Pub Date : 2024-05-09DOI: 10.3389/fnuen.2024.1385478
Seda Yilmaz, Paul K. Romano, Lorenzo Chierici, Erik B. Knudsen, P. Shriwise
In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4–2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CAD-based particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer.
{"title":"CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC","authors":"Seda Yilmaz, Paul K. Romano, Lorenzo Chierici, Erik B. Knudsen, P. Shriwise","doi":"10.3389/fnuen.2024.1385478","DOIUrl":"https://doi.org/10.3389/fnuen.2024.1385478","url":null,"abstract":"In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4–2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CAD-based particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":" 20","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140995751","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-05-09DOI: 10.3389/fnuen.2024.1379414
Levi Morin Larsen, Kathryn Biegel, Nahuel Guaita, Andrew Foss
The U.S. Inflation Reduction Act (IRA) of 2022 provides a wide array of tax credits and other incentives for low-carbon energy. The technology-neutral clean generation production tax credit (PTC) (Section 45Y of the U.S. Internal Revenue Code) and the technology-neutral investment tax credit (ITC) (Section 48E) lower the net cost of new electricity generation projects with zero or negative greenhouse gas emission rates. We evaluate the impact of the IRA legislation—specifically the PTC and ITC—on the cost-competitiveness of small modular reactors (SMRs). We use the Argonne Low-carbon Energy Analysis Framework (A-LEAF) model to calculate the capacity factor of an SMR with a range of hypothetical variable operating and maintenance (O&M) costs in the Electric Reliability Council of Texas (ERCOT) electricity market. We selected ERCOT for market modeling because of its competitive structure, available data, and extensive use in prior literature. We use a discounted cash flow model to calculate the SMR’s net present value based on the market prices and capacity factors from A-LEAF, hypothetical ranges of capital and variable O&M costs, and other input parameters, with or without the IRA tax credits. We determine the SMR owner’s optimal choice of PTC or ITC for the hypothetical ranges of capital and variable O&M costs. We also evaluate potential shifts in the SMR owner’s optimal choice of PTC or ITC based on historical patterns of nuclear capital cost overruns in the United States. We also assess the sensitivity of our results to longer PTC period and electricity prices from the New England market, which tend to be higher than electricity prices in ERCOT. We find that even with the IRA tax credits, only SMRs with low capital and variable O&M costs would be economically feasible in the low-price ERCOT market scenario modeled. A longer PTC period and higher-price market such as New England, however, would significantly expand the economic feasibility of SMRs in the United States.
{"title":"Effects of the U.S. inflation reduction act on SMR economics","authors":"Levi Morin Larsen, Kathryn Biegel, Nahuel Guaita, Andrew Foss","doi":"10.3389/fnuen.2024.1379414","DOIUrl":"https://doi.org/10.3389/fnuen.2024.1379414","url":null,"abstract":"The U.S. Inflation Reduction Act (IRA) of 2022 provides a wide array of tax credits and other incentives for low-carbon energy. The technology-neutral clean generation production tax credit (PTC) (Section 45Y of the U.S. Internal Revenue Code) and the technology-neutral investment tax credit (ITC) (Section 48E) lower the net cost of new electricity generation projects with zero or negative greenhouse gas emission rates. We evaluate the impact of the IRA legislation—specifically the PTC and ITC—on the cost-competitiveness of small modular reactors (SMRs). We use the Argonne Low-carbon Energy Analysis Framework (A-LEAF) model to calculate the capacity factor of an SMR with a range of hypothetical variable operating and maintenance (O&M) costs in the Electric Reliability Council of Texas (ERCOT) electricity market. We selected ERCOT for market modeling because of its competitive structure, available data, and extensive use in prior literature. We use a discounted cash flow model to calculate the SMR’s net present value based on the market prices and capacity factors from A-LEAF, hypothetical ranges of capital and variable O&M costs, and other input parameters, with or without the IRA tax credits. We determine the SMR owner’s optimal choice of PTC or ITC for the hypothetical ranges of capital and variable O&M costs. We also evaluate potential shifts in the SMR owner’s optimal choice of PTC or ITC based on historical patterns of nuclear capital cost overruns in the United States. We also assess the sensitivity of our results to longer PTC period and electricity prices from the New England market, which tend to be higher than electricity prices in ERCOT. We find that even with the IRA tax credits, only SMRs with low capital and variable O&M costs would be economically feasible in the low-price ERCOT market scenario modeled. A longer PTC period and higher-price market such as New England, however, would significantly expand the economic feasibility of SMRs in the United States.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":" 3","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140995628","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-05-07DOI: 10.3389/fnuen.2024.1380108
A. E. White, E. Baglietto, M. Bucci, N. T. Howard, P. Rodriguez-Fernandez
In fusion, the validation of turbulent transport models is undertaken with the goals of making basic physics discoveries as well as for development of new predictive models to improve the operation and enhance the performance of existing and future fusion reactors. A fusion industry is just beginning to emerge globally. Like fission, validation in fusion energy research is a vibrant research area, but unlike fusion, a fission industry exists. The fission power industry motivates validation efforts, often performed at universities with small-scale experiments and advanced models and simulations developed in-house. Because fission research spans basic physics and applications, and addresses near-term and long-term industry interests, validation is thriving. This perspective article describes the validation of turbulent transport models in both fusion research and fission research, draws parallels between the validation methods and techniques used in two areas of the fields, and presents an outlook for thriving university fusion and fission research programs underpinned by a virtual cycle of basic and applied research that supports industry needs as well as tackling intellectual grand challenges.
{"title":"Fusion plasma turbulence research beyond the burning plasma era: perspectives on transport model validation in fusion and fission","authors":"A. E. White, E. Baglietto, M. Bucci, N. T. Howard, P. Rodriguez-Fernandez","doi":"10.3389/fnuen.2024.1380108","DOIUrl":"https://doi.org/10.3389/fnuen.2024.1380108","url":null,"abstract":"In fusion, the validation of turbulent transport models is undertaken with the goals of making basic physics discoveries as well as for development of new predictive models to improve the operation and enhance the performance of existing and future fusion reactors. A fusion industry is just beginning to emerge globally. Like fission, validation in fusion energy research is a vibrant research area, but unlike fusion, a fission industry exists. The fission power industry motivates validation efforts, often performed at universities with small-scale experiments and advanced models and simulations developed in-house. Because fission research spans basic physics and applications, and addresses near-term and long-term industry interests, validation is thriving. This perspective article describes the validation of turbulent transport models in both fusion research and fission research, draws parallels between the validation methods and techniques used in two areas of the fields, and presents an outlook for thriving university fusion and fission research programs underpinned by a virtual cycle of basic and applied research that supports industry needs as well as tackling intellectual grand challenges.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"28 18","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141005463","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-03-21DOI: 10.3389/fnuen.2024.1321627
Rosemary Montgomery, B. Bevard, P. Cantonwine, Yadukrishnan Sasikumar
At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.
{"title":"Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods","authors":"Rosemary Montgomery, B. Bevard, P. Cantonwine, Yadukrishnan Sasikumar","doi":"10.3389/fnuen.2024.1321627","DOIUrl":"https://doi.org/10.3389/fnuen.2024.1321627","url":null,"abstract":"At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"136 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-03-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140223708","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-03-12DOI: 10.3389/fnuen.2024.1363880
M. S. Espain, J. S. Viglietti, C. Vidal, D. Martin, V. Trivillin, A. Monti Hughes, M. A. Palmieri, L. De Leo, M. A. Garabalino, P. Curotto, S. I. Thorp, E. C. C. Pozzi, G. Saint Martin, A. Portu
The distribution of boron in biological samples coming from BNCT protocols can be determined by the analysis of autoradiographic images formed by nuclear tracks on polymeric detectors. The tissue section is first explored to delimit regions of interest (ROIs), and then it is removed in order to reveal nuclear tracks through an etching process. A quantitative autoradiographic technique was developed to measure the boron concentration in the sample by counting the nuclear tracks on the detector. However, under certain circumstances, rapid measurements of boron distribution are required. For that purpose, a methodology was developed, allowing a rapid and easy determination of boron concentration and distribution in a tissue section by measuring grey levels in the corresponding high fluence autoradiographic image (HFA). The grey levels are then converted to optical density values which are then correlated with boron concentration values. In order to validate the technique, consecutive sections of several tissue samples were analyzed by the conventional counting method and by HFA. The image analysis and boron concentration estimation can be carried out in under 15 min, and relative differences between zones can be easily determined.
{"title":"Optical density of high fluence autoradiographic images to study boron microdistribution: validation of the technique in tissue sections","authors":"M. S. Espain, J. S. Viglietti, C. Vidal, D. Martin, V. Trivillin, A. Monti Hughes, M. A. Palmieri, L. De Leo, M. A. Garabalino, P. Curotto, S. I. Thorp, E. C. C. Pozzi, G. Saint Martin, A. Portu","doi":"10.3389/fnuen.2024.1363880","DOIUrl":"https://doi.org/10.3389/fnuen.2024.1363880","url":null,"abstract":"The distribution of boron in biological samples coming from BNCT protocols can be determined by the analysis of autoradiographic images formed by nuclear tracks on polymeric detectors. The tissue section is first explored to delimit regions of interest (ROIs), and then it is removed in order to reveal nuclear tracks through an etching process. A quantitative autoradiographic technique was developed to measure the boron concentration in the sample by counting the nuclear tracks on the detector. However, under certain circumstances, rapid measurements of boron distribution are required. For that purpose, a methodology was developed, allowing a rapid and easy determination of boron concentration and distribution in a tissue section by measuring grey levels in the corresponding high fluence autoradiographic image (HFA). The grey levels are then converted to optical density values which are then correlated with boron concentration values. In order to validate the technique, consecutive sections of several tissue samples were analyzed by the conventional counting method and by HFA. The image analysis and boron concentration estimation can be carried out in under 15 min, and relative differences between zones can be easily determined.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"62 5","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140249354","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-22DOI: 10.3389/fnuen.2023.1340426
T. Ulrich, T. Besmann
The phases in uranium-silicide binary system were evaluated in regards to their stabilities, phase boundaries, crystal structures, and phase transitions. The results from this study were used in combination with a well assessed literature to optimize the U-Si phase diagram using the CALPHAD method. A thermodynamic database was developed, which could be used to guide nuclear fuel fabrication, could be incorporated into other nuclear fuel thermodynamic databases, or could be used to generate data required by fuel performance codes to model fuel behavior in normal or off-normal reactor operations. The U3Si2 and U3Si5 phases were modeled using the Compound Energy Formalism model with 3 sublattices to account for the variation in composition. The crystal structure used for the USi phase was the tetragonal with an I4/mmm space. Above 450°C, the U3Si5 phase was modeled. The composition of the USi2 phase was adjusted to USi1.84. The calculated invariant reactions and the enthalpy of formation for the stoichiometric phases were in agreement with experimental data.
{"title":"Phase equilibria of advanced technology uranium silicide-based nuclear fuel","authors":"T. Ulrich, T. Besmann","doi":"10.3389/fnuen.2023.1340426","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1340426","url":null,"abstract":"The phases in uranium-silicide binary system were evaluated in regards to their stabilities, phase boundaries, crystal structures, and phase transitions. The results from this study were used in combination with a well assessed literature to optimize the U-Si phase diagram using the CALPHAD method. A thermodynamic database was developed, which could be used to guide nuclear fuel fabrication, could be incorporated into other nuclear fuel thermodynamic databases, or could be used to generate data required by fuel performance codes to model fuel behavior in normal or off-normal reactor operations. The U3Si2 and U3Si5 phases were modeled using the Compound Energy Formalism model with 3 sublattices to account for the variation in composition. The crystal structure used for the USi phase was the tetragonal with an I4/mmm space. Above 450°C, the U3Si5 phase was modeled. The composition of the USi2 phase was adjusted to USi1.84. The calculated invariant reactions and the enthalpy of formation for the stoichiometric phases were in agreement with experimental data.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"69 7","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139606914","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-22DOI: 10.3389/fnuen.2023.1332377
Hazel M. Gardner, G. Zilahi, James Wade-Zhu
Two-micron thick erbium oxide tritium barrier coatings have been prepared by aerosol injection chemical vapor deposition and subsequently irradiated with 33 MeV Au 6+ ions at fluences up to 2.1 × 1016 Au/m2 at 550°C. Scanning electron microscopy, X-ray diffraction and transmission electron microscopy were used to investigate the coating surface morphologies, phase structures and cross-sectional microstructures as a function of irradiation and thermal treatment. XRD data was also used to extract information about the evolution of lattice strain in the coating. Some of the cubic erbia transformed to the monoclinic phase in the sample that was ion irradiated at temperature, and this was accompanied by a change from columnar to a more equiaxed grain structure. All coatings were found to experience out-of-plane tensile strain, thought to originate from thermal stresses created during coating manufacture. Thermal treatment reduced microstrains present in the as-deposited sample, whilst the cubic-to-monoclinic phase transformation reduced strain in the cubic phase but increased strain in the monoclinic phase.
{"title":"Investigating the thermal and irradiation stability of chemical vapor deposited erbium oxide tritium barrier coatings for Li breeder blanket applications","authors":"Hazel M. Gardner, G. Zilahi, James Wade-Zhu","doi":"10.3389/fnuen.2023.1332377","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1332377","url":null,"abstract":"Two-micron thick erbium oxide tritium barrier coatings have been prepared by aerosol injection chemical vapor deposition and subsequently irradiated with 33 MeV Au 6+ ions at fluences up to 2.1 × 1016 Au/m2 at 550°C. Scanning electron microscopy, X-ray diffraction and transmission electron microscopy were used to investigate the coating surface morphologies, phase structures and cross-sectional microstructures as a function of irradiation and thermal treatment. XRD data was also used to extract information about the evolution of lattice strain in the coating. Some of the cubic erbia transformed to the monoclinic phase in the sample that was ion irradiated at temperature, and this was accompanied by a change from columnar to a more equiaxed grain structure. All coatings were found to experience out-of-plane tensile strain, thought to originate from thermal stresses created during coating manufacture. Thermal treatment reduced microstrains present in the as-deposited sample, whilst the cubic-to-monoclinic phase transformation reduced strain in the cubic phase but increased strain in the monoclinic phase.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"46 22","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139608408","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-12DOI: 10.3389/fnuen.2023.1322209
Jordan Stone, Mira Khair, Steven Cavazos, Elizabeth S. Sooby
The advancement of nuclear energy technology necessitates the development of novel materials and synthesis methods to produce materials which enable new fuel cycles. Alongside the maturation of R&D scale technologies to produce these materials, there is an ongoing effort to develop in situ monitoring capabilities to reduce the time to the discovery and development of these fuels. Monitoring data can be leveraged in artificial intelligence platforms to detect phenomena which lead to varied macro- and microstructural features which impact the application and performance of samples synthesized. The present study presents early-stage findings of the implementation of high-temperature, high-frame-rate infrared thermal imaging to monitor the arc-melt synthesis of novel fuels and compounds relevant to advanced nuclear reactors. The study illustrates both the challenges and opportunities of this methodology, highlighting the importance of internal standards while determining emissivity and transmission values as well as visualizing volatilization during melt synthesis.
{"title":"High-temperature thermal imaging to inform the arc-melt synthesis of nuclear materials","authors":"Jordan Stone, Mira Khair, Steven Cavazos, Elizabeth S. Sooby","doi":"10.3389/fnuen.2023.1322209","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1322209","url":null,"abstract":"The advancement of nuclear energy technology necessitates the development of novel materials and synthesis methods to produce materials which enable new fuel cycles. Alongside the maturation of R&D scale technologies to produce these materials, there is an ongoing effort to develop in situ monitoring capabilities to reduce the time to the discovery and development of these fuels. Monitoring data can be leveraged in artificial intelligence platforms to detect phenomena which lead to varied macro- and microstructural features which impact the application and performance of samples synthesized. The present study presents early-stage findings of the implementation of high-temperature, high-frame-rate infrared thermal imaging to monitor the arc-melt synthesis of novel fuels and compounds relevant to advanced nuclear reactors. The study illustrates both the challenges and opportunities of this methodology, highlighting the importance of internal standards while determining emissivity and transmission values as well as visualizing volatilization during melt synthesis.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"59 14","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139532782","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-01-03DOI: 10.3389/fnuen.2023.1215418
Leonid Burakovsky, S. Ramsey, R. Baty
As UO2 is easily oxidized during the nuclear fuel cycle it is important to have a detailed understanding of the structures and properties of the oxidation products. Experimental work over the years has revealed many stable uranium oxides including UO2, U4O9 (UO2.25), U3O7 (UO2.33), U2O5 (UO2.5), U3O8 (UO2.67), and UO3, all with a number of different polymorphs. These oxides are broadly split into two categories, fluorite-based structures with stoichiometries in the range of UO2 to UO2.5 and less dense layered-type structures with stoichiometries in the range of UO2.5 to UO3. While UO2 is well characterized, both experimentally and computationally, there is a paucity of data concerning higher stoichiometry oxides in the literature. In this work we determine the ambient melting points of all the six stoichiometric uranium oxides listed above and compare them to the available experimental and/or theoretical data. We demonstrate that a family of the six ambient melting points map out a solid-liquid transition boundary consistent with the high-temperature portion of the phase diagram of uranium-oxygen system suggested by Babelot et al.
{"title":"Ambient melting behavior of stoichiometric uranium oxides","authors":"Leonid Burakovsky, S. Ramsey, R. Baty","doi":"10.3389/fnuen.2023.1215418","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1215418","url":null,"abstract":"As UO2 is easily oxidized during the nuclear fuel cycle it is important to have a detailed understanding of the structures and properties of the oxidation products. Experimental work over the years has revealed many stable uranium oxides including UO2, U4O9 (UO2.25), U3O7 (UO2.33), U2O5 (UO2.5), U3O8 (UO2.67), and UO3, all with a number of different polymorphs. These oxides are broadly split into two categories, fluorite-based structures with stoichiometries in the range of UO2 to UO2.5 and less dense layered-type structures with stoichiometries in the range of UO2.5 to UO3. While UO2 is well characterized, both experimentally and computationally, there is a paucity of data concerning higher stoichiometry oxides in the literature. In this work we determine the ambient melting points of all the six stoichiometric uranium oxides listed above and compare them to the available experimental and/or theoretical data. We demonstrate that a family of the six ambient melting points map out a solid-liquid transition boundary consistent with the high-temperature portion of the phase diagram of uranium-oxygen system suggested by Babelot et al.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"102 18","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139388026","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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