Abstract In this study, radionuclide behavior in high-level radioactive waste (HLW) disposal repositories is complicated because of the spatial heterogeneity of porous media, coupled flow-transport mechanisms, and multiple chemical reaction processes. Discrepancies in the diffusion behavior of a non-sorbing tracer (HTO) and a reactive tracer (137Cs) in porous media have long been recognized but are not yet fully understood, which hinders effective assessment of the capabilities of buffer materials. This paper was dedicated to exploring and explaining the discrepancies in the transport behavior of non-sorbing and reactive tracers through laboratory experiments and an investigation of contributing mechanisms. Our results showed that for a bentonite sample of the same thickness, 137Cs has smaller apparent and less effective diffusion coefficients than those for HTO. These discrepancies can be attributed to the negative surface electric effects, atomic properties, and chemical reactions. In the case of bentonite samples with different thicknesses (0.5, 0.75, 2.0, 2.5 cm), the apparent and effective diffusion coefficients show an increasing trend with bentonite thickness. According to the experimental data and fitting results, the apparent and effective diffusion coefficients are highly related to bentonite thickness. Thus, scaling effects on transport parameters were proposed to explain the results, which were attributed to the nonuniform distribution of the pore space in the bentonite sample. The scale effect behavior of radionuclide was quantified through a regression analysis. The results can be used to improve buffer designs for radionuclides diffusion.
{"title":"Scaling effect on cesium diffusion in compacted MX-80 bentonite for buffer materials in HLW repository","authors":"Yi-Ling Liu, Tzu-Ting Lin, C. Lee","doi":"10.1515/kern-2022-0122","DOIUrl":"https://doi.org/10.1515/kern-2022-0122","url":null,"abstract":"Abstract In this study, radionuclide behavior in high-level radioactive waste (HLW) disposal repositories is complicated because of the spatial heterogeneity of porous media, coupled flow-transport mechanisms, and multiple chemical reaction processes. Discrepancies in the diffusion behavior of a non-sorbing tracer (HTO) and a reactive tracer (137Cs) in porous media have long been recognized but are not yet fully understood, which hinders effective assessment of the capabilities of buffer materials. This paper was dedicated to exploring and explaining the discrepancies in the transport behavior of non-sorbing and reactive tracers through laboratory experiments and an investigation of contributing mechanisms. Our results showed that for a bentonite sample of the same thickness, 137Cs has smaller apparent and less effective diffusion coefficients than those for HTO. These discrepancies can be attributed to the negative surface electric effects, atomic properties, and chemical reactions. In the case of bentonite samples with different thicknesses (0.5, 0.75, 2.0, 2.5 cm), the apparent and effective diffusion coefficients show an increasing trend with bentonite thickness. According to the experimental data and fitting results, the apparent and effective diffusion coefficients are highly related to bentonite thickness. Thus, scaling effects on transport parameters were proposed to explain the results, which were attributed to the nonuniform distribution of the pore space in the bentonite sample. The scale effect behavior of radionuclide was quantified through a regression analysis. The results can be used to improve buffer designs for radionuclides diffusion.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"16 1","pages":"253 - 261"},"PeriodicalIF":0.5,"publicationDate":"2023-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74680458","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}
Abstract Flow-induced vibrations in nuclear power plants may lead to material fatigue, fretting wear, and eventually to loss of component integrity. The consequences might be substantial costs due to long unplanned outages or a fault that requires safety provisions to perform as intended. To avoid these, Fluid-Structure Interaction analyses are performed to understand and predict the complex thermal-hydraulic and structural mechanics phenomena. To further advance the knowledge of solving FSI problems with the help of numerical tools, in the beginning of 2020, the joint industry VIKING project was established in Europe. Further, OECD/NEA initiated in 2021 an FSI Benchmark on FIV that should be finished by the end of 2022 and the final synthesis report should be published in 2023. This paper provides a short overview of the GRS contributions within these two international activities on the prediction of FIV in nuclear power reactors. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.
{"title":"GRS contributions to flow-induced vibrations related activities in Europe","authors":"A. Papukchiev","doi":"10.1515/kern-2022-0110","DOIUrl":"https://doi.org/10.1515/kern-2022-0110","url":null,"abstract":"Abstract Flow-induced vibrations in nuclear power plants may lead to material fatigue, fretting wear, and eventually to loss of component integrity. The consequences might be substantial costs due to long unplanned outages or a fault that requires safety provisions to perform as intended. To avoid these, Fluid-Structure Interaction analyses are performed to understand and predict the complex thermal-hydraulic and structural mechanics phenomena. To further advance the knowledge of solving FSI problems with the help of numerical tools, in the beginning of 2020, the joint industry VIKING project was established in Europe. Further, OECD/NEA initiated in 2021 an FSI Benchmark on FIV that should be finished by the end of 2022 and the final synthesis report should be published in 2023. This paper provides a short overview of the GRS contributions within these two international activities on the prediction of FIV in nuclear power reactors. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"16 1","pages":"155 - 173"},"PeriodicalIF":0.5,"publicationDate":"2023-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87349238","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}
Abstract Accelerator Driven sub-critical System (ADS), which employs the high-energy proton beam generated by accelerator to bombard the target nucleus and generate spallation neutrons as external neutrons to drive and maintain the operation of its sub-critical reactor, is of great significance in nuclear waste treatment and disposal. As the instability of proton beam would affect the power level of the reactor and threaten the safety of ADS, Beam Trip (BT) and Beam OverPower (BOP) are commonly considered to be its two typical transient accidents. As for the sub-critical reactor, the Transient OverPower (TOP) is also one of typical transient accidents that should be considered, which is mainly caused by reactivity insertion under certain cases, such as SGTR (Steam Generator Tube Rupture) accident. For the subcritical reactors, the transient evolution behaviors are strongly affected by the subcriticality value. On the one hand, the subcriticality values of ADS design should take safety margin and power gain into consideration. On the other hand, the subcriticality value is variable with the burnup of reactors. So it is necessary to study the safety characteristics of the subcritical reactors under different subcriticality values, in this paper, the transient safety characteristics of a single channel for XADS under BT, BOP and TOP accidents of different subcriticality values were investigated by using MPC-LBE code.
{"title":"Study on the accidents analyses of a single channel for XADS by using MPC-LBE code","authors":"Ling Zhang, Tianxin Song, Zhi-Xing Gu, Jianing Dai, Wenlan Ou, Qiwen Pan, Zhengyu Gong","doi":"10.1515/kern-2022-0085","DOIUrl":"https://doi.org/10.1515/kern-2022-0085","url":null,"abstract":"Abstract Accelerator Driven sub-critical System (ADS), which employs the high-energy proton beam generated by accelerator to bombard the target nucleus and generate spallation neutrons as external neutrons to drive and maintain the operation of its sub-critical reactor, is of great significance in nuclear waste treatment and disposal. As the instability of proton beam would affect the power level of the reactor and threaten the safety of ADS, Beam Trip (BT) and Beam OverPower (BOP) are commonly considered to be its two typical transient accidents. As for the sub-critical reactor, the Transient OverPower (TOP) is also one of typical transient accidents that should be considered, which is mainly caused by reactivity insertion under certain cases, such as SGTR (Steam Generator Tube Rupture) accident. For the subcritical reactors, the transient evolution behaviors are strongly affected by the subcriticality value. On the one hand, the subcriticality values of ADS design should take safety margin and power gain into consideration. On the other hand, the subcriticality value is variable with the burnup of reactors. So it is necessary to study the safety characteristics of the subcritical reactors under different subcriticality values, in this paper, the transient safety characteristics of a single channel for XADS under BT, BOP and TOP accidents of different subcriticality values were investigated by using MPC-LBE code.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"139 1","pages":"240 - 250"},"PeriodicalIF":0.5,"publicationDate":"2023-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75646003","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}
S. Elnaggar, Samaa. A. Wasfy, S. Abdel-Latif, A. Refaey
Abstract The spent nuclear fuel discharged from power reactors is a very important problem facing the future of using power reactors in electricity production. This paper focuses on the thermal-hydraulic behaviour of the VVER spent fuel in the vault dry storage system under forced convection mode, which is experimentally and numerically investigated. For this purpose, a test rig is designed and constrained to simulate the cooling loop vault system that contains four spent fuel assemblies discharged from the VVER reactor, which are represented by four electric heaters. A numerical simulation is performed by the ANSYS-CFX fluid dynamics code. The effects of decay heat generation and inlet air velocity are investigated as an operating condition. Also, the effect of the type of the Vault System tube material is being studied. The results show that the increase in the inlet air velocity improves the coolability of the fuel, while the increase in decay heat leads to a decrease in the coolability of the fuel. The used velocity range is (0.1 < V < 0.5 m/s) for inlet coolant air and heater power (20 < P < 100 W). Three tube materials (aluminum, copper, and stainless steel) were evaluated for mechanical properties, including thermal conductivity, to assess the feasibility of their use as tubes in the spent fuel storage.
动力堆的乏燃料排放是未来动力堆发电面临的一个非常重要的问题。本文对VVER乏燃料在强制对流模式下在拱顶干蓄系统中的热水力特性进行了实验和数值研究。为此，设计并约束了一个试验台来模拟冷却回路拱顶系统，该系统包含从VVER反应堆排放的四个乏燃料组件，由四个电加热器代表。利用ANSYS-CFX流体动力学软件进行了数值模拟。研究了衰变热产生和入口风速作为运行条件的影响。此外，还研究了拱顶系统管材料类型的影响。结果表明:进气速度的增加提高了燃料的冷却性，而衰变热的增加导致燃料的冷却性降低。入口冷却剂空气的使用速度范围为(0.1 < V < 0.5 m/s)，加热器功率为(20 < P < 100 W)。对三种管道材料(铝、铜和不锈钢)的机械性能(包括导热性)进行了评估，以评估它们作为乏燃料储存管道的可行性。
{"title":"Thermal hydraulic analysis of VVER spent fuels stored in vault dry system under different operating and design conditions","authors":"S. Elnaggar, Samaa. A. Wasfy, S. Abdel-Latif, A. Refaey","doi":"10.1515/kern-2022-0096","DOIUrl":"https://doi.org/10.1515/kern-2022-0096","url":null,"abstract":"Abstract The spent nuclear fuel discharged from power reactors is a very important problem facing the future of using power reactors in electricity production. This paper focuses on the thermal-hydraulic behaviour of the VVER spent fuel in the vault dry storage system under forced convection mode, which is experimentally and numerically investigated. For this purpose, a test rig is designed and constrained to simulate the cooling loop vault system that contains four spent fuel assemblies discharged from the VVER reactor, which are represented by four electric heaters. A numerical simulation is performed by the ANSYS-CFX fluid dynamics code. The effects of decay heat generation and inlet air velocity are investigated as an operating condition. Also, the effect of the type of the Vault System tube material is being studied. The results show that the increase in the inlet air velocity improves the coolability of the fuel, while the increase in decay heat leads to a decrease in the coolability of the fuel. The used velocity range is (0.1 < V < 0.5 m/s) for inlet coolant air and heater power (20 < P < 100 W). Three tube materials (aluminum, copper, and stainless steel) were evaluated for mechanical properties, including thermal conductivity, to assess the feasibility of their use as tubes in the spent fuel storage.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"269 1","pages":"341 - 353"},"PeriodicalIF":0.5,"publicationDate":"2023-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80162379","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}
R. Lehnigk, M. Bruschewski, Tobias Huste, D. Lucas, Markus Rehm, F. Schlegel
Abstract Open-source environments such as the Computational Fluid Dynamics software OpenFOAM are very appealing for research groups since they allow for an efficient prototyping of new models or concepts. However, for downstream developments to be sustainable, i.e. reproducible and reusable in the long term, a significant amount of maintenance work must be accounted for. To allow for growth and extensibility, the maintenance work should be underpinned by a high degree of automation for repetitive tasks such as build tests, code deployment and validation runs, in order to keep the focus on scientific work. Here, an information technology environment is presented that aids the centralized maintenance of addon code and setup files with relation to reactor coolant system safety research. It fosters collaborative developments and review processes. State-of-the-art tools for managing software developments are adapted to meet the requirements of OpenFOAM. A flexible approach for upgrading the underlying installation is proposed, based on snapshots of the OpenFOAM development line rather than yearly version releases, to make new functionality available when needed by associated research projects. The process of upgrading within so-called sprint cycles is accompanied by several checks to ensure compatibility of downstream code and simulation setups. Furthermore, the foundation for building a validation data base from contributed simulation setups is laid, creating a basis for continuous quality assurance.
{"title":"Sustainable development of simulation setups and addons for OpenFOAM for nuclear reactor safety research","authors":"R. Lehnigk, M. Bruschewski, Tobias Huste, D. Lucas, Markus Rehm, F. Schlegel","doi":"10.1515/kern-2022-0107","DOIUrl":"https://doi.org/10.1515/kern-2022-0107","url":null,"abstract":"Abstract Open-source environments such as the Computational Fluid Dynamics software OpenFOAM are very appealing for research groups since they allow for an efficient prototyping of new models or concepts. However, for downstream developments to be sustainable, i.e. reproducible and reusable in the long term, a significant amount of maintenance work must be accounted for. To allow for growth and extensibility, the maintenance work should be underpinned by a high degree of automation for repetitive tasks such as build tests, code deployment and validation runs, in order to keep the focus on scientific work. Here, an information technology environment is presented that aids the centralized maintenance of addon code and setup files with relation to reactor coolant system safety research. It fosters collaborative developments and review processes. State-of-the-art tools for managing software developments are adapted to meet the requirements of OpenFOAM. A flexible approach for upgrading the underlying installation is proposed, based on snapshots of the OpenFOAM development line rather than yearly version releases, to make new functionality available when needed by associated research projects. The process of upgrading within so-called sprint cycles is accompanied by several checks to ensure compatibility of downstream code and simulation setups. Furthermore, the foundation for building a validation data base from contributed simulation setups is laid, creating a basis for continuous quality assurance.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"16 1","pages":"131 - 140"},"PeriodicalIF":0.5,"publicationDate":"2023-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80098725","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}
Abstract Critical (choked) flow is a highly concerning phenomenon in safety analysis for nuclear energy. The discharge mass flow rate prediction is crucial for engineering design and emergency response in case of nuclear accidents. Unfortunately, the critical flow is difficult to predict especially when the two-phase flow exists. The accuracy is based on a deeper understanding of the complex phenomenon of critical flow. The influence of virtual mass force on the two-phase critical flow was seldom concentrated on owing to the lack of suitable critical flow models for studies in detail. This study is based on a developed 6-equation two-phase critical flow model. It is confirmed that the virtual mass force contributes to the stability and convergence of the critical flow simulation and it will impact not only the critical mass flux but also the thermal hydraulic parameters along the discharge duct. The magnitude depends on the geometry of the discharge duct and the upstream condition. It is larger when the duct is longer and the pressure is lower. Furthermore, the virtual mass force for each flow regime was studied in detail with a sensitivity study. The results show that the most sensible condition for the virtual mass force is annular flow along a long tube under relatively low pressure. The future work is to develop a correlation of virtual mass force for critical flow specifically since the correlations in the literature were developed under general two-phase flow process conditions.
{"title":"Study of the effect of virtual mass force on two-phase critical flow","authors":"Hong Xu, Jiayue Chen, P. Ming, A. Badea, Xu Cheng","doi":"10.1515/kern-2022-0072","DOIUrl":"https://doi.org/10.1515/kern-2022-0072","url":null,"abstract":"Abstract Critical (choked) flow is a highly concerning phenomenon in safety analysis for nuclear energy. The discharge mass flow rate prediction is crucial for engineering design and emergency response in case of nuclear accidents. Unfortunately, the critical flow is difficult to predict especially when the two-phase flow exists. The accuracy is based on a deeper understanding of the complex phenomenon of critical flow. The influence of virtual mass force on the two-phase critical flow was seldom concentrated on owing to the lack of suitable critical flow models for studies in detail. This study is based on a developed 6-equation two-phase critical flow model. It is confirmed that the virtual mass force contributes to the stability and convergence of the critical flow simulation and it will impact not only the critical mass flux but also the thermal hydraulic parameters along the discharge duct. The magnitude depends on the geometry of the discharge duct and the upstream condition. It is larger when the duct is longer and the pressure is lower. Furthermore, the virtual mass force for each flow regime was studied in detail with a sensitivity study. The results show that the most sensible condition for the virtual mass force is annular flow along a long tube under relatively low pressure. The future work is to develop a correlation of virtual mass force for critical flow specifically since the correlations in the literature were developed under general two-phase flow process conditions.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"36 1","pages":"203 - 212"},"PeriodicalIF":0.5,"publicationDate":"2023-02-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85654097","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}
Abstract A CANDU reactor core comprises several hundred horizontal fuel channels spanning a calandria vessel. Loss of sufficient cooling during a severe accident could result in collapse of the core to the bottom of the calandria. A simple computational tool for simulating, in two dimensions, the resulting build-up of a terminal debris bed is described. The tool is used to model a variety of core collapse scenarios. Computed debris beds are generally lower in the middle, ∼10 fuel channels deep, and have higher decay power in their interiors. The initial debris bed porosity is estimated to be 0.65 ± 0.15. High porosity could augment in-vessel hydrogen generation and fission product release during subsequent debris bed heat-up.
{"title":"Model of terminal debris bed formation after a CANDU core collapse","authors":"R. David","doi":"10.1515/kern-2022-0095","DOIUrl":"https://doi.org/10.1515/kern-2022-0095","url":null,"abstract":"Abstract A CANDU reactor core comprises several hundred horizontal fuel channels spanning a calandria vessel. Loss of sufficient cooling during a severe accident could result in collapse of the core to the bottom of the calandria. A simple computational tool for simulating, in two dimensions, the resulting build-up of a terminal debris bed is described. The tool is used to model a variety of core collapse scenarios. Computed debris beds are generally lower in the middle, ∼10 fuel channels deep, and have higher decay power in their interiors. The initial debris bed porosity is estimated to be 0.65 ± 0.15. High porosity could augment in-vessel hydrogen generation and fission product release during subsequent debris bed heat-up.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"70 1","pages":"186 - 193"},"PeriodicalIF":0.5,"publicationDate":"2023-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88532473","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}
Abstract Remaining Useful Life (RUL) estimation has been extensively explored in recent years. RUL could be used in deciding the maintenance timeline or inspection interval for the Reaktor Serba Guna – G. A. Siwabessy (RSG-GAS reactor). RSG-GAS reactor is a pool-type research reactor (built by the Interatom Internationale of Germany) and has been operating for more than 30 years to date. This study aimed to propose a Weibull model to find the RUL estimation value of the distribution parameters of the mean time to failure (MTTF). Therefore, the RSG-GAS reactor would be higher safety, longer lifetime and higher reliability with a smaller failure rate including for the PA01-AP01 secondary pump. The research methodology is processing data collection and estimating the parameters of the Weibull model to determine maintenance timeline or inspection intervals based on the MTTF value in case the reliability has reached the targeted percentage. Results show that the RUL estimation has been obtained for the RSG-GAS reactor. In the implemented study, a maintenance timeline has been stipulated for the PA01-AP01 secondary pump (with the model of KSB and type of CPK-S350-400) for the reliability of 90% and RUL estimation of circa 29 days.
摘要剩余使用寿命(RUL)的估算是近年来研究的热点。RUL可用于确定Serba Guna - g.a. Siwabessy反应堆(RSG-GAS反应堆)的维护时间表或检查间隔。RSG-GAS反应堆是池式研究堆(由德国国际原子能机构建造)，迄今已运行30多年。本研究旨在建立威布尔模型，以求平均失效时间(MTTF)分布参数的RUL估计值。因此，包括PA01-AP01二次泵在内，RSG-GAS反应器将具有更高的安全性、更长的使用寿命和更高的可靠性，故障率更小。研究方法是在可靠性达到目标百分比的情况下，对收集的数据进行处理并估计威布尔模型的参数，根据MTTF值确定维修时间表或检查间隔。结果表明，得到了RSG-GAS反应器的RUL估计。在实施的研究中，对PA01-AP01次泵(型号为KSB，型号为CPK-S350-400)规定了维护时间表，可靠性为90%，RUL估计约为29天。
{"title":"Weibull model for RUL estimation at RSG-GAS reactor implemented on PA01-AP01 secondary pump","authors":"S. Sudadiyo","doi":"10.1515/kern-2022-0080","DOIUrl":"https://doi.org/10.1515/kern-2022-0080","url":null,"abstract":"Abstract Remaining Useful Life (RUL) estimation has been extensively explored in recent years. RUL could be used in deciding the maintenance timeline or inspection interval for the Reaktor Serba Guna – G. A. Siwabessy (RSG-GAS reactor). RSG-GAS reactor is a pool-type research reactor (built by the Interatom Internationale of Germany) and has been operating for more than 30 years to date. This study aimed to propose a Weibull model to find the RUL estimation value of the distribution parameters of the mean time to failure (MTTF). Therefore, the RSG-GAS reactor would be higher safety, longer lifetime and higher reliability with a smaller failure rate including for the PA01-AP01 secondary pump. The research methodology is processing data collection and estimating the parameters of the Weibull model to determine maintenance timeline or inspection intervals based on the MTTF value in case the reliability has reached the targeted percentage. Results show that the RUL estimation has been obtained for the RSG-GAS reactor. In the implemented study, a maintenance timeline has been stipulated for the PA01-AP01 secondary pump (with the model of KSB and type of CPK-S350-400) for the reliability of 90% and RUL estimation of circa 29 days.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"22 1","pages":"194 - 202"},"PeriodicalIF":0.5,"publicationDate":"2023-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74197618","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}
Abstract The 33rd German CFD Network of Competence Meeting was held in March 2022 at the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH in Garching, Germany. In 2022 the meeting celebrates its 20th anniversary with 17 scientific presentations, distributed in two main sessions: “Simulation of Reactor Cooling Circuit Flows” and “Simulation of Reactor Containment Flows”. This paper gives an overview of the different contributions, presented at this anniversary meeting, and also provides information on the background and the objectives of the German CFD Network of Competence.
{"title":"The 33rd German CFD Network of Competence Meeting: 20 years of advances in the numerical 3D simulation of reactor relevant flows","authors":"A. Papukchiev, B. Schramm","doi":"10.1515/kern-2022-0108","DOIUrl":"https://doi.org/10.1515/kern-2022-0108","url":null,"abstract":"Abstract The 33rd German CFD Network of Competence Meeting was held in March 2022 at the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH in Garching, Germany. In 2022 the meeting celebrates its 20th anniversary with 17 scientific presentations, distributed in two main sessions: “Simulation of Reactor Cooling Circuit Flows” and “Simulation of Reactor Containment Flows”. This paper gives an overview of the different contributions, presented at this anniversary meeting, and also provides information on the background and the objectives of the German CFD Network of Competence.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"103 1","pages":"121 - 130"},"PeriodicalIF":0.5,"publicationDate":"2023-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80838230","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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