak Ridge National Laboratory ,Neutron temperature ,020501 mining & metallurgy ,0205 materials engineering ,Nuclear Energy and Engineering ,Heat flux ,Absorption (electromagnetic radiation) ,High Flux Isotope Reactor - Abstract
Additive manufacturing is under investigation as a novel method of fabricating the control elements (CEs) of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory with greater simplicity, eliminating numerous highly complex fabrication steps and thereby offering potential for significant savings in cost, time, and effort. This process yields a unique CE design with lumped absorbers, a departure from traditionally manufactured CEs with uniformly distributed absorbing material. This study undertakes a neutronics analysis of the impact of additively manufactured CEs on the HFIR core physics, seeking preliminary assessment of the feasibility of their practical use. The results of the MCNP transport simulations reveal changes in the HFIR reactor physics arising from geometric and nuclear effects. Absorber lumping in the discrete CEs yields a large volume of unpoisoned material that is not present in the homogeneous design, in turn yielding increases in free thermal flux in the CE absorbing regions and their immediate vicinity. The availability of additional free thermal neutrons in the core yields an increase in fission rate density in the fuel closest to the CEs and a corresponding increase in neutron multiplication on the order of 100 pcm. The absorption behavior exhibited by the discrete CEs is markedly different from the homogeneous CEs due to several competing effects. Self-shielding arising from absorber lumping acts to reduce the effective absorption cross section of the discrete CEs, but this effect is offset by geometric and spectral effects. The operational performance of the discrete CEs is found to be comparable to the homogeneous CEs, with only limited deficiencies in reactivity worth that are expected to be operationally recoverable via limited adjustment of the CE positions and withdrawal rate. On the whole, these results indicate that the discrete CEs perform reasonably similarly to the homogeneous CEs and appear feasible for application in HFIR. The physical phenomena identified in this study provide valuable background for follow-up design studies.
- Published
- 2018
- Full Text
- View/download PDF
44. Evaluation of microstructure stability at the interfaces of Al-6061 welds fabricated using ultrasonic additive manufacturing
- Author
-
Niyanth Sridharan, S. Suresh Babu, Dieter Isheim, Maxim N. Gussev, Kurt A. Terrani, Chad M. Parish, and David N. Seidman
- Subjects
010302 applied physics ,Ultrasonic welding ,Materials science ,Mechanical Engineering ,Alloy ,chemistry.chemical_element ,02 engineering and technology ,engineering.material ,021001 nanoscience & nanotechnology ,Condensed Matter Physics ,Microstructure ,01 natural sciences ,Grain growth ,chemistry ,Mechanics of Materials ,Aluminium ,0103 physical sciences ,Ultimate tensile strength ,engineering ,General Materials Science ,Grain boundary ,Ultrasonic sensor ,Composite material ,0210 nano-technology - Abstract
Ultrasonic additive manufacturing (UAM) is a solid-state additive manufacturing process that uses fundamental principles of ultrasonic welding and sequential layering of tapes to fabricate complex three-dimensional (3-D) components. One of the factors limiting the use of this technology is the poor tensile strength along the z-axis. Recent work has demonstrated the improvement of the z-axis properties after post-processing treatments. The abnormally high stability of the grains at the interface during post-weld heat treatments is, however, not yet well understood. In this work we use multiscale characterization to understand the stability of the grains during post-weld heat treatments. Aluminum alloy (6061) builds, fabricated using ultrasonic additive manufacturing, were post-weld heat treated at 180, 330 and 580 °C. The grains close to the tape interfaces are stable during post-weld heat treatments at high temperatures (i.e., 580 °C). This is in contrast to rapid grain growth that takes place in the bulk. Transmission electron microscopy and atom-probe tomography display a significant enrichment of oxygen and magnesium near the stable interfaces. Based on the detailed characterization, two mechanisms are proposed and evaluated: nonequilibrium nano-dispersed oxides impeding the grain growth due to grain boundary pinning, or grain boundary segregation of magnesium and oxygen reducing the grain boundary energy.
- Published
- 2018
- Full Text
- View/download PDF
45. Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics
- Author
-
Kurt A. Terrani, Briana Hiscox, Nicholas R. Brown, and Cihang Lu
- Subjects
Nuclear fission product ,Control rod ,Nuclear engineering ,02 engineering and technology ,Nuclear reactor ,021001 nanoscience & nanotechnology ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Thermal hydraulics ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,law ,Neutron flux ,0103 physical sciences ,Silicon carbide ,Neutron ,0210 nano-technology ,Burnup - Abstract
Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tr istructural iso tropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar. The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design. The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238 U in the fuel. In one case considered, this will increase the absorption of neutrons in 238 U, and the resultant impact of 239 Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power.
- Published
- 2018
- Full Text
- View/download PDF
46. Accident tolerant fuel cladding development: Promise, status, and challenges
- Author
-
Kurt A. Terrani
- Subjects
Cladding (metalworking) ,Nuclear and High Energy Physics ,Zirconium ,Nuclear engineering ,Iron alloys ,chemistry.chemical_element ,02 engineering and technology ,021001 nanoscience & nanotechnology ,01 natural sciences ,010305 fluids & plasmas ,Carbide ,Surface coating ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Nuclear industry ,0103 physical sciences ,Silicon carbide ,Environmental science ,General Materials Science ,0210 nano-technology ,Energy source - Abstract
The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber–reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.
- Published
- 2018
- Full Text
- View/download PDF
47. Fuel performance simulation of iron-chrome-aluminum (FeCrAl) cladding during steady-state LWR operation
- Author
-
Ryan Sweet, G. I. Maldonado, Nathan M George, Kurt A. Terrani, and Brian D. Wirth
- Subjects
Cladding (metalworking) ,Nuclear and High Energy Physics ,Neutron transport ,Zirconium ,Materials science ,020209 energy ,Mechanical Engineering ,Nuclear engineering ,chemistry.chemical_element ,02 engineering and technology ,01 natural sciences ,Finite element method ,Rod ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,Creep ,chemistry ,Aluminium ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,General Materials Science ,Light-water reactor ,Safety, Risk, Reliability and Quality ,Waste Management and Disposal - Abstract
Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys that exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. Within the development of these alloys, suitability for normal operation must also be demonstrated. This article is focused on modeling the integral thermo-mechanical performance of FeCrAl clad UO 2 fuel during normal reactor operation. Finite element analysis has been performed to assess commercially available FeCrAl alloys (namely Alkrothal 720 and APMT) as a candidate fuel cladding replacement for Zr-alloys, using the MOOSE-based fuel performance code BISON. These simulations identify the effects of the mechanical-stress and irradiation responses of FeCrAl and provide a comparison with Zr-alloys. In comparing these cladding materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (∼4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. These power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. The fuel rod designs and operating conditions used here are based on the Peach Bottom BWR with representative GE-12/14 fuel geometries, and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. Individual sensitivity analyses of the fuel and cladding creep responses were also performed, which indicated the influence of compliance for each material, separately, on the stress state of the fuel cladding. These parametric analyses are performed using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history.
- Published
- 2018
- Full Text
- View/download PDF
48. Influence of hot isostatic pressing on the performance of aluminum alloy fabricated by ultrasonic additive manufacturing
- Author
-
Kurt A. Terrani, S. Suresh Babu, Niyanth Sridharan, Maxim N. Gussev, and Z. Thompson
- Subjects
0209 industrial biotechnology ,Ultrasonic welding ,Materials science ,Mechanical Engineering ,Metallurgy ,Alloy ,Metals and Alloys ,02 engineering and technology ,Welding ,engineering.material ,021001 nanoscience & nanotechnology ,Condensed Matter Physics ,Microstructure ,law.invention ,020901 industrial engineering & automation ,Brittleness ,Mechanics of Materials ,law ,Hot isostatic pressing ,engineering ,General Materials Science ,Composite material ,0210 nano-technology ,Ductility ,Tensile testing - Abstract
Ultrasonic additive manufacturing (UAM) is a solid-state manufacturing technique employing principles of ultrasonic welding coupled with mechanized tape layering to fabricate fully functional parts. However, UAM-fabricated parts often exhibit a reduction in strength when loaded normal to the welding interfaces (Z-direction). Here, the effect of hot isostatic pressing (HIP) on UAM builds of aluminum alloy was explored. Tensile testing and microstructure characterization were conducted; it was established that HIP eliminated the brittle Z-direction fracture and improved the strength and ductility of the Z-direction specimens. HIP eliminated voids and produced recrystallized structure; however, welding interfaces survived the HIP treatment.
- Published
- 2018
- Full Text
- View/download PDF
49. Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa
- Author
-
Lance Lewis Snead, Kurt A. Terrani, Caen Ang, and Yutai Katoh
- Subjects
Nuclear and High Energy Physics ,Materials science ,Sintering ,02 engineering and technology ,Raw material ,021001 nanoscience & nanotechnology ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,0103 physical sciences ,General Materials Science ,Transient (oscillation) ,Irradiation ,Composite material ,0210 nano-technology ,Thermo mechanical ,Eutectic system - Abstract
Five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of
- Published
- 2018
- Full Text
- View/download PDF
50. Evaluating the irradiation effects on the elastic properties of miniature monolithic SiC tubular specimens
- Author
-
Takaaki Koyanagi, Gyanender Singh, Yutai Katoh, Kurt A. Terrani, and Christian M. Petrie
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,business.industry ,Modulus ,02 engineering and technology ,021001 nanoscience & nanotechnology ,01 natural sciences ,Finite element method ,law.invention ,chemistry.chemical_compound ,Nuclear Energy and Engineering ,chemistry ,Heat flux ,Optical microscope ,law ,Nondestructive testing ,0103 physical sciences ,Silicon carbide ,General Materials Science ,Irradiation ,Composite material ,0210 nano-technology ,business ,Elastic modulus - Abstract
The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. The optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.
- Published
- 2018
- Full Text
- View/download PDF
Catalog
Discovery Service for Jio Institute Digital Library
For full access to our library's resources, please sign in.