Pub Date : 2023-06-19DOI: 10.3389/fnuen.2022.1050262
Tiankai Yao, M. Bachhav, F. D. Di Lemma, Fei Xu, Fei Teng, D. Murray, M. Benson, L. Capriotti
U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.
{"title":"The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel","authors":"Tiankai Yao, M. Bachhav, F. D. Di Lemma, Fei Xu, Fei Teng, D. Murray, M. Benson, L. Capriotti","doi":"10.3389/fnuen.2022.1050262","DOIUrl":"https://doi.org/10.3389/fnuen.2022.1050262","url":null,"abstract":"U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127062441","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 : 2023-06-01DOI: 10.3389/fnuen.2023.1214113
J. Lechelle, L. Luzzi, Luteng Zhang
During the nuclear fuel cycle, stages of fabrication and in-pile processing imply diffusion phenomena are taking place in mixed oxide ceramics. These diffusion phenomena change the microstructure and key properties of the fuel, such as thermal conductivity and creep behavior. For instance, in the fabrication process, the interdiffusion of uranium and plutonium cations leads to a homogeneous solid solution and an increased grain size. Diffusion of oxygen anions leads to deviation from the stoichiometry of the fuel, which controls physical properties such as thermal conductivity. These phenomena can also occur during the in-pile stay of the fuel with an impact on creep properties. Gaseous fission products tend to remain in the bulk of fuel pellets. Thus, understanding diffusion phenomena makes it possible to avoid bad cooling of the fuel (due to the degradation of the fuel-cladding gap thermal conductance) at the beginning of in-pile operation and an increase in rod or pin internal pressure at the end of fuel life, which may lead to potential sheath embrittlement. The time required for the fission gas to reach grain boundaries depends on grain size. For all these reasons, it is necessary to be able to quantify these diffusion phenomena that occur either in the bulk of grains, at their surfaces, or along grain boundaries. This series of articles provides insight into the current knowledge in this field and can help in fuel performance codes, such as BISON (Yu et al., 2021), SCIANTIX (Pizzocri et al., 2020), or PLEIADES (Marelle et al., 2017). For future reactors with higher levels of Pu and, to a lesser extent, Am and Np, deviation from stoichiometry may reach higher values with an oxygen-to-metal ratio between 1.5 and 2.0, which occurs due to the behavior of the 5f electrons involved in the chemical binding. Deviation from stoichiometry may also lead to a reordering of oxygen vacancies and even to a different crystal structure. OPEN ACCESS
{"title":"Editorial: Experimental and model-based assessment of diffusion kinetics of actinides and oxygen during fuel fabrication and of fission products over in-pile use","authors":"J. Lechelle, L. Luzzi, Luteng Zhang","doi":"10.3389/fnuen.2023.1214113","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1214113","url":null,"abstract":"During the nuclear fuel cycle, stages of fabrication and in-pile processing imply diffusion phenomena are taking place in mixed oxide ceramics. These diffusion phenomena change the microstructure and key properties of the fuel, such as thermal conductivity and creep behavior. For instance, in the fabrication process, the interdiffusion of uranium and plutonium cations leads to a homogeneous solid solution and an increased grain size. Diffusion of oxygen anions leads to deviation from the stoichiometry of the fuel, which controls physical properties such as thermal conductivity. These phenomena can also occur during the in-pile stay of the fuel with an impact on creep properties. Gaseous fission products tend to remain in the bulk of fuel pellets. Thus, understanding diffusion phenomena makes it possible to avoid bad cooling of the fuel (due to the degradation of the fuel-cladding gap thermal conductance) at the beginning of in-pile operation and an increase in rod or pin internal pressure at the end of fuel life, which may lead to potential sheath embrittlement. The time required for the fission gas to reach grain boundaries depends on grain size. For all these reasons, it is necessary to be able to quantify these diffusion phenomena that occur either in the bulk of grains, at their surfaces, or along grain boundaries. This series of articles provides insight into the current knowledge in this field and can help in fuel performance codes, such as BISON (Yu et al., 2021), SCIANTIX (Pizzocri et al., 2020), or PLEIADES (Marelle et al., 2017). For future reactors with higher levels of Pu and, to a lesser extent, Am and Np, deviation from stoichiometry may reach higher values with an oxygen-to-metal ratio between 1.5 and 2.0, which occurs due to the behavior of the 5f electrons involved in the chemical binding. Deviation from stoichiometry may also lead to a reordering of oxygen vacancies and even to a different crystal structure. OPEN ACCESS","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116913327","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 : 2023-05-30DOI: 10.3389/fnuen.2023.1188789
M. Pannach, Iris Paschke, V. Metz, M. Altmaier, W. Voigt, D. Freyer
Sorel phases are the binder phases of the magnesia building material (Sorel cement/concrete) and of special concern for the construction of long-term stable geotechnical barriers in repositories for radioactive waste in rock salt, as potentially occurring brines are expected to contain MgCl2. Sorel phases, in addition to Mg(OH)2, are equally important as pH buffers to minimize solubility and potential mobilization of radionuclides in brine systems. In order to obtain a detailed database of the relevant solid-liquid equilibria and the related pHm values of the equilibrium solutions, extensive experimental investigations were carried out. Solid phase formation was studied by suspending MgO and Mg(OH)2 in NaCl saturated MgCl2-solutions at 25°C. Mg(OH)2 and the 3-1-8 Sorel phase were identified as the stable solid phases, while the 5-1-8 Sorel phase is metastable. Equilibration at 40°C did not lead to any solid phase changes. Both OH− and H+ equilibrium concentrations were analyzed as a function of MgCl2 concentration at 25°C and 40°C. In addition to our already published solid-liquid equilibria for the ternary system Mg-Cl-OH-H2O (25°C–120°C), the equilibrium H+ concentrations (pHm) determined at 25°C, 40°C and 60°C are now reported. Analyzing these data together with known ion-interaction Pitzer coefficients, the solubility constants for Mg(OH)2 and the 3-1-8 phase at these three temperatures, for the metastable 5-1-8 phase at 25°C and for the 2-1-4 phase at 60°C have been consistently calculated.
{"title":"Solid-liquid equilibria of Sorel phases and Mg (OH)2 in the system Na-Mg-Cl-OH-H2O. Part I: experimental determination of OH− and H+ equilibrium concentrations and solubility constants at 25°C, 40°C, and 60°C","authors":"M. Pannach, Iris Paschke, V. Metz, M. Altmaier, W. Voigt, D. Freyer","doi":"10.3389/fnuen.2023.1188789","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1188789","url":null,"abstract":"Sorel phases are the binder phases of the magnesia building material (Sorel cement/concrete) and of special concern for the construction of long-term stable geotechnical barriers in repositories for radioactive waste in rock salt, as potentially occurring brines are expected to contain MgCl2. Sorel phases, in addition to Mg(OH)2, are equally important as pH buffers to minimize solubility and potential mobilization of radionuclides in brine systems. In order to obtain a detailed database of the relevant solid-liquid equilibria and the related pHm values of the equilibrium solutions, extensive experimental investigations were carried out. Solid phase formation was studied by suspending MgO and Mg(OH)2 in NaCl saturated MgCl2-solutions at 25°C. Mg(OH)2 and the 3-1-8 Sorel phase were identified as the stable solid phases, while the 5-1-8 Sorel phase is metastable. Equilibration at 40°C did not lead to any solid phase changes. Both OH− and H+ equilibrium concentrations were analyzed as a function of MgCl2 concentration at 25°C and 40°C. In addition to our already published solid-liquid equilibria for the ternary system Mg-Cl-OH-H2O (25°C–120°C), the equilibrium H+ concentrations (pHm) determined at 25°C, 40°C and 60°C are now reported. Analyzing these data together with known ion-interaction Pitzer coefficients, the solubility constants for Mg(OH)2 and the 3-1-8 phase at these three temperatures, for the metastable 5-1-8 phase at 25°C and for the 2-1-4 phase at 60°C have been consistently calculated.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"166 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115041106","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 : 2023-05-18DOI: 10.3389/fnuen.2023.1170424
Andrew M. Ritzmann, Michael D. LaCount, M. Sassi, A. Johnson, N. Henson
In the event of a nuclear accident, fission products may be released into the environment. The release of 131I is of particular concern to human health. Iodine can be captured using a number of materials and frequently, this is accomplished with activated carbon impregnated with organic bases. Previous studies have used DFT and the graphite (0001) surface as a surrogate for adsorption, those studies focus on the species I•, I2, and CH3I. In this work we perform an ab initio study of the adsorption onto the surface of a graphite sheet of I2, CH3I, and inorganic acidic iodine species (HI, HOI, HIO2, and HIO3), which were selected to examine the possible effect of oxidation state on adsorption. The PBE exchange-correlation functional with D3 dispersion was employed. It was found that for molecular iodine, the iodine atoms tended to either situate above the center of a hexagonal site on the graphite or directly atop a carbon atom with the lighter components resting closer to the graphite. For each species the relative binding energies spanned the range of 21–33 kJ mol-1 and graphite-iodine distance was in the range of 3.52–3.93 Å. In all cases we found no significant charge transfer between the iodine species and the graphite, thus we conclude that all the iodine species studied undergo strong physisorption to the graphite.
{"title":"A density functional theory analysis of the adsorption and surface chemistry of inorganic iodine species on graphitea","authors":"Andrew M. Ritzmann, Michael D. LaCount, M. Sassi, A. Johnson, N. Henson","doi":"10.3389/fnuen.2023.1170424","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1170424","url":null,"abstract":"In the event of a nuclear accident, fission products may be released into the environment. The release of 131I is of particular concern to human health. Iodine can be captured using a number of materials and frequently, this is accomplished with activated carbon impregnated with organic bases. Previous studies have used DFT and the graphite (0001) surface as a surrogate for adsorption, those studies focus on the species I•, I2, and CH3I. In this work we perform an ab initio study of the adsorption onto the surface of a graphite sheet of I2, CH3I, and inorganic acidic iodine species (HI, HOI, HIO2, and HIO3), which were selected to examine the possible effect of oxidation state on adsorption. The PBE exchange-correlation functional with D3 dispersion was employed. It was found that for molecular iodine, the iodine atoms tended to either situate above the center of a hexagonal site on the graphite or directly atop a carbon atom with the lighter components resting closer to the graphite. For each species the relative binding energies spanned the range of 21–33 kJ mol-1 and graphite-iodine distance was in the range of 3.52–3.93 Å. In all cases we found no significant charge transfer between the iodine species and the graphite, thus we conclude that all the iodine species studied undergo strong physisorption to the graphite.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"50 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129723023","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 : 2023-04-24DOI: 10.3389/fnuen.2023.1176288
A. Ulbricht, J. Dykas, P. Chekhonin, E. Altstadt, F. Bergner
Post-irradiation annealing of neutron-irradiated reactor pressure vessel steels is a matter of both technical and scientific interest. Small-angle neutron scattering (SANS), while being sensitive to nm-sized irradiation-induced solute-atom clusters, provides macroscopically representative and statistically reliable measures of cluster volume fraction, number density and size. In the present study, SANS was applied to uncover the size distribution of clusters in as-irradiated samples of a VVER-1000 weld and their gradual dissolution as function of the post-irradiation annealing temperature. The same samples were used to measure Vickers hardness. The results are consistent with Mn-Ni-Si-rich clusters of less than 2 nm radius to be the dominant source of both scattering and hardening. Annealing gave rise to small but significant partial recovery at 350°C and almost complete recovery at 475°C. The dispersed-barrier hardening model was applied to bridge the gap between the characteristics of nano-features and macro-hardness.
中子辐照反应堆压力容器钢的辐照后退火是一个既有技术意义又有科学意义的问题。小角中子散射(SANS)虽然对纳米尺寸的辐照诱导的溶质原子团簇敏感，但提供了宏观上具有代表性和统计上可靠的团簇体积分数、数量密度和大小的测量。在本研究中，应用SANS揭示了VVER-1000焊缝辐照后样品中团簇的尺寸分布以及它们随辐照后退火温度的逐渐溶解。使用相同的样品测量维氏硬度。结果表明，半径小于2 nm的富mn - ni - si团簇是散射和硬化的主要来源。退火在350°C时产生小但显著的部分恢复，在475°C时几乎完全恢复。采用分散性屏障硬化模型来弥补纳米特征与宏观硬度之间的差距。
{"title":"Small-angle neutron scattering study of neutron-irradiated and post-irradiation annealed VVER-1000 reactor pressure vessel weld material","authors":"A. Ulbricht, J. Dykas, P. Chekhonin, E. Altstadt, F. Bergner","doi":"10.3389/fnuen.2023.1176288","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1176288","url":null,"abstract":"Post-irradiation annealing of neutron-irradiated reactor pressure vessel steels is a matter of both technical and scientific interest. Small-angle neutron scattering (SANS), while being sensitive to nm-sized irradiation-induced solute-atom clusters, provides macroscopically representative and statistically reliable measures of cluster volume fraction, number density and size. In the present study, SANS was applied to uncover the size distribution of clusters in as-irradiated samples of a VVER-1000 weld and their gradual dissolution as function of the post-irradiation annealing temperature. The same samples were used to measure Vickers hardness. The results are consistent with Mn-Ni-Si-rich clusters of less than 2 nm radius to be the dominant source of both scattering and hardening. Annealing gave rise to small but significant partial recovery at 350°C and almost complete recovery at 475°C. The dispersed-barrier hardening model was applied to bridge the gap between the characteristics of nano-features and macro-hardness.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125010770","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 : 2023-04-14DOI: 10.3389/fnuen.2023.1158109
F. Bok, H. Moog, V. Brendler
Oxygen is one of the key reaction partners for many redox reactions also in the context of nuclear waste disposal. Its solubility influences radionuclides’ behavior, corrosion processes and even microbial activity. Therefore, a reliable calculation of the solubility of molecular oxygen in aqueous solutions is relevant for any safety assessment. Available geochemical speciation and reactive transport programs handle these data very differently. In some codes, the hypothetical equilibrium between dissolved oxygen and water is used to balance redox reactions. Equilibrium constants are given in “temperature grids” for up to 573.15 K. In other cases, temperature functions for the solubility of gaseous oxygen in water are given, without any reference to a valid temperature range. These settings become even more complicated when used in the context of modeling equilibria in high-saline solutions applying the Pitzer formalism. This raised the question about the experimental foundation of equilibrium constants given in such data files and their validity for the solubility of molecular oxygen in saline solutions. For this article, a thorough literature review was conducted with respect to the solubility of molecular oxygen in pure water and saline solutions. From these primary experimental O2 solubility data a temperature-dependent Henry’s law function as well as temperature-dependent binary and ternary Pitzer ion-interaction coefficients were derived. An internally consistent set of thermodynamic data for dissolved oxygen is presented, along with statements about its validity in terms of temperature and, as far as Pitzer interaction coefficients are concerned, of solution composition. This self-consistent activity-fugacity model containing thermodynamic data, Henry’s law temperature equation, and Pitzer interaction coefficients is capable of providing a more accurate description of redox transformations, allowing a reduction of conservatism in safety assessment calculations, not only in the context of a nuclear repository. The model reproduces well the reliable experimental data available, and is capable to predict the oxygen solubility in complex solution media. The temperature functions used to describe Henry’s constant and the Pitzer interaction coefficients are consistent with the implementation in commonly used geochemical computational programs, allowing direct use without further modification.
{"title":"The solubility of oxygen in water and saline solutions","authors":"F. Bok, H. Moog, V. Brendler","doi":"10.3389/fnuen.2023.1158109","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1158109","url":null,"abstract":"Oxygen is one of the key reaction partners for many redox reactions also in the context of nuclear waste disposal. Its solubility influences radionuclides’ behavior, corrosion processes and even microbial activity. Therefore, a reliable calculation of the solubility of molecular oxygen in aqueous solutions is relevant for any safety assessment. Available geochemical speciation and reactive transport programs handle these data very differently. In some codes, the hypothetical equilibrium between dissolved oxygen and water is used to balance redox reactions. Equilibrium constants are given in “temperature grids” for up to 573.15 K. In other cases, temperature functions for the solubility of gaseous oxygen in water are given, without any reference to a valid temperature range. These settings become even more complicated when used in the context of modeling equilibria in high-saline solutions applying the Pitzer formalism. This raised the question about the experimental foundation of equilibrium constants given in such data files and their validity for the solubility of molecular oxygen in saline solutions. For this article, a thorough literature review was conducted with respect to the solubility of molecular oxygen in pure water and saline solutions. From these primary experimental O2 solubility data a temperature-dependent Henry’s law function as well as temperature-dependent binary and ternary Pitzer ion-interaction coefficients were derived. An internally consistent set of thermodynamic data for dissolved oxygen is presented, along with statements about its validity in terms of temperature and, as far as Pitzer interaction coefficients are concerned, of solution composition. This self-consistent activity-fugacity model containing thermodynamic data, Henry’s law temperature equation, and Pitzer interaction coefficients is capable of providing a more accurate description of redox transformations, allowing a reduction of conservatism in safety assessment calculations, not only in the context of a nuclear repository. The model reproduces well the reliable experimental data available, and is capable to predict the oxygen solubility in complex solution media. The temperature functions used to describe Henry’s constant and the Pitzer interaction coefficients are consistent with the implementation in commonly used geochemical computational programs, allowing direct use without further modification.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133880652","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 : 2023-04-04DOI: 10.3389/fnuen.2023.1118594
Hye-Ryun Cho, Sangki Cho, Jueun Kim, S. Han, Hee-Kyung Kim, W. Um
PuO2(cr) dissolution in natural water was investigated at 25°C and 60°C under atmospheric conditions. The concentration of Pu in solutions [Pu], was monitored for 1 year of reaction time. PuO2(cr) dissolution in natural water reached a steady state within 2 months at 25°C. The [Pu] in groundwater and seawater at pH 8 were in the range of [Pu] = 0.9–34 and 3.4–27 nM, respectively. The [Pu] in concrete porewater (rainwater equilibrated with concrete) at pH 8.1–10.9 was in the range of 0.1–3.2 nM. The [Pu] and pH values of groundwater were similar to those of seawater samples having a high ionic strength. The measured [Pu] at equilibrium in all samples was higher than the calculated solubility curves for PuO2(am, hyd). Experimental evidence is insufficient to confirm the oxidation state of Pu in solution and solid phases. However, the results of geochemical modeling indicate that PuO2(am, hyd) and aqueous Pu(IV) species are dominant in natural water samples of this work. The dissolution behavior of PuO2(cr) in natural waters is comparable to the oxidative dissolution of PuO2(am, hyd) in the presence of PuO2(coll, hyd). The dissolution of PuO2 in groundwater decreased at higher temperatures, whereas the influence of temperature in seawater and porewater was not significant under these experimental conditions.
{"title":"Dissolution behaviors of PuO2(cr) in natural waters","authors":"Hye-Ryun Cho, Sangki Cho, Jueun Kim, S. Han, Hee-Kyung Kim, W. Um","doi":"10.3389/fnuen.2023.1118594","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1118594","url":null,"abstract":"PuO2(cr) dissolution in natural water was investigated at 25°C and 60°C under atmospheric conditions. The concentration of Pu in solutions [Pu], was monitored for 1 year of reaction time. PuO2(cr) dissolution in natural water reached a steady state within 2 months at 25°C. The [Pu] in groundwater and seawater at pH 8 were in the range of [Pu] = 0.9–34 and 3.4–27 nM, respectively. The [Pu] in concrete porewater (rainwater equilibrated with concrete) at pH 8.1–10.9 was in the range of 0.1–3.2 nM. The [Pu] and pH values of groundwater were similar to those of seawater samples having a high ionic strength. The measured [Pu] at equilibrium in all samples was higher than the calculated solubility curves for PuO2(am, hyd). Experimental evidence is insufficient to confirm the oxidation state of Pu in solution and solid phases. However, the results of geochemical modeling indicate that PuO2(am, hyd) and aqueous Pu(IV) species are dominant in natural water samples of this work. The dissolution behavior of PuO2(cr) in natural waters is comparable to the oxidative dissolution of PuO2(am, hyd) in the presence of PuO2(coll, hyd). The dissolution of PuO2 in groundwater decreased at higher temperatures, whereas the influence of temperature in seawater and porewater was not significant under these experimental conditions.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129007585","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 : 2023-03-29DOI: 10.3389/fnuen.2023.1151271
Erik Coppens, K. Wouters, Benny de Blochouse, D. Durce
In the frame of the safe disposal of short-lived low and intermediate level nuclear waste (SL-ILW), ONDRAF/NIRAS (Belgium) has submitted a license application for the exploitation of a near surface facility in Dessel (Belgium). A significant part of the waste intended for the surface repository is Pu-contaminated and has been conditioned by means of CEM III/C based mortar, produced in the CILVA-installation at the Belgoprocess site in Dessel. To establish more accurate data on sorption of Pu to the CILVA mortar, an experimental test set-up was designed in order to screen which factors were likely to affect Pu sorption to the mortar. The different factors of the design were variables related to the pore water composition of the mortar on the one hand (concentrations of Ca2+, Cl−, SO4 2−, S2−, K+ and OH− (pH)), and variables characteristic for batch sorption experiments on the other hand ([Pu], solid-to-liquid ratio and equilibration time). The results of this screening indicate that over the tested variables, only the concentration of Ca2+ in the synthetic pore water affects Pu sorption to the CILVA matrix to a significant extent. Additionally, from literature it is expected that the presence of isosaccharinic acid (ISA), a cellulose degradation product, would affect Pu sorption, with increasing concentrations of ISA frequently correlated with decreased sorption. To address the nature and extent of the impact of both [Ca2+] and [ISA] and their combined effect on sorption of Pu to the mortar, an experimental set-up for surface response measurement (SRM) was designed. A Central Composite Design (CCD) in two factors was selected for the SRM, with three test points and a four point repetition of the centre point. The execution of this experimental set-up and the resulting responses, allowed for the development of a polynomial model to predict the average response of Pu sorption (expressed as R d ) as a function of [ISA] and of [Ca2+]. In addition, the [Ca2+] in solution in equilibrium with the mortar could be assessed from the established dataset, which allowed to predict Pu sorption as a function of [ISA] at the intrinsic [Ca2+] in the mortar’s pore water.
{"title":"Investigating the effect of pore water composition and isosaccharinic acid (ISA) on sorption of Pu to a CEM III/C-based mortar","authors":"Erik Coppens, K. Wouters, Benny de Blochouse, D. Durce","doi":"10.3389/fnuen.2023.1151271","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1151271","url":null,"abstract":"In the frame of the safe disposal of short-lived low and intermediate level nuclear waste (SL-ILW), ONDRAF/NIRAS (Belgium) has submitted a license application for the exploitation of a near surface facility in Dessel (Belgium). A significant part of the waste intended for the surface repository is Pu-contaminated and has been conditioned by means of CEM III/C based mortar, produced in the CILVA-installation at the Belgoprocess site in Dessel. To establish more accurate data on sorption of Pu to the CILVA mortar, an experimental test set-up was designed in order to screen which factors were likely to affect Pu sorption to the mortar. The different factors of the design were variables related to the pore water composition of the mortar on the one hand (concentrations of Ca2+, Cl−, SO4 2−, S2−, K+ and OH− (pH)), and variables characteristic for batch sorption experiments on the other hand ([Pu], solid-to-liquid ratio and equilibration time). The results of this screening indicate that over the tested variables, only the concentration of Ca2+ in the synthetic pore water affects Pu sorption to the CILVA matrix to a significant extent. Additionally, from literature it is expected that the presence of isosaccharinic acid (ISA), a cellulose degradation product, would affect Pu sorption, with increasing concentrations of ISA frequently correlated with decreased sorption. To address the nature and extent of the impact of both [Ca2+] and [ISA] and their combined effect on sorption of Pu to the mortar, an experimental set-up for surface response measurement (SRM) was designed. A Central Composite Design (CCD) in two factors was selected for the SRM, with three test points and a four point repetition of the centre point. The execution of this experimental set-up and the resulting responses, allowed for the development of a polynomial model to predict the average response of Pu sorption (expressed as R d ) as a function of [ISA] and of [Ca2+]. In addition, the [Ca2+] in solution in equilibrium with the mortar could be assessed from the established dataset, which allowed to predict Pu sorption as a function of [ISA] at the intrinsic [Ca2+] in the mortar’s pore water.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"61 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126540379","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 : 2023-03-27DOI: 10.3389/fnuen.2023.1142823
I. Maamoun, K. Tokunaga, T. Dohi, Futoshi Kanno, O. Falyouna, O. Eljamal, Kazuya Tanaka
Introduction: In this work, bimetallic nickel/iron nanoparticles (Ni/Fe0) were prepared to enhance rhenium (Re(VII)) immobilization from aqueous solutions, as the surrogate of technetium (Tc(VII)). Methods: Two synthesis approaches of Ni/Fe0, pre-, and post-nucleation, were investigated towards Re(VII) removal. Different characterization techniques were considered to elucidate the physicochemical features of the fresh and reacted materials, such as scanning transmission electron microscopy-energy dispersive X-ray spectroscopy (STEM-EDX), X-ray diffraction (XRD), and X-ray absorption near edge spectroscopy (XANES). The influence of several reaction parameters on Re(VII) removal was investigated, including Ni/Fe0 ratio, dosage, initial pH, temperature, and initial concentration. Results and discussion: Results showed a promising potential of Ni/Fe0, either pre- or post-nucleation synthesized, in Re(VII) removal, especially at the early stage of the reaction, where Ni/Fe0: 0.4 yielded almost full removal efficiency of initial 15.0 μM-Re(VII) within the first 10 min of reaction. Even at low Ni/Fe0 dosages, such as 0.25 and 0.5 g/L, reasonable removal efficiency was achieved after 2 h reaction time of ∼73% and ∼98%, respectively. Unlike acidic/neutral pH, alkaline conditions were not favorable for Re(VII) removal by Ni/Fe0 owing to the delayed aqueous corrosion of Fe0-core resulting in insufficiency of electrons available for Re(VII) reduction. The reductive abilities were confirmed by XANES, revealing Re(VII) reduction to Re(IV)/(III) by the released electrons from Fe0-core in both Fe0 and Ni/Fe0 materials. Pseudo-first- and second-order kinetic models were suitable to describe Re(VII) removal by Ni/Fe0, implying physical and/or chemical processes were involved. Zeta potential measurements depicted the point of zero charge (pHPZC) of Fe0 and Ni/Fe0 to be 8.24 and 7.63, respectively, suggesting the involvement of electrostatic sorption of ReO4− on the positively charged surface of Ni/Fe0. The occurrence of multi- and mono-layer sorption within Re(VII) removal process was implicated, following Freundlich and Sips isotherm models. The presence of Ni0/NiO on Fe0-surface resulted in providing an efficient electron-transfer medium that facilitated Re(VII) reduction, leading to impressive kinetic rates. Overall, our study provided valuable insights into the use of Ni/Fe0 for Re(VII) removal from water and offered guidance for future research in such an aspect towards the pilot-scale applications of Tc(VII) removal from nuclear wastewater.
{"title":"Improved immobilization of Re(VII) from aqueous solutions via bimetallic Ni/Fe0 nanoparticles: Implications towards Tc(VII) removal","authors":"I. Maamoun, K. Tokunaga, T. Dohi, Futoshi Kanno, O. Falyouna, O. Eljamal, Kazuya Tanaka","doi":"10.3389/fnuen.2023.1142823","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1142823","url":null,"abstract":"Introduction: In this work, bimetallic nickel/iron nanoparticles (Ni/Fe0) were prepared to enhance rhenium (Re(VII)) immobilization from aqueous solutions, as the surrogate of technetium (Tc(VII)). Methods: Two synthesis approaches of Ni/Fe0, pre-, and post-nucleation, were investigated towards Re(VII) removal. Different characterization techniques were considered to elucidate the physicochemical features of the fresh and reacted materials, such as scanning transmission electron microscopy-energy dispersive X-ray spectroscopy (STEM-EDX), X-ray diffraction (XRD), and X-ray absorption near edge spectroscopy (XANES). The influence of several reaction parameters on Re(VII) removal was investigated, including Ni/Fe0 ratio, dosage, initial pH, temperature, and initial concentration. Results and discussion: Results showed a promising potential of Ni/Fe0, either pre- or post-nucleation synthesized, in Re(VII) removal, especially at the early stage of the reaction, where Ni/Fe0: 0.4 yielded almost full removal efficiency of initial 15.0 μM-Re(VII) within the first 10 min of reaction. Even at low Ni/Fe0 dosages, such as 0.25 and 0.5 g/L, reasonable removal efficiency was achieved after 2 h reaction time of ∼73% and ∼98%, respectively. Unlike acidic/neutral pH, alkaline conditions were not favorable for Re(VII) removal by Ni/Fe0 owing to the delayed aqueous corrosion of Fe0-core resulting in insufficiency of electrons available for Re(VII) reduction. The reductive abilities were confirmed by XANES, revealing Re(VII) reduction to Re(IV)/(III) by the released electrons from Fe0-core in both Fe0 and Ni/Fe0 materials. Pseudo-first- and second-order kinetic models were suitable to describe Re(VII) removal by Ni/Fe0, implying physical and/or chemical processes were involved. Zeta potential measurements depicted the point of zero charge (pHPZC) of Fe0 and Ni/Fe0 to be 8.24 and 7.63, respectively, suggesting the involvement of electrostatic sorption of ReO4− on the positively charged surface of Ni/Fe0. The occurrence of multi- and mono-layer sorption within Re(VII) removal process was implicated, following Freundlich and Sips isotherm models. The presence of Ni0/NiO on Fe0-surface resulted in providing an efficient electron-transfer medium that facilitated Re(VII) reduction, leading to impressive kinetic rates. Overall, our study provided valuable insights into the use of Ni/Fe0 for Re(VII) removal from water and offered guidance for future research in such an aspect towards the pilot-scale applications of Tc(VII) removal from nuclear wastewater.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"50 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127432678","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 : 2023-03-17DOI: 10.3389/fnuen.2023.1162751
D. LaBrier, C. Pope, W. Marcum
The Versatile Test Reactor is meant to fill a need in the advanced nuclear reactor development community. Accelerated nuclear testing of fuels and materials is needed to allow designs to proceed. To accomplish this, a fast neutron spectrum is needed. To have a fast neutron spectrum, the reactor coolant needs to not only have excellent heat transfer properties to sufficiently cool the reactor core, but also not significantly slow neutrons down following their birth in the fission process. Sodium is an excellent coolant choice that can fulfill these needs and many others. Sodium is the sixth most abundant element in the Earth’s crust and can be acquired easily through simple chemical separation methods. Sodium, however, reacts when in contact with water and air, but with proper precautions, these scenarios can be avoided. Nuclear reactors that utilize molten sodium as a primary heat transfer medium [a.k.a. sodium-cooled fast-neutron reactors, or SFRs, Figure 1 (1)] provide unique challenges in their daily operation, and therefore require focused training for researchers, scientists, staff, and employees alike to understand these unique challenges. Therefore, a comprehensive understanding of sodium properties is needed by all personnel involved in the design, construction, operation, experiment planning and execution of SFRs. Additionally, several commercial advanced reactor designs are under development which will use sodium for the reactor coolant. These advanced reactor designs include the Natrium reactor from TerraPower and GE-Hitachi (2), the ARC-100 reactor from Advanced Reactor Concepts (3), and the Aurora reactor from Oklo (4).
{"title":"On developing a practical safety culture for the advanced reactor workforce: Experiences of working with sodium","authors":"D. LaBrier, C. Pope, W. Marcum","doi":"10.3389/fnuen.2023.1162751","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1162751","url":null,"abstract":"The Versatile Test Reactor is meant to fill a need in the advanced nuclear reactor development community. Accelerated nuclear testing of fuels and materials is needed to allow designs to proceed. To accomplish this, a fast neutron spectrum is needed. To have a fast neutron spectrum, the reactor coolant needs to not only have excellent heat transfer properties to sufficiently cool the reactor core, but also not significantly slow neutrons down following their birth in the fission process. Sodium is an excellent coolant choice that can fulfill these needs and many others. Sodium is the sixth most abundant element in the Earth’s crust and can be acquired easily through simple chemical separation methods. Sodium, however, reacts when in contact with water and air, but with proper precautions, these scenarios can be avoided. Nuclear reactors that utilize molten sodium as a primary heat transfer medium [a.k.a. sodium-cooled fast-neutron reactors, or SFRs, Figure 1 (1)] provide unique challenges in their daily operation, and therefore require focused training for researchers, scientists, staff, and employees alike to understand these unique challenges. Therefore, a comprehensive understanding of sodium properties is needed by all personnel involved in the design, construction, operation, experiment planning and execution of SFRs. Additionally, several commercial advanced reactor designs are under development which will use sodium for the reactor coolant. These advanced reactor designs include the Natrium reactor from TerraPower and GE-Hitachi (2), the ARC-100 reactor from Advanced Reactor Concepts (3), and the Aurora reactor from Oklo (4).","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115724668","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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