The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10–250 nm and 33–64 μg·L −1 , respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.
{"title":"Characterization of Groundwater Colloids From the Granitic KURT Site and Their Roles in Radionuclide Migration","authors":"M. Baik, Tae-Jin Park, Hye-Ryun Cho, E. Jung","doi":"10.7733/jnfcwt.2022.025","DOIUrl":"https://doi.org/10.7733/jnfcwt.2022.025","url":null,"abstract":"The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10–250 nm and 33–64 μg·L −1 , respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"52 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85182750","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}
Minseop Kim, Seungrae Lee, Seok Yoon, Min-Kyung Jeon
{"title":"Identification of Mechanical Parameters of Kyeongju Bentonite Based on Artificial Neural Network Technique","authors":"Minseop Kim, Seungrae Lee, Seok Yoon, Min-Kyung Jeon","doi":"10.7733/jnfcwt.2022.022","DOIUrl":"https://doi.org/10.7733/jnfcwt.2022.022","url":null,"abstract":"","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"61 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80547605","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}
{"title":"Countermeasures for Management of Off-site Radioactive Wastes in the Event of a Major Accident at Nuclear Power Plants","authors":"Ji-Min Lee, D. Hong, Hyeong-Ki Shin, Hyun Ki Kim","doi":"10.7733/jnfcwt.2022.023","DOIUrl":"https://doi.org/10.7733/jnfcwt.2022.023","url":null,"abstract":",","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"74 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83768538","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}
In the reactor, the main purpose of radial shielding was to protect the internal structure and staff from radiation damage. The good shielding material must have both a high elastic scattering cross-section and a high absorption cross-section to reduce the radiation level outside the reactor core and ensure the safety of workers and equipment outside the reactor. In this paper, a new type selection scheme of shielding material was proposed, which read the reaction cross-sections of all-natural stable nuclides in the ENDF/B-VII database, calculated the single group cross-sections of various reactions of all nuclides by the integral method according to the neutron energy spectrum. Then ranked the nuclides with different cross-sections as the target, to screen the nuclides with the best neutron and photon shielding effect. In this paper, a typical Monte Carlo particle transport program was used for preliminary verification, which was applied to the energy spectrum of the pressurized water reactor and sodium-cooled fast reactor. The results show that for neutron shielding, photon shielding, and neutron-photon shielding, in the thermal neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 95 %, 98.95 %, and 94.87 % respectively, and in the fast neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 72.5 %, 95.79 %, and 82.5 % respectively. The calculation time of non-dominated sorting was less than 3% of that of the Monte Carlo method. It provided an efficient method for the selection of shielding nuclides in advanced reactors and had engineering application significance.
{"title":"Study on Selection Method of Shielding Nuclides in Nuclear Reactor Based on Non-Dominated Sorting","authors":"Rui Li, Shichang Liu","doi":"10.1115/icone29-90902","DOIUrl":"https://doi.org/10.1115/icone29-90902","url":null,"abstract":"\u0000 In the reactor, the main purpose of radial shielding was to protect the internal structure and staff from radiation damage. The good shielding material must have both a high elastic scattering cross-section and a high absorption cross-section to reduce the radiation level outside the reactor core and ensure the safety of workers and equipment outside the reactor. In this paper, a new type selection scheme of shielding material was proposed, which read the reaction cross-sections of all-natural stable nuclides in the ENDF/B-VII database, calculated the single group cross-sections of various reactions of all nuclides by the integral method according to the neutron energy spectrum. Then ranked the nuclides with different cross-sections as the target, to screen the nuclides with the best neutron and photon shielding effect. In this paper, a typical Monte Carlo particle transport program was used for preliminary verification, which was applied to the energy spectrum of the pressurized water reactor and sodium-cooled fast reactor. The results show that for neutron shielding, photon shielding, and neutron-photon shielding, in the thermal neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 95 %, 98.95 %, and 94.87 % respectively, and in the fast neutron spectrum, the coincidence rates between the non-dominated sorting scheme and the Monte Carlo method were 72.5 %, 95.79 %, and 82.5 % respectively. The calculation time of non-dominated sorting was less than 3% of that of the Monte Carlo method. It provided an efficient method for the selection of shielding nuclides in advanced reactors and had engineering application significance.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"16 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74919524","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}
During fuel storage, lifting fuel assembly is an essential process. The fuel assembly dropping accident may lead to consequences, such as fuel assembly lying on the storage rack, fuel assembly deformation leading to out of control of fuel rod pitch distance, fuel rod breakage and other accidents, and then affect the critical safety of fuel storage system. Therefore, it is necessary to analyze the critical safety of fuel assembly storage under fuel assembly dropping accident. This study analyzes all kinds of accident conditions that may occur under fuel assembly dropping accident, and uses Monte Carlo particle transport code JMCT to simulate the neutron effective multiplication factor (keff) at typical conditions. The results show that the calculated keff of the fuel storage system will not exceed the criteria when the fuel assembly drops on the fuel storage rack without deformation. When fuel assembly is deformed due to dropping from a height, change of fuel rod pitch distance in fuel assembly will affect the moderator-uranium ratio of the fuel assembly storage system, which may lead to an increase in keff in the system. However, the fuel storage system can also will not exceed the criteria when 2000 ppm soluble boron acid is taken credit in the spent fuel pool. Based on the above analysis, the fuel storage system can ensure the subcritical state at different condition of fuel assembly dropping.
{"title":"Criticality Safety Analysis of Fuel Storage Under Fuel Assembly Dropping Accident","authors":"Mingliang Dai, Pengtao Fu, Jun Zhao","doi":"10.1115/icone29-90509","DOIUrl":"https://doi.org/10.1115/icone29-90509","url":null,"abstract":"\u0000 During fuel storage, lifting fuel assembly is an essential process. The fuel assembly dropping accident may lead to consequences, such as fuel assembly lying on the storage rack, fuel assembly deformation leading to out of control of fuel rod pitch distance, fuel rod breakage and other accidents, and then affect the critical safety of fuel storage system. Therefore, it is necessary to analyze the critical safety of fuel assembly storage under fuel assembly dropping accident.\u0000 This study analyzes all kinds of accident conditions that may occur under fuel assembly dropping accident, and uses Monte Carlo particle transport code JMCT to simulate the neutron effective multiplication factor (keff) at typical conditions. The results show that the calculated keff of the fuel storage system will not exceed the criteria when the fuel assembly drops on the fuel storage rack without deformation. When fuel assembly is deformed due to dropping from a height, change of fuel rod pitch distance in fuel assembly will affect the moderator-uranium ratio of the fuel assembly storage system, which may lead to an increase in keff in the system. However, the fuel storage system can also will not exceed the criteria when 2000 ppm soluble boron acid is taken credit in the spent fuel pool. Based on the above analysis, the fuel storage system can ensure the subcritical state at different condition of fuel assembly dropping.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"30 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81138196","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}
The cubic silicon carbide (3C-SiC) has been considered as a candidate structural material for several types of advanced nuclear reactors. The effects of cascade collision on thermal conductivity in symmetrical tilt grain boundary (GB) were studied by Molecular dynamics (MD) simulations. The thermal conductivity of 3C-SiC at Σ5(210)[001] GB was calculated using non-equilibrium molecular dynamics (NEMD) methods. A relatively small simulation unit was used to analyze the effect of different energies of incident PKA (primary knock-on atoms) on the thermal conductivity of 3C-SiC and to compare the results with perfect structure GB system. Finally, the vibrational density of states (VDOS) of atoms in the GB region was calculated to analyze the phonon mismatch at the interface. Calculations show that cascade collisions generated by energetic atoms will result in a decrease in thermal conductivity of the Σ5(210) GB system, but the effect varies in different regions, with a sharp decrease in thermal conductivity and an increase in thermal resistance for the intracrystalline region, while the magnitude of change in either thermal resistance or thermal conductivity is not significant in the GB region. Irradiated model shows a higher GB energy compared to the unirradiated model. For all irradiated models, lattice defects have a significant effect on the thermal conductivity of the GB system, depending on the spatial structure of the GBs. the results of the VDOS analysis suggest that an increase in the degree of atomic lattice mismatch near the interface is responsible for a further increase in the thermal resistance of the irradiated GB system.
{"title":"Molecular Dynamics Study on Thermal Conductivity of Unirradiated and Irradiated Symmetrical Tilt Grain Boundary 3C-SiC","authors":"Ziqi Cai, Qingmin Zhang, Z. Shao, Yuanming Li","doi":"10.1115/icone29-92136","DOIUrl":"https://doi.org/10.1115/icone29-92136","url":null,"abstract":"\u0000 The cubic silicon carbide (3C-SiC) has been considered as a candidate structural material for several types of advanced nuclear reactors. The effects of cascade collision on thermal conductivity in symmetrical tilt grain boundary (GB) were studied by Molecular dynamics (MD) simulations. The thermal conductivity of 3C-SiC at Σ5(210)[001] GB was calculated using non-equilibrium molecular dynamics (NEMD) methods. A relatively small simulation unit was used to analyze the effect of different energies of incident PKA (primary knock-on atoms) on the thermal conductivity of 3C-SiC and to compare the results with perfect structure GB system. Finally, the vibrational density of states (VDOS) of atoms in the GB region was calculated to analyze the phonon mismatch at the interface. Calculations show that cascade collisions generated by energetic atoms will result in a decrease in thermal conductivity of the Σ5(210) GB system, but the effect varies in different regions, with a sharp decrease in thermal conductivity and an increase in thermal resistance for the intracrystalline region, while the magnitude of change in either thermal resistance or thermal conductivity is not significant in the GB region. Irradiated model shows a higher GB energy compared to the unirradiated model. For all irradiated models, lattice defects have a significant effect on the thermal conductivity of the GB system, depending on the spatial structure of the GBs. the results of the VDOS analysis suggest that an increase in the degree of atomic lattice mismatch near the interface is responsible for a further increase in the thermal resistance of the irradiated GB system.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"366 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76428963","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}
Guangdong Liu, Wei-liang Wu, He Zhu, Qingjun Ma, Qipeng Li
According to IAEA regulations, the spent nuclear fuel transportation cask must be able to withstand a free drop from the height of 9 meters on an unyielding surface. The impact limiter of the cask absorbs most of the impact energy and reduces the deceleration loads on the cask and contents. In order to accurately estimate the safety of the cask and the spent nuclear fuel, the calculation of drop acceleration is very important. In this study, based on the absorption of energy by an impact limiter is equal to the work done in crushing a volume, the maximum deceleration is estimated during collision process between cask and the ground. Then using the LS-DYNA computer program to simulate the cask drop behavior of 9 meters free drop. By comparing the results of theoretical analysis, numerical simulation and actual drop tests, the cask maximum acceleration and the impact limiter deformation are in good agreement, which prove the validity of the spent nuclear fuel transportation cask drop analysis. The stress intensity results for 9 meters free drop loading cases satisfy the allowable stress limit criteria as presented in “ASME Boiler and Pressure Vessel Code”. The spent nuclear fuel transportation cask meets the requirements of IAEA in the 9m free drop hypothetical accident conditions.
{"title":"Drop Analyses and Verification of the Spent Nuclear Fuel Transportation Cask Under Hypothetical Accident Conditions","authors":"Guangdong Liu, Wei-liang Wu, He Zhu, Qingjun Ma, Qipeng Li","doi":"10.1115/icone29-91927","DOIUrl":"https://doi.org/10.1115/icone29-91927","url":null,"abstract":"\u0000 According to IAEA regulations, the spent nuclear fuel transportation cask must be able to withstand a free drop from the height of 9 meters on an unyielding surface. The impact limiter of the cask absorbs most of the impact energy and reduces the deceleration loads on the cask and contents. In order to accurately estimate the safety of the cask and the spent nuclear fuel, the calculation of drop acceleration is very important. In this study, based on the absorption of energy by an impact limiter is equal to the work done in crushing a volume, the maximum deceleration is estimated during collision process between cask and the ground. Then using the LS-DYNA computer program to simulate the cask drop behavior of 9 meters free drop. By comparing the results of theoretical analysis, numerical simulation and actual drop tests, the cask maximum acceleration and the impact limiter deformation are in good agreement, which prove the validity of the spent nuclear fuel transportation cask drop analysis. The stress intensity results for 9 meters free drop loading cases satisfy the allowable stress limit criteria as presented in “ASME Boiler and Pressure Vessel Code”. The spent nuclear fuel transportation cask meets the requirements of IAEA in the 9m free drop hypothetical accident conditions.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"139 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86829322","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}
Jingjing Zhao, Shiyin Xu, Wenan Pan, Jingya Sun, Y. Du
As the core component and the first safety barrier of a nuclear power plant, the structural integrity of fuel assemblies (FAs) is an important factor in ensuring the safety of nuclear power. Safety analysis and evaluation of fuel assemblies under seismic conditions is a requirement of nuclear power plant design codes and safety reviews. The mechanical properties of a single FA determine the operational and seismic performance of the core fuel assembly. Due to the complexity of the FAs in terms of structure and boundary conditions, test is the most common method to obtain the mechanical properties. In order to predict the mechanical properties of the fuel assembly, a nonlinear finite element detailed model was developed. The contact and frictional behavior between the fuel rod cladding and the spring/dimple convexity were considered in the modeling to simulate the nonlinear behavior of the fuel rod better. To verify the validity of the model, virtual tests of lateral stiffness test, forced vibration test and impact test were conducted on the model. The comparison with the experimental results shows that the model can reflect the mechanical properties of the fuel assembly well. The modeling method and calculation conclusions in this paper are of guidance for fuel assembly design.
{"title":"A Detailed Model to Predict Mechanical Characteristics of Fuel Assembly","authors":"Jingjing Zhao, Shiyin Xu, Wenan Pan, Jingya Sun, Y. Du","doi":"10.1115/icone29-91931","DOIUrl":"https://doi.org/10.1115/icone29-91931","url":null,"abstract":"\u0000 As the core component and the first safety barrier of a nuclear power plant, the structural integrity of fuel assemblies (FAs) is an important factor in ensuring the safety of nuclear power. Safety analysis and evaluation of fuel assemblies under seismic conditions is a requirement of nuclear power plant design codes and safety reviews. The mechanical properties of a single FA determine the operational and seismic performance of the core fuel assembly. Due to the complexity of the FAs in terms of structure and boundary conditions, test is the most common method to obtain the mechanical properties. In order to predict the mechanical properties of the fuel assembly, a nonlinear finite element detailed model was developed. The contact and frictional behavior between the fuel rod cladding and the spring/dimple convexity were considered in the modeling to simulate the nonlinear behavior of the fuel rod better. To verify the validity of the model, virtual tests of lateral stiffness test, forced vibration test and impact test were conducted on the model. The comparison with the experimental results shows that the model can reflect the mechanical properties of the fuel assembly well. The modeling method and calculation conclusions in this paper are of guidance for fuel assembly design.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"48 1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90060064","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}
Hansen Chen, Changyuan Gao, Liu-tao Chen, Xu Wang, J. Tan
The corrosion behavior of zirconium alloys is closely related to the water chemistry and operating temperature of the primary circuit. The increase of operating temperature, Li concentration and pH value will have a significant negative impact on the corrosion behavior of zirconium alloy. Autoclave corrosion tests of CZ and Zr-4 alloys were carried out under different water chemistry conditions (Li concentration and pH300 were 3.5ppm/7.2, 4.5ppm/7.2, 4.5ppm/7.5, 7ppm/7.5 respectively) and temperature conditions (360 °C, 400 °C, 430 °C, 450 °C, 500 °C). The effects of different temperature and water chemistry conditions on the corrosion rate of different zirconium alloys were obtained and the information of oxide film thickness and hydrogen content of different samples in different corrosion stages were obtained to study the temperature and water chemistry adaptability of zirconium alloys. The experimental results show that breakaway corrosion and nodule corrosion occur in Zr-4 alloy above 430 °C, and the thickness of oxide film and hydrogen absorption increase significantly. However, the corrosion law of CZ alloys keep the same from 360 °C to 500 °C, and there is no breakaway corrosion or nodule corrosion. The relationships between corrosion rate and temperature accord with Arrhenius formula, showing good temperature adaptability. Under relatively severe water chemistry conditions (Li concentration and pH300 are 4.5 ppm and 7.5 respectively), the corrosion rate of various zirconium alloys does not increase obviously, and there is no nodule corrosion phenomenon. The increase of oxide film thickness and hydrogen absorption capacity is not obvious, showing good water chemistry adaptability.
{"title":"Study on Water Chemistry and Temperature Adaptability of CZ Zirconium Alloy","authors":"Hansen Chen, Changyuan Gao, Liu-tao Chen, Xu Wang, J. Tan","doi":"10.1115/icone29-93820","DOIUrl":"https://doi.org/10.1115/icone29-93820","url":null,"abstract":"\u0000 The corrosion behavior of zirconium alloys is closely related to the water chemistry and operating temperature of the primary circuit. The increase of operating temperature, Li concentration and pH value will have a significant negative impact on the corrosion behavior of zirconium alloy. Autoclave corrosion tests of CZ and Zr-4 alloys were carried out under different water chemistry conditions (Li concentration and pH300 were 3.5ppm/7.2, 4.5ppm/7.2, 4.5ppm/7.5, 7ppm/7.5 respectively) and temperature conditions (360 °C, 400 °C, 430 °C, 450 °C, 500 °C). The effects of different temperature and water chemistry conditions on the corrosion rate of different zirconium alloys were obtained and the information of oxide film thickness and hydrogen content of different samples in different corrosion stages were obtained to study the temperature and water chemistry adaptability of zirconium alloys. The experimental results show that breakaway corrosion and nodule corrosion occur in Zr-4 alloy above 430 °C, and the thickness of oxide film and hydrogen absorption increase significantly. However, the corrosion law of CZ alloys keep the same from 360 °C to 500 °C, and there is no breakaway corrosion or nodule corrosion. The relationships between corrosion rate and temperature accord with Arrhenius formula, showing good temperature adaptability. Under relatively severe water chemistry conditions (Li concentration and pH300 are 4.5 ppm and 7.5 respectively), the corrosion rate of various zirconium alloys does not increase obviously, and there is no nodule corrosion phenomenon. The increase of oxide film thickness and hydrogen absorption capacity is not obvious, showing good water chemistry adaptability.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"21 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80706703","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}
Reactivity-initiated accident (RIA) is postulated design-basis accidents (DBAs) in light-water reactor (LWR). Moreover, Pellet-cladding mechanical interaction (PCMI) can cause a failure of the cladding in the early transient. In recent RIA regulatory guide released by NRC which points out that PCMI failure threshold depends on total hydrogen content in cladding during RIA. In order to evaluate the performance of independently-developed zirconium alloy under RIA conditions, the mechanical behavior and the fracture of 2 different cladding tubes (CZ, and SR (Stress Relief) Zr-4) with different hydrogen contents are investigated under thermal-mechanical loading conditions representative of PCMI during RIAs. Ring tensile tests are performed at room temperature, 350 °C on 2 different materials containing various hydrogen concentrations up to 1000 wt. ppm. Test results indicate that the ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking, the ductility and strength results of SR Zr-4 have a good agreement with reference paper, confirming the rationality of experimental method applied and reliability of test facilities. According to the results of Zr-4 and CZ, a conclusion can be made is that the ductility of independently-developed zirconium alloy (CZ) is better than Zr-4, which can provide the technical support when licensing.
{"title":"Mechanical Performance Assessment of Independently-Developed Zirconium Alloy Under RIA Condition","authors":"Jie Wang, Yang Wang, Jun Wei, Yong-jun Deng","doi":"10.1115/icone29-90521","DOIUrl":"https://doi.org/10.1115/icone29-90521","url":null,"abstract":"\u0000 Reactivity-initiated accident (RIA) is postulated design-basis accidents (DBAs) in light-water reactor (LWR). Moreover, Pellet-cladding mechanical interaction (PCMI) can cause a failure of the cladding in the early transient. In recent RIA regulatory guide released by NRC which points out that PCMI failure threshold depends on total hydrogen content in cladding during RIA. In order to evaluate the performance of independently-developed zirconium alloy under RIA conditions, the mechanical behavior and the fracture of 2 different cladding tubes (CZ, and SR (Stress Relief) Zr-4) with different hydrogen contents are investigated under thermal-mechanical loading conditions representative of PCMI during RIAs. Ring tensile tests are performed at room temperature, 350 °C on 2 different materials containing various hydrogen concentrations up to 1000 wt. ppm. Test results indicate that the ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking, the ductility and strength results of SR Zr-4 have a good agreement with reference paper, confirming the rationality of experimental method applied and reliability of test facilities. According to the results of Zr-4 and CZ, a conclusion can be made is that the ductility of independently-developed zirconium alloy (CZ) is better than Zr-4, which can provide the technical support when licensing.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"48 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76265253","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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