A. Pavliuk, Evgeniy V. Bespala, S. Kotlyarevskiy, I. Novoselov, V. N. Kotov
The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors. The analysis of heat release processes inside storages containing irradiated nuclear graphite, representing a potential hazard due to the possible heating and, accordingly, the release of long-lived radionuclides during oxidation was carried out. The following factors were considered as the main factors that can lead to an increase in the temperature inside the storage facility: corrosion of metallic radioactive waste, the presence of fuel fragments, and also the random exposure of irradiated graphite to local sources of thermal energy (spark, etc.). It was noted in the work that the combined or separate influence of some factors can lead to an increase in the temperature of the onset of the initiation of Wigner energy release in graphite radwaste (Tin ≈ 90–100°C for the “Worst-case” graphite). The model of heat generation in the storage was developed based on the analysis of the features of graphite radioactive waste storage and Wigner energy release. The layered location of different types of waste (graphite and aluminum) and the local character of the distribution of heat sources were adopted in this model. The greatest heating is achieved if graphite radioactive waste is located near the concrete walls of the storage facility, as well as in direct contact with irradiated aluminum radioactive waste, which was shown in this paper.
{"title":"Analysis of Heat Release Processes inside Storage Facilities Containing Irradiated Nuclear Graphite","authors":"A. Pavliuk, Evgeniy V. Bespala, S. Kotlyarevskiy, I. Novoselov, V. N. Kotov","doi":"10.1155/2022/2957310","DOIUrl":"https://doi.org/10.1155/2022/2957310","url":null,"abstract":"The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors. The analysis of heat release processes inside storages containing irradiated nuclear graphite, representing a potential hazard due to the possible heating and, accordingly, the release of long-lived radionuclides during oxidation was carried out. The following factors were considered as the main factors that can lead to an increase in the temperature inside the storage facility: corrosion of metallic radioactive waste, the presence of fuel fragments, and also the random exposure of irradiated graphite to local sources of thermal energy (spark, etc.). It was noted in the work that the combined or separate influence of some factors can lead to an increase in the temperature of the onset of the initiation of Wigner energy release in graphite radwaste (Tin ≈ 90–100°C for the “Worst-case” graphite). The model of heat generation in the storage was developed based on the analysis of the features of graphite radioactive waste storage and Wigner energy release. The layered location of different types of waste (graphite and aluminum) and the local character of the distribution of heat sources were adopted in this model. The greatest heating is achieved if graphite radioactive waste is located near the concrete walls of the storage facility, as well as in direct contact with irradiated aluminum radioactive waste, which was shown in this paper.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46561240","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm ∗ 60 mm and 2.8 mm ∗ 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that Δ T ONB (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, Δ T ONB in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.
在质量通量为100 ~ 300 kg/(m2·s)，进水温度为70 ~ 100℃，热流密度为10 ~ 70 kW/m2，局部压力为0.145 MPa的条件下，对两个垂直矩形通道内的ONB(核沸腾起始)进行了实验研究。截面尺寸分别为1.8 mm∗60 mm和2.8 mm∗60 mm。利用三种沸腾起始判断方法对ONB点进行了定位，发现Δ T ONB (ONB点壁面过热度)随入口温度的降低而增大，随质量通量的增大而增大。结果还表明，尽管在ONB位置，窄通道内的气泡大小和行为与非窄通道内的气泡大小和行为不同，但传热没有受到明显影响。此外，通过Thom提出的相关性可以在±30%的误差范围内预测两个通道中的Δ T ONB。
{"title":"Experimental Study of Onset of Nucleate Boiling in Vertical Rectangular Channels with Different Flow Path Heights","authors":"N. Cheng, Shuwen Yu, J. Xiao, C. Peng","doi":"10.1155/2022/7760569","DOIUrl":"https://doi.org/10.1155/2022/7760569","url":null,"abstract":"A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm \u0000 \u0000 ∗\u0000 \u0000 60 mm and 2.8 mm \u0000 \u0000 ∗\u0000 \u0000 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that \u0000 \u0000 Δ\u0000 \u0000 \u0000 T\u0000 \u0000 \u0000 ONB\u0000 \u0000 \u0000 \u0000 (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, \u0000 \u0000 Δ\u0000 \u0000 \u0000 T\u0000 \u0000 \u0000 ONB\u0000 \u0000 \u0000 \u0000 in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41486924","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Weibao Li, Lei Yang, Junling Chen, Jiansheng Hu, B. Guo, Lili Zhu
In the experimental advanced superconducting tokamak (EAST), the cooling channels of plasma-facing components (PFCs) are familiarly connected in parallel through manifolds. According to the drainage performance of the PFCs, the amount of water trapped in the cooling channels is directly correlated with the type of manifolds. To date, manifold types have been well studied with respect to single-phase and mixed multiphase flow characteristics. However, there are few studies on the drainage performance relevant to the type of manifolds. The friction effect and inertia effect in the manifold intake and exhaust are studied through theoretical analysis. In addition, the draining liquid effect in the branches is dependent on the pressure difference and the resistance coefficient. Furthermore, U-type and Z-type manifolds are studied with FLUENT to discuss their drainage capability in this study. The distribution of the water volume fraction, pressure, and flow ratio is obtained to verify that the Z type is more effective than the U type when applied in the drainage process. This is also supported by comparing the experimental value of the two in drainage discharge. This study will provide a theoretical basis for future upgrades of EAST drainage systems.
{"title":"Investigation of the Water Draining Process Pushed by Gas for U-Type and Z-Type Manifolds","authors":"Weibao Li, Lei Yang, Junling Chen, Jiansheng Hu, B. Guo, Lili Zhu","doi":"10.1155/2022/5555759","DOIUrl":"https://doi.org/10.1155/2022/5555759","url":null,"abstract":"In the experimental advanced superconducting tokamak (EAST), the cooling channels of plasma-facing components (PFCs) are familiarly connected in parallel through manifolds. According to the drainage performance of the PFCs, the amount of water trapped in the cooling channels is directly correlated with the type of manifolds. To date, manifold types have been well studied with respect to single-phase and mixed multiphase flow characteristics. However, there are few studies on the drainage performance relevant to the type of manifolds. The friction effect and inertia effect in the manifold intake and exhaust are studied through theoretical analysis. In addition, the draining liquid effect in the branches is dependent on the pressure difference and the resistance coefficient. Furthermore, U-type and Z-type manifolds are studied with FLUENT to discuss their drainage capability in this study. The distribution of the water volume fraction, pressure, and flow ratio is obtained to verify that the Z type is more effective than the U type when applied in the drainage process. This is also supported by comparing the experimental value of the two in drainage discharge. This study will provide a theoretical basis for future upgrades of EAST drainage systems.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48627977","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yuman Sun, H. Xue, K. Zhao, Yubiao Zhang, Y. Zhao, W. Yan, R. Bashir
The complicated driving force at the stress corrosion cracking (SCC) tip of the safe-end dissimilar metal-welded joints (DMWJs) in the pressurized water reactor (PWR) is mainly caused by the heterogeneous material mechanical properties. In this research, to accurately evaluate the crack driving force at the SCC in DMWJs, the stress-strain condition, stress triaxiality, and J-integral of the crack tip at different positions are analyzed based on the heterogeneous material properties model. The results indicate that the larger driving force will be provided for the I-type crack when the crack is in the SA508 zone and the interface between the 316L region and base metal. In addition, the heterogeneous material properties inhibit the J-integral of the crack in the 316L region, which has a promoting effect when the crack is in the SA508 zone and weld metal. It provides a new idea for analyzing driving force at the crack tip and safety evaluation of DMWJs in PWRs.
{"title":"Cracking Driving Force at the Tip of SCC under Heterogeneous Material Mechanics Model of Safe-End Dissimilar Metal-Welded Joints in PWR","authors":"Yuman Sun, H. Xue, K. Zhao, Yubiao Zhang, Y. Zhao, W. Yan, R. Bashir","doi":"10.1155/2022/6605101","DOIUrl":"https://doi.org/10.1155/2022/6605101","url":null,"abstract":"The complicated driving force at the stress corrosion cracking (SCC) tip of the safe-end dissimilar metal-welded joints (DMWJs) in the pressurized water reactor (PWR) is mainly caused by the heterogeneous material mechanical properties. In this research, to accurately evaluate the crack driving force at the SCC in DMWJs, the stress-strain condition, stress triaxiality, and J-integral of the crack tip at different positions are analyzed based on the heterogeneous material properties model. The results indicate that the larger driving force will be provided for the I-type crack when the crack is in the SA508 zone and the interface between the 316L region and base metal. In addition, the heterogeneous material properties inhibit the J-integral of the crack in the 316L region, which has a promoting effect when the crack is in the SA508 zone and weld metal. It provides a new idea for analyzing driving force at the crack tip and safety evaluation of DMWJs in PWRs.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47208310","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yuman Sun, H. Xue, Fu-qiang Yang, Shuai Wang, Shun Zhang, Jinxuan He, R. Bashir
The material mechanical properties and crack propagation behavior of dissimilar metal welded joint (DMWJ) of pressurized water reactor (PWR) was investigated. In this research, the mechanical parameters of the cladding layer materials (304L-SA508) of the DMWJ in PWRs were obtained by the continuous indentation test. Simultaneously, the user-defined (USDFLD) subroutine in ABAQUS was used to establish the heterogeneous materials model of the welded joint. On this basis, the local crack propagation path of DMWJs has been discussed based on the extended finite element method (XFEM). The result indicated that the strength value at the fusion boundary line (FB line) is the largest, and the yield strength reaches 689 MPa. The yield stress values of the cladding metal (304 L) and base metal (SA508) are 371 MPa and 501 MPa, respectively. Affected by the material constraint effect of the DMWJ, the crack will propagate through the FB line when the initial crack is perpendicular to the FB line. And when the initial crack parallels the FB line, the crack will deviate from it. Meanwhile, the crack propagation length is smaller as the initial crack tip is closer to the FB line when the load condition is constant.
{"title":"Mechanical Properties Evaluation and Crack Propagation Behavior in Dissimilar Metal Welded Joints of 304 L Austenitic Stainless Steel and SA508 Low-Alloy Steel","authors":"Yuman Sun, H. Xue, Fu-qiang Yang, Shuai Wang, Shun Zhang, Jinxuan He, R. Bashir","doi":"10.1155/2022/3038397","DOIUrl":"https://doi.org/10.1155/2022/3038397","url":null,"abstract":"The material mechanical properties and crack propagation behavior of dissimilar metal welded joint (DMWJ) of pressurized water reactor (PWR) was investigated. In this research, the mechanical parameters of the cladding layer materials (304L-SA508) of the DMWJ in PWRs were obtained by the continuous indentation test. Simultaneously, the user-defined (USDFLD) subroutine in ABAQUS was used to establish the heterogeneous materials model of the welded joint. On this basis, the local crack propagation path of DMWJs has been discussed based on the extended finite element method (XFEM). The result indicated that the strength value at the fusion boundary line (FB line) is the largest, and the yield strength reaches 689 MPa. The yield stress values of the cladding metal (304 L) and base metal (SA508) are 371 MPa and 501 MPa, respectively. Affected by the material constraint effect of the DMWJ, the crack will propagate through the FB line when the initial crack is perpendicular to the FB line. And when the initial crack parallels the FB line, the crack will deviate from it. Meanwhile, the crack propagation length is smaller as the initial crack tip is closer to the FB line when the load condition is constant.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45021712","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Izza Shahid, Nadeem Shaukat, Amjad Ali, Meer Bacha, Ammar Ahmad, M. Siddique, R. Khan, Sajjad Tahir, Zeeshan Jamil
A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.
{"title":"Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION","authors":"Izza Shahid, Nadeem Shaukat, Amjad Ali, Meer Bacha, Ammar Ahmad, M. Siddique, R. Khan, Sajjad Tahir, Zeeshan Jamil","doi":"10.1155/2022/6319628","DOIUrl":"https://doi.org/10.1155/2022/6319628","url":null,"abstract":"A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2022-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44343134","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
High-temperature superconducting material is a promising candidate to fabricate superconducting magnet for magnetic confinement fusion reactors. The DPA number of the 1 µm thick superconducting layer in a high temperature superconducting tape under neutron irradiation needs to be calculated to predict the property changes. The DPA cross sections, which ignore the spatial distribution of vacancies caused by PKAs, are commonly used to obtain the results of the damage energy and DPA. However, for geometric models with the thickness as small as 1 µm, the energy and angular distribution of PKAs reveal that a significant number of PKAs with relatively high energy tend to scatter forward and cross the boundary of model, so the thickness of model has the potential to affect the number of displaced atoms. In this paper, we developed a method based on Geant4 and SRIM to evaluate the deviation of the traditional analytic method caused by the thickness. Geant4 is used to obtain the location, direction, and energy of PKAs, while SRIM is used to track every PKA and obtain damage energy and the number of displaced atoms. The radiation damage calculation of simple thin plate models with different thicknesses and the tape model are conducted with the neutron energies from 1 to 14 MeV. The results show that PKAs need to be tracked continuously for models with thickness less than 10 µm and the deviation of the analytic formulas increases rapidly with the decrease of thickness. For the superconducting layer composed of four different elements in the tape, the deviation also depends on the proportion of each atomic species and the neutron-atom interaction cross sections under different incident neutron energy.
{"title":"Study on the Applicability of Neutron Radiation Damage Method Used for High-Temperature Superconducting Tape Based on Geant4 and SRIM","authors":"Ying-Ying Zheng, Jinxing Zheng, Xudong Wang","doi":"10.1155/2021/2839746","DOIUrl":"https://doi.org/10.1155/2021/2839746","url":null,"abstract":"High-temperature superconducting material is a promising candidate to fabricate superconducting magnet for magnetic confinement fusion reactors. The DPA number of the 1 µm thick superconducting layer in a high temperature superconducting tape under neutron irradiation needs to be calculated to predict the property changes. The DPA cross sections, which ignore the spatial distribution of vacancies caused by PKAs, are commonly used to obtain the results of the damage energy and DPA. However, for geometric models with the thickness as small as 1 µm, the energy and angular distribution of PKAs reveal that a significant number of PKAs with relatively high energy tend to scatter forward and cross the boundary of model, so the thickness of model has the potential to affect the number of displaced atoms. In this paper, we developed a method based on Geant4 and SRIM to evaluate the deviation of the traditional analytic method caused by the thickness. Geant4 is used to obtain the location, direction, and energy of PKAs, while SRIM is used to track every PKA and obtain damage energy and the number of displaced atoms. The radiation damage calculation of simple thin plate models with different thicknesses and the tape model are conducted with the neutron energies from 1 to 14 MeV. The results show that PKAs need to be tracked continuously for models with thickness less than 10 µm and the deviation of the analytic formulas increases rapidly with the decrease of thickness. For the superconducting layer composed of four different elements in the tape, the deviation also depends on the proportion of each atomic species and the neutron-atom interaction cross sections under different incident neutron energy.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2021-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44231437","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The nickel-base superalloy Hastelloy N was irradiated using 1 MeV Xe20+ and 7 MeV Xe26+ ions with displacement damage ranging from 0.5 dPa to 10 dPa at room temperature (RT). The irradiated Ni-based superalloy was characterized with transmission electron microscopy (TEM), XRD, and nanoindenter to determine the changes in microstructural evolution and nanoindentation hardness. The TEM results showed that ion irradiation induced a large number of defects such as black spots and corrugated structures and the second phase was rapidly amorphized after being irradiated to a fluence of 0.5 dPa. The XRD results showed that the Hastelloy N alloy sample did not undergo lattice distortion after ion irradiation. An obvious irradiation hardening phenomenon was observed in this study, and the hardness increased with Xe ion fluence. The pinning effect in which the defects can become obstacles to the free movement of dislocation may be responsible for the irradiation-induced hardening.
{"title":"Xe Ion Irradiation-Induced Microstructural Evolution and Hardening Effect of Nickel-Base Superalloy","authors":"Yu Hou, DeHui Li, Yan-Er Lu, Hefei Huang, WeiGuo Yang, Renduo Liu","doi":"10.1155/2021/5535478","DOIUrl":"https://doi.org/10.1155/2021/5535478","url":null,"abstract":"The nickel-base superalloy Hastelloy N was irradiated using 1 MeV Xe20+ and 7 MeV Xe26+ ions with displacement damage ranging from 0.5 dPa to 10 dPa at room temperature (RT). The irradiated Ni-based superalloy was characterized with transmission electron microscopy (TEM), XRD, and nanoindenter to determine the changes in microstructural evolution and nanoindentation hardness. The TEM results showed that ion irradiation induced a large number of defects such as black spots and corrugated structures and the second phase was rapidly amorphized after being irradiated to a fluence of 0.5 dPa. The XRD results showed that the Hastelloy N alloy sample did not undergo lattice distortion after ion irradiation. An obvious irradiation hardening phenomenon was observed in this study, and the hardness increased with Xe ion fluence. The pinning effect in which the defects can become obstacles to the free movement of dislocation may be responsible for the irradiation-induced hardening.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2021-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46305549","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In the planning and management of the interim storage of spent nuclear fuel, the technical and economic parameters that are involved have a significant role in increasing the efficiency of the storage system. Optimal parameters will reduce the total economic costs for countries embarking on nuclear energy, such as the UAE. This study evaluated the design performance and economic feasibility of various structures and schedules, to determine an optimal combination of parameters for the management of spent nuclear fuel. With the introduction of various storage technology arrangements and expected costs per unit for the storage system design, we evaluated eight major scenarios, each with a cost analysis based on technological and economic issues. We executed a number of calculations based on the use of these storage technologies, and considered their investment costs. These calculations, which were aligned with the net present value approach and conducted using MS Project and MATLAB software programs, considered the capacities of the spent fuel pools and the amount of spent nuclear fuel (SNF) that will be transferred to dry storage facilities. As soon as they sufficiently cool, the spent nuclear fuel is to be stored in a pool storage facility. The results show that applying a centralized dry storage (CDS) system strategy is not an economically feasible solution, compared with using a permanent disposal facility (PDF) (unless the variable investment cost is reduced or changed). The optimal strategy involves operating a spent fuel pool island (SFPI) storage after the first 20 years of the start of the permanent shutdown of the reactor. After 20 years, the spent fuel is then transferred to a PDF. This strategy also results in a 20.9% to 26.1% reduction in the total cost compared with those of the other strategies. The total cost of the proposed strategy is approximately 4,307 million USD. The duration of the fuel storage and the investment cost, particularly the variable investment cost, directly affect the choice of facility storage.
{"title":"Technical Options and Cost Estimates for Spent Nuclear Fuel Management at the Barakah Nuclear Power Plants","authors":"Shadwan M. M. Esmail, J. Cheong","doi":"10.1155/2021/3133433","DOIUrl":"https://doi.org/10.1155/2021/3133433","url":null,"abstract":"In the planning and management of the interim storage of spent nuclear fuel, the technical and economic parameters that are involved have a significant role in increasing the efficiency of the storage system. Optimal parameters will reduce the total economic costs for countries embarking on nuclear energy, such as the UAE. This study evaluated the design performance and economic feasibility of various structures and schedules, to determine an optimal combination of parameters for the management of spent nuclear fuel. With the introduction of various storage technology arrangements and expected costs per unit for the storage system design, we evaluated eight major scenarios, each with a cost analysis based on technological and economic issues. We executed a number of calculations based on the use of these storage technologies, and considered their investment costs. These calculations, which were aligned with the net present value approach and conducted using MS Project and MATLAB software programs, considered the capacities of the spent fuel pools and the amount of spent nuclear fuel (SNF) that will be transferred to dry storage facilities. As soon as they sufficiently cool, the spent nuclear fuel is to be stored in a pool storage facility. The results show that applying a centralized dry storage (CDS) system strategy is not an economically feasible solution, compared with using a permanent disposal facility (PDF) (unless the variable investment cost is reduced or changed). The optimal strategy involves operating a spent fuel pool island (SFPI) storage after the first 20 years of the start of the permanent shutdown of the reactor. After 20 years, the spent fuel is then transferred to a PDF. This strategy also results in a 20.9% to 26.1% reduction in the total cost compared with those of the other strategies. The total cost of the proposed strategy is approximately 4,307 million USD. The duration of the fuel storage and the investment cost, particularly the variable investment cost, directly affect the choice of facility storage.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2021-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48535187","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.
{"title":"Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models","authors":"Si-Chan Lyu, D. Lu, D. Sui","doi":"10.1155/2021/5843910","DOIUrl":"https://doi.org/10.1155/2021/5843910","url":null,"abstract":"The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2021-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48885091","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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