Kerntechnik最新文献_第6页

Scaling effect on cesium diffusion in compacted MX-80 bentonite for buffer materials in HLW repository 高浓缩铀储存库缓冲材料用MX-80膨润土中铯扩散的结垢效应

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-03-14 DOI: 10.1515/kern-2022-0122

Yi-Ling Liu, Tzu-Ting Lin, C. Lee

{"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}

引用次数: 0

GRS contributions to flow-induced vibrations related activities in Europe GRS对欧洲流动诱发振动相关活动的贡献

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-03-13 DOI: 10.1515/kern-2022-0110

A. Papukchiev

引用次数: 0

Study on the accidents analyses of a single channel for XADS by using MPC-LBE code 基于MPC-LBE码的XADS单信道事故分析研究

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-03-13 DOI: 10.1515/kern-2022-0085

Ling Zhang, Tianxin Song, Zhi-Xing Gu, Jianing Dai, Wenlan Ou, Qiwen Pan, Zhengyu Gong

{"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}

引用次数: 0

Thermal hydraulic analysis of VVER spent fuels stored in vault dry system under different operating and design conditions 不同工况和设计工况下VVER拱顶干燥系统乏燃料的热水力分析

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-03-13 DOI: 10.1515/kern-2022-0096

S. Elnaggar, Samaa. A. Wasfy, S. Abdel-Latif, A. Refaey

{"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}

引用次数: 0

Sustainable development of simulation setups and addons for OpenFOAM for nuclear reactor safety research 用于核反应堆安全研究的OpenFOAM模拟设置和附加组件的可持续发展

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-02-16 DOI: 10.1515/kern-2022-0107

R. Lehnigk, M. Bruschewski, Tobias Huste, D. Lucas, Markus Rehm, F. Schlegel

{"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}

引用次数: 0

Study of the effect of virtual mass force on two-phase critical flow 虚质量力对两相临界流影响的研究

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-02-15 DOI: 10.1515/kern-2022-0072

Hong Xu, Jiayue Chen, P. Ming, A. Badea, Xu Cheng

{"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}

引用次数: 0

Model of terminal debris bed formation after a CANDU core collapse CANDU岩心崩塌后末端碎屑床形成模型

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-02-08 DOI: 10.1515/kern-2022-0095

R. David

引用次数: 0

Frontmatter 头版头条

4区工程技术

Kerntechnik Pub Date : 2023-02-01 DOI: 10.1515/kern-2023-frontmatter1

引用次数: 0

Weibull model for RUL estimation at RSG-GAS reactor implemented on PA01-AP01 secondary pump 基于PA01-AP01二次泵的RSG-GAS反应器RUL估计Weibull模型

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-01-06 DOI: 10.1515/kern-2022-0080

S. Sudadiyo

{"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}

引用次数: 0

The 33rd German CFD Network of Competence Meeting: 20 years of advances in the numerical 3D simulation of reactor relevant flows 第33届德国CFD能力网络会议:反应堆相关流动数值三维模拟的20年进展

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2023-01-06 DOI: 10.1515/kern-2022-0108

A. Papukchiev, B. Schramm

引用次数: 0