Pellet-to-cladding mechanical interaction is an important physical phenomenon during reactor power change as well as a multi-phenomenal fuel rod failure mechanism involving stress, strain and material irradiation properties. In order to avoid the failure caused by PCMI, a large number of power ramp tests have been carried out by international organizations over the past decades. The typical way of power ramp test in a high temperature and pressure loop requires stringent test capabilities and high test costs. The key parameters of the PCMI phenomena are stress and strain of cladding, which are generally chosen as the evaluation indicators of PCMI. If it is possible to simulate the pellet-to-cladding contact state, i.e. the gap between pellet and cladding under basic irradiation power level in atmospheric pressure and ordinary temperature environment, then the high stress and strain state of the cladding during the subsequent power ramp test could also be simulated in the same environment, which means lower test costs and test loop requirements. Therefore, a sensitivity analysis using the fuel rod performance analysis code RoPE, was carried out on factors such as initial pellet-to-cladding gap and fuel densification in the power ramp test rod design. By adapting the manufacturing parameters of the test rod and coolant conditions, the high stress and strain state of the cladding could be simulated in the test environment at normal temperature and pressure. The sensitivity analysis provides a theoretical basis for conducting power ramp tests in an atmospheric pressure and ordinary temperature loop.
{"title":"Feasibility Study on Power Ramp Test Under Atmospheric Pressure and Ordinary Temperature","authors":"Xiangyu Wei, Wenhua Zhang, Yingchun Zhao","doi":"10.1115/icone29-93417","DOIUrl":"https://doi.org/10.1115/icone29-93417","url":null,"abstract":"\u0000 Pellet-to-cladding mechanical interaction is an important physical phenomenon during reactor power change as well as a multi-phenomenal fuel rod failure mechanism involving stress, strain and material irradiation properties. In order to avoid the failure caused by PCMI, a large number of power ramp tests have been carried out by international organizations over the past decades. The typical way of power ramp test in a high temperature and pressure loop requires stringent test capabilities and high test costs. The key parameters of the PCMI phenomena are stress and strain of cladding, which are generally chosen as the evaluation indicators of PCMI. If it is possible to simulate the pellet-to-cladding contact state, i.e. the gap between pellet and cladding under basic irradiation power level in atmospheric pressure and ordinary temperature environment, then the high stress and strain state of the cladding during the subsequent power ramp test could also be simulated in the same environment, which means lower test costs and test loop requirements. Therefore, a sensitivity analysis using the fuel rod performance analysis code RoPE, was carried out on factors such as initial pellet-to-cladding gap and fuel densification in the power ramp test rod design. By adapting the manufacturing parameters of the test rod and coolant conditions, the high stress and strain state of the cladding could be simulated in the test environment at normal temperature and pressure. The sensitivity analysis provides a theoretical basis for conducting power ramp tests in an atmospheric pressure and ordinary temperature loop.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"46 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83728750","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}
Feng Wen, Bolong Guo, Mingyang Li, Jiachen Wang, Fan Zhang, W. Liu, Wenkai Ma, Zhihua Liu
Since uranium carbide (UC) fuel has higher thermal conductivity and uranium density than UO2 fuel, good inherent safety and economy can be achieved by UC dispersion fuel. This paper focuses on the research of UC phase control during sol-gel preparation of UC ceramic microsphere. Using nano-carbon black powder as carbon source, UC ceramic microspheres were prepared by sol-gel process combined with carbothermal reaction process. Specific amount of carbon black was uniformly added into the glue solution, which was then dispersed to prepare carbon ADU gel spheres. With heat treatment of carbon ADU gel spheres subject to carbothermic reaction, ceramic microspheres containing UC phase were obtained. During the experiment, the effects of different carbon black addition ratio, thickener addition level and dispersion process on the chemical composition, sphericity and surface morphology on UC ceramic microspheres were investigated. The effects of heat treatment temperature, atmosphere and pressure on the preparation of UC microspheres were also studied during the carbothermic experiment. The process parameters for UC phase control were determined through the experiment and XRD, and the ceramic microsphere samples containing the UC phase were obtained as well. Analysis results show that crack-free UC microspheres with smooth surface can be obtained after drying when sol solution C/U is 2.5∼6.0 and thickener addition level is 1.2∼2.0 kg/kgU. By reaction treatment at high temperature, ceramic microspheres with good sphericity, compact structure and UC phase composition were obtained. A preparation method of UC ceramic microspheres was developed as a result.
{"title":"Study on UC Phase Control of UC Ceramic Microspheres Prepared by Sol-Gel Method","authors":"Feng Wen, Bolong Guo, Mingyang Li, Jiachen Wang, Fan Zhang, W. Liu, Wenkai Ma, Zhihua Liu","doi":"10.1115/icone29-92188","DOIUrl":"https://doi.org/10.1115/icone29-92188","url":null,"abstract":"\u0000 Since uranium carbide (UC) fuel has higher thermal conductivity and uranium density than UO2 fuel, good inherent safety and economy can be achieved by UC dispersion fuel. This paper focuses on the research of UC phase control during sol-gel preparation of UC ceramic microsphere. Using nano-carbon black powder as carbon source, UC ceramic microspheres were prepared by sol-gel process combined with carbothermal reaction process. Specific amount of carbon black was uniformly added into the glue solution, which was then dispersed to prepare carbon ADU gel spheres. With heat treatment of carbon ADU gel spheres subject to carbothermic reaction, ceramic microspheres containing UC phase were obtained. During the experiment, the effects of different carbon black addition ratio, thickener addition level and dispersion process on the chemical composition, sphericity and surface morphology on UC ceramic microspheres were investigated. The effects of heat treatment temperature, atmosphere and pressure on the preparation of UC microspheres were also studied during the carbothermic experiment. The process parameters for UC phase control were determined through the experiment and XRD, and the ceramic microsphere samples containing the UC phase were obtained as well. Analysis results show that crack-free UC microspheres with smooth surface can be obtained after drying when sol solution C/U is 2.5∼6.0 and thickener addition level is 1.2∼2.0 kg/kgU. By reaction treatment at high temperature, ceramic microspheres with good sphericity, compact structure and UC phase composition were obtained. A preparation method of UC ceramic microspheres was developed as a result.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"5 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73094391","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}
A monolithic plate-type fuel system has been under development to convert high-performance test reactors from highly enriched uranium to low-enrichment uranium fuels and is now moving into the qualification phase, a predecessor to the timely conversion of the target reactors. To qualify this fuel system, the plates must meet the safety standards and perform well in a reactor. The plates must maintain mechanical integrity, exhibit geometric stability, and have stable and predictable in-reactor behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by successful testing. However, each high-performance reactor employs distinct design, resulting in distinct plate geometries, with unique features, attributes, irregularities, and tolerances. Due to the abundance of such distinct geometric varieties, a single “generic” plate geometry capturing all the extremes is not achievable. It is also impractical to test each of these proposed designs in a reactor. This limitation necessitates cautious evaluations since the thermo-mechanical response of a plate with a certain geometry may not be representative for a plate with a significantly different geometry. To address concerns related to in-reactor performance of the plates, large set of sensitivity studies were performed. These parametric studies aimed to better understand irradiation performance, while evaluating the sensitivity of results to various modeling inputs, including geometric, operational, and material parameters. This work studied selected geometric parameters based on provided fuel specifications and performed a series of parametric simulations. The resulting temperature, displacement and stress-strains were comparatively evaluated to determine the effects of various geometric parameters. This draft provides a “high-level summary” of parametric sensitivity studies performed and summarizes the key findings from those studies.
{"title":"Summary of Geometric Parameters and Their Effects on Performance of U-10Mo Fuel Plates","authors":"H. Ozaltun, H. Roh, W. Mohamed","doi":"10.1115/icone29-93700","DOIUrl":"https://doi.org/10.1115/icone29-93700","url":null,"abstract":"\u0000 A monolithic plate-type fuel system has been under development to convert high-performance test reactors from highly enriched uranium to low-enrichment uranium fuels and is now moving into the qualification phase, a predecessor to the timely conversion of the target reactors. To qualify this fuel system, the plates must meet the safety standards and perform well in a reactor. The plates must maintain mechanical integrity, exhibit geometric stability, and have stable and predictable in-reactor behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by successful testing. However, each high-performance reactor employs distinct design, resulting in distinct plate geometries, with unique features, attributes, irregularities, and tolerances. Due to the abundance of such distinct geometric varieties, a single “generic” plate geometry capturing all the extremes is not achievable. It is also impractical to test each of these proposed designs in a reactor. This limitation necessitates cautious evaluations since the thermo-mechanical response of a plate with a certain geometry may not be representative for a plate with a significantly different geometry. To address concerns related to in-reactor performance of the plates, large set of sensitivity studies were performed. These parametric studies aimed to better understand irradiation performance, while evaluating the sensitivity of results to various modeling inputs, including geometric, operational, and material parameters. This work studied selected geometric parameters based on provided fuel specifications and performed a series of parametric simulations. The resulting temperature, displacement and stress-strains were comparatively evaluated to determine the effects of various geometric parameters. This draft provides a “high-level summary” of parametric sensitivity studies performed and summarizes the key findings from those studies.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"45 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73827670","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}
As the technology of small modular reactor (SMR) becomes mature, its safety is getting more attention. In recent research of accident tolerant fuel (ATF) such as the U3Si2 used in a large commercial pressurized water reactor can significantly improve the safety of the core due to its high thermal conductivity and good physicochemical properties. Therefore, applying the ATF in the SMR can also enhance its safety. The mPower with a thermal power output of 530 MW designed by B&W company is selected as one of the SMR. In the design and production process of fuel rods, the design and manufacturing parameters such as fuel pellet diameter, inner diameter of cladding, fuel pellet density and other parameters have uncertain factors. These factors affect the results of effective multiplication factor (keff). In this work, we carried out preliminary sensitivity and uncertainty analysis of the U3Si2 with accident tolerant cladding (FeCrAl, SiC) in mPower using SEPRENT and DAKOTA. Sensitivity analysis performed the relative importance of each input parameters for different ATF combination. Uncertainty analysis gave the tolerance range and uncertainty for different ATF combination.
{"title":"Preliminary Sensitivity and Uncertainty Analysis of Accident Tolerant Fuel in SMR","authors":"Yuxuan Liang, Xiang Wang","doi":"10.1115/icone29-92887","DOIUrl":"https://doi.org/10.1115/icone29-92887","url":null,"abstract":"\u0000 As the technology of small modular reactor (SMR) becomes mature, its safety is getting more attention. In recent research of accident tolerant fuel (ATF) such as the U3Si2 used in a large commercial pressurized water reactor can significantly improve the safety of the core due to its high thermal conductivity and good physicochemical properties. Therefore, applying the ATF in the SMR can also enhance its safety. The mPower with a thermal power output of 530 MW designed by B&W company is selected as one of the SMR. In the design and production process of fuel rods, the design and manufacturing parameters such as fuel pellet diameter, inner diameter of cladding, fuel pellet density and other parameters have uncertain factors. These factors affect the results of effective multiplication factor (keff). In this work, we carried out preliminary sensitivity and uncertainty analysis of the U3Si2 with accident tolerant cladding (FeCrAl, SiC) in mPower using SEPRENT and DAKOTA. Sensitivity analysis performed the relative importance of each input parameters for different ATF combination. Uncertainty analysis gave the tolerance range and uncertainty for different ATF combination.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"2 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74020643","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}
As one of the six nuclear reactor candidates selected by the Generation IV International Forum (GIF), lead-cooled fast reactor (LFR) has become one of the most promising concepts and attracted more attention from the institutes. A lead-bismuth eutectic (LBE) cooled fast reactor called BLESS-D has been proposed by State Power Investment Corporation Research Institute of China. The fuel management is critical to reactor design because it affects reactor economics. Although the research on fuel management of PWR has matured, there are few studies on fuel management for LFR. Fuel management design includes the entire process from initial cycle to transition cycle, and to equilibrium cycle. To obtain a better refueling scheme and optimize the key parameters of equilibrium cycle, a new refueling scheme is first proposed, and then traditionally, the simulation covering the entire refueling operation is required cycle by cycle, resulting in consuming a massive computing resource. If the scheme fails to meet expectations, expenditure will multiply until the modified scheme meets the safety design criteria. Fuel management design using “pseudo-equilibrium cycle” instead of traditional method is carried out in this work. The “pseudo-equilibrium cycle” method can directly solve the core key parameters of equilibrium cycle by replacing fuel model with approximate nuclides densities estimated from initial core arrangement and refueling scheme. In this paper, a two-batch refueling scheme is proposed with “pseudo-equilibrium cycle” method and then transition cycle are designed to ensure the feasibility that the fresh core successfully transition to the “pseudo-equilibrium cycle” state. Afterwards, neutronics parameters are solved for each cycle from the fresh core and results show that when the burnup calculation comes to the 5th cycle, the reactor key parameters, including assembly peaking factor, linear power density, delayed neutron fraction, and prompt neutron lifetime are in good agreement with “pseudo-equilibrium cycle”, which proves that the “pseudo-equilibrium cycle” method can be used accurately and efficiently to design the refueling scheme.
{"title":"Fuel Management Strategy of LBE-Cooled Fast Reactor BLESS-D","authors":"Zhen Luo, Eing Yee Yeoh, Xiaosong Chen, Linsen Li, Mian Xing","doi":"10.1115/icone29-90443","DOIUrl":"https://doi.org/10.1115/icone29-90443","url":null,"abstract":"\u0000 As one of the six nuclear reactor candidates selected by the Generation IV International Forum (GIF), lead-cooled fast reactor (LFR) has become one of the most promising concepts and attracted more attention from the institutes. A lead-bismuth eutectic (LBE) cooled fast reactor called BLESS-D has been proposed by State Power Investment Corporation Research Institute of China. The fuel management is critical to reactor design because it affects reactor economics. Although the research on fuel management of PWR has matured, there are few studies on fuel management for LFR. Fuel management design includes the entire process from initial cycle to transition cycle, and to equilibrium cycle. To obtain a better refueling scheme and optimize the key parameters of equilibrium cycle, a new refueling scheme is first proposed, and then traditionally, the simulation covering the entire refueling operation is required cycle by cycle, resulting in consuming a massive computing resource. If the scheme fails to meet expectations, expenditure will multiply until the modified scheme meets the safety design criteria. Fuel management design using “pseudo-equilibrium cycle” instead of traditional method is carried out in this work. The “pseudo-equilibrium cycle” method can directly solve the core key parameters of equilibrium cycle by replacing fuel model with approximate nuclides densities estimated from initial core arrangement and refueling scheme. In this paper, a two-batch refueling scheme is proposed with “pseudo-equilibrium cycle” method and then transition cycle are designed to ensure the feasibility that the fresh core successfully transition to the “pseudo-equilibrium cycle” state. Afterwards, neutronics parameters are solved for each cycle from the fresh core and results show that when the burnup calculation comes to the 5th cycle, the reactor key parameters, including assembly peaking factor, linear power density, delayed neutron fraction, and prompt neutron lifetime are in good agreement with “pseudo-equilibrium cycle”, which proves that the “pseudo-equilibrium cycle” method can be used accurately and efficiently to design the refueling scheme.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"98 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83589454","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}
IN617 was considered the primary material candidate for the IHX in the VHTR. Researching microstructure evolution during high-temperature creep of IN617 helped understand its fracture laws and guide IHX operation under creep loading. Creep tests of IN617 were conducted under 19 MPa, 24 MPa, 27 MPa, and 38 MPa at 950 °C. Creep rupture mechanisms of IN617 were discussed by correlating creep performance, microstructure characteristics and fracture-surface morphology. The results indicated that DRX, creep voids and brittle-phase precipitation were found under different stresses during microstructure observation, which would cause the specimen ductile rupture, intergranular rupture and brittle rupture, respectively. Specifically, under the highest stress 38 MPa, DRX occurred and grain size was decreased greatly to 11.1 μm from 99.7 μm. Fine grains were easy to migrate, causing significant plastic deformation and ductile rupture of specimens. With stress decreased to 27 MPa, grain boundaries became vulnerable and intergranular rupture occurred because intergranular carbides dissolved and their pinning effect was weakened. As stresses were lowered to 24 MPa and 19 MPa, nitrogen was diffused into specimens and brittle nitrides precipitated into continuous networks along GBs. The internal cracking of nitride networks caused brittle rupture. Meanwhile, steady creep rates were increased, and creep rupture lives were shortened greatly, especially under 19 MPa.
{"title":"Creep Performance and Microstructure Evolution of IN617 at 950 °C for VHTR Applications","authors":"Yue Wang, Haitao Wang, Kejian Li","doi":"10.1115/icone29-88939","DOIUrl":"https://doi.org/10.1115/icone29-88939","url":null,"abstract":"\u0000 IN617 was considered the primary material candidate for the IHX in the VHTR. Researching microstructure evolution during high-temperature creep of IN617 helped understand its fracture laws and guide IHX operation under creep loading. Creep tests of IN617 were conducted under 19 MPa, 24 MPa, 27 MPa, and 38 MPa at 950 °C. Creep rupture mechanisms of IN617 were discussed by correlating creep performance, microstructure characteristics and fracture-surface morphology. The results indicated that DRX, creep voids and brittle-phase precipitation were found under different stresses during microstructure observation, which would cause the specimen ductile rupture, intergranular rupture and brittle rupture, respectively. Specifically, under the highest stress 38 MPa, DRX occurred and grain size was decreased greatly to 11.1 μm from 99.7 μm. Fine grains were easy to migrate, causing significant plastic deformation and ductile rupture of specimens. With stress decreased to 27 MPa, grain boundaries became vulnerable and intergranular rupture occurred because intergranular carbides dissolved and their pinning effect was weakened. As stresses were lowered to 24 MPa and 19 MPa, nitrogen was diffused into specimens and brittle nitrides precipitated into continuous networks along GBs. The internal cracking of nitride networks caused brittle rupture. Meanwhile, steady creep rates were increased, and creep rupture lives were shortened greatly, especially under 19 MPa.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"15 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83645192","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}
Fission gas release was modeled by COMSOL Multiphysics in oxide fuels, whose theory is based on the effect of grain boundary network percolation. The model is contributed to the conventional Booth model. In this model, the fuel pellet microstructure adopts 2D axisymmetric geometry. The effect of different bubble growth and coalescence rates on the independent grain boundaries are investigated, such as bubble contact angle, resolution rate, and radial position. The modeled physical phenomena are similar to the existing experiments and can be observed in the fission gas release process. The fission gas release is sensitive to the variations of these parameters. Therefore, the parameters are essential to the fission gas release on the microscopic or macroscopic scales. The long-range percolation on the networked grain boundaries is also considered in this work, but it is not considered in the Booth model. The gas resolution on the grain boundaries is also taken into account. At last, this model’s results will be compared with the outcomes of the Booth model as well as the other models.
{"title":"Fission Gas Release Grain Boundary Network Percolation Mechanistic Studies in Oxide Fuels Based on COMSOL Multiphysics Framework","authors":"Jingyu Guo, Wenzhong Zhou","doi":"10.1115/icone29-92732","DOIUrl":"https://doi.org/10.1115/icone29-92732","url":null,"abstract":"\u0000 Fission gas release was modeled by COMSOL Multiphysics in oxide fuels, whose theory is based on the effect of grain boundary network percolation. The model is contributed to the conventional Booth model. In this model, the fuel pellet microstructure adopts 2D axisymmetric geometry. The effect of different bubble growth and coalescence rates on the independent grain boundaries are investigated, such as bubble contact angle, resolution rate, and radial position. The modeled physical phenomena are similar to the existing experiments and can be observed in the fission gas release process. The fission gas release is sensitive to the variations of these parameters. Therefore, the parameters are essential to the fission gas release on the microscopic or macroscopic scales. The long-range percolation on the networked grain boundaries is also considered in this work, but it is not considered in the Booth model. The gas resolution on the grain boundaries is also taken into account. At last, this model’s results will be compared with the outcomes of the Booth model as well as the other models.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"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":"75325094","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}
Liu Yiqing, Wang Wei, Zhao Xiaodong, Gui Zhong, Chen Jiangtao
The safety of hazardous materials marine transportation is an important aspect of marine transportation safety. As the seventh category of hazardous materials, spent fuel has strong radioactivity and high decay heat, which makes the marine transportation of spent fuel concerned by the national competent department. Risk analysis is based on the identification of risk factors, through qualitative or quantitative analysis methods, to analyze the possible consequences of these risks and the possibility of these consequences, so as to determine the corresponding risk level and formulate targeted risk control measures. The safety of marine transportation depends not only on the safety of the cargo itself, but also on the navigation safety of the ship. On the basis of introducing the steps and methods of risk analysis, this paper selects various factors affecting the safety of spent fuel marine transportation from the aspects of cargo itself and ship navigation safety, and establishes a scientific and reasonable spent fuel marine transportation safety evaluation index system; The risk factors of maritime transportation are evaluated by risk evaluation methods such as expert scoring and risk coordinate diagram, some measures and suggestions are put forward according to the evaluation results.
{"title":"Study on Risk of Nuclear Spent Fuel Marine Transportation","authors":"Liu Yiqing, Wang Wei, Zhao Xiaodong, Gui Zhong, Chen Jiangtao","doi":"10.1115/icone29-92870","DOIUrl":"https://doi.org/10.1115/icone29-92870","url":null,"abstract":"\u0000 The safety of hazardous materials marine transportation is an important aspect of marine transportation safety. As the seventh category of hazardous materials, spent fuel has strong radioactivity and high decay heat, which makes the marine transportation of spent fuel concerned by the national competent department. Risk analysis is based on the identification of risk factors, through qualitative or quantitative analysis methods, to analyze the possible consequences of these risks and the possibility of these consequences, so as to determine the corresponding risk level and formulate targeted risk control measures. The safety of marine transportation depends not only on the safety of the cargo itself, but also on the navigation safety of the ship. On the basis of introducing the steps and methods of risk analysis, this paper selects various factors affecting the safety of spent fuel marine transportation from the aspects of cargo itself and ship navigation safety, and establishes a scientific and reasonable spent fuel marine transportation safety evaluation index system; The risk factors of maritime transportation are evaluated by risk evaluation methods such as expert scoring and risk coordinate diagram, some measures and suggestions are put forward according to the evaluation results.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"32 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85005853","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}
Doctor Enivweru, Qingyu Wang, A. Ayodeji, John Njoroge
The integrity of structural components is a major concern in reactor safety. Cladding materials can fail when hydrogen uptake is in excess of the terminal solid solubility (TSS). Previous studies on hydrogenation or secondary hydriding (SH) focused mainly on loss of coolant accident (LOCA) and fuel ballooning. Most of these studies are carried out in autoclave conditions and they preclude the contribution of ionizing radiations (neutrons and gamma) to hydrogen generation during normal reactor condition. This study is devoted to the hydrogenation of Zircaloy-4 (Zry-4) cladding material during normal reactor operation and during the early phase of core degradation in a LOCA event. Radiolysis and corrosion are considered as hydrogen sources during normal reactor operation while corrosion is considered in the early phase of core degradation. Using appropriate initial and boundary conditions peculiar to pressurized water reactors (PWRs), diffusion equation is solved in a defective Zry-4 fuel system using COMSOL Multiphysics 5.2. Results show that the average corrosive source term (1.279E−4 mol m−3 s−1) is higher than the radiolytic source term (3.6594E−7 mol m−3 s−1) by a factor of 350 during normal reactor operation. By integrating in two kinetics regimes and setting 6 years as the maximum lifetime of Zry-4, the amount of H dissolved in the cladding material at a temperature of 633 K are 6.54E−03 and 2.02 wt. ppm due to radiolysis and corrosion respectively. These values are safe compared to the TSS of dissolution (CTSSD) of Zry-4 which is 117 wt. ppm at the same referenced temperature. However, in the early phase of core degradation where the source term is 9.405E−01 mol m−3 s−1, SH is observed in 13 days. A comparison is made between hydrogenation by integration in two regimes and hydrogenation using Sievert’s law. The results obtained in this study are useful for the prediction of the onset of hydrogen embrittlement, and to determine the service time of Zry-4 cladding material.
{"title":"Hydrogenation of Zircaloy-4 in a Defective Fuel Pellet for Pressurized Water Reactors","authors":"Doctor Enivweru, Qingyu Wang, A. Ayodeji, John Njoroge","doi":"10.1115/icone29-90736","DOIUrl":"https://doi.org/10.1115/icone29-90736","url":null,"abstract":"The integrity of structural components is a major concern in reactor safety. Cladding materials can fail when hydrogen uptake is in excess of the terminal solid solubility (TSS). Previous studies on hydrogenation or secondary hydriding (SH) focused mainly on loss of coolant accident (LOCA) and fuel ballooning. Most of these studies are carried out in autoclave conditions and they preclude the contribution of ionizing radiations (neutrons and gamma) to hydrogen generation during normal reactor condition. This study is devoted to the hydrogenation of Zircaloy-4 (Zry-4) cladding material during normal reactor operation and during the early phase of core degradation in a LOCA event. Radiolysis and corrosion are considered as hydrogen sources during normal reactor operation while corrosion is considered in the early phase of core degradation. Using appropriate initial and boundary conditions peculiar to pressurized water reactors (PWRs), diffusion equation is solved in a defective Zry-4 fuel system using COMSOL Multiphysics 5.2. Results show that the average corrosive source term (1.279E−4 mol m−3 s−1) is higher than the radiolytic source term (3.6594E−7 mol m−3 s−1) by a factor of 350 during normal reactor operation. By integrating in two kinetics regimes and setting 6 years as the maximum lifetime of Zry-4, the amount of H dissolved in the cladding material at a temperature of 633 K are 6.54E−03 and 2.02 wt. ppm due to radiolysis and corrosion respectively. These values are safe compared to the TSS of dissolution (CTSSD) of Zry-4 which is 117 wt. ppm at the same referenced temperature. However, in the early phase of core degradation where the source term is 9.405E−01 mol m−3 s−1, SH is observed in 13 days. A comparison is made between hydrogenation by integration in two regimes and hydrogenation using Sievert’s law. The results obtained in this study are useful for the prediction of the onset of hydrogen embrittlement, and to determine the service time of Zry-4 cladding material.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"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":"76397372","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 order to monitor the radiation effects of advanced fuel element cladding materials, it is necessary to install irradiation supervision tanks in the reactor, which contain kinds of detector foils. And then the detector foils are used to determine the neutron energy spectrum and neutron flux, so as to evaluate the radiation damage of the material. Taking China Experimental Fast Reactor (CEFR) as the research object, seven neutron detector foils were selected and five supervision schemes were designed in 316 stainless steel. The size of the neutron detector, the irradiation time and the corresponding relationship with the radiation damage are calculated by the Monte Carlo code and SPECTER program, and finally the sound monitoring scheme is given. It has guiding significance for irradiation experiments of structural materials, and provides a basis for monitoring the irradiation damage of large commercial fast reactor structures.
{"title":"Design of Radiation Damage Monitoring Scheme for Materials in China Experimental Fast Reactor","authors":"Hu Ziao, Chen Xiao-xian, Chen Xiao-liang","doi":"10.1115/icone29-90146","DOIUrl":"https://doi.org/10.1115/icone29-90146","url":null,"abstract":"\u0000 In order to monitor the radiation effects of advanced fuel element cladding materials, it is necessary to install irradiation supervision tanks in the reactor, which contain kinds of detector foils. And then the detector foils are used to determine the neutron energy spectrum and neutron flux, so as to evaluate the radiation damage of the material. Taking China Experimental Fast Reactor (CEFR) as the research object, seven neutron detector foils were selected and five supervision schemes were designed in 316 stainless steel. The size of the neutron detector, the irradiation time and the corresponding relationship with the radiation damage are calculated by the Monte Carlo code and SPECTER program, and finally the sound monitoring scheme is given. It has guiding significance for irradiation experiments of structural materials, and provides a basis for monitoring the irradiation damage of large commercial fast reactor structures.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"17 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77492713","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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