In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.
{"title":"Experimental Research for CHF Sensitivity of Heat Flux Distribution under IVR Conditions","authors":"Shilei Han, Pengfei Liu, B. Kuang, Yanhua Yang","doi":"10.1155/2022/3522470","DOIUrl":"https://doi.org/10.1155/2022/3522470","url":null,"abstract":"In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"12 4","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41295859","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}
Dae-Hwa Hong, D. Cho, Jinwoo Kim, A. Diab, Cigdem Cildag
The presence of a stable stratified gas cloud inside the containment near or at the flammability limit may lead to deflagration or even detonation which may challenge the containment and cause a radioactive material release into the environment. To mitigate this risk, a number of approaches have been proposed, for example, containment inerting or venting and use of passive autocatalytic recombiners or igniters. However, for these measures to be effective, a thorough analysis of the hydrogen dispersion and associated phenomena is indispensable during the design phase as well as the mitigation phase during a severe accident. In this work, a MAAP analysis is performed to assess the hydrogen risk in a typical pressurized water reactor (PWR) containment. An extended station blackout (SBO) was chosen as an initiating event given its high contribution to the core damage frequency. RCS depressurization and external injection are mitigation techniques implemented consecutively to extend the coping capability of the plant for the extended SBO scenario. A sensitivity study is performed to select the combination of timing and flow rate that generate the most severe case for the “in-vessel phase of hydrogen generation.” Subsequently, a number of passive autocatalytic recombiners (PARs) were implemented to mitigate the hydrogen risk during the first three days of the accident. The Shapiro diagram is used to assess the flammability condition of the containment atmosphere based on MAAP analysis. The results show that the gas mixture composition is acceptable in the majority of the containment compartments and only marginally acceptable in the cavity. Even under the conservative conditions of the accident, the simulation results confirmed the sufficiency of recombiners alone without igniters in the low hydrogen concentration zones, while for compartments close to the sources, additional mitigation may be needed.
{"title":"Verification of the Efficacy of Passive Autocatalytic Recombiners in a Typical Pressurized Water Reactor under a Station Blackout Condition","authors":"Dae-Hwa Hong, D. Cho, Jinwoo Kim, A. Diab, Cigdem Cildag","doi":"10.1155/2022/7129092","DOIUrl":"https://doi.org/10.1155/2022/7129092","url":null,"abstract":"The presence of a stable stratified gas cloud inside the containment near or at the flammability limit may lead to deflagration or even detonation which may challenge the containment and cause a radioactive material release into the environment. To mitigate this risk, a number of approaches have been proposed, for example, containment inerting or venting and use of passive autocatalytic recombiners or igniters. However, for these measures to be effective, a thorough analysis of the hydrogen dispersion and associated phenomena is indispensable during the design phase as well as the mitigation phase during a severe accident. In this work, a MAAP analysis is performed to assess the hydrogen risk in a typical pressurized water reactor (PWR) containment. An extended station blackout (SBO) was chosen as an initiating event given its high contribution to the core damage frequency. RCS depressurization and external injection are mitigation techniques implemented consecutively to extend the coping capability of the plant for the extended SBO scenario. A sensitivity study is performed to select the combination of timing and flow rate that generate the most severe case for the “in-vessel phase of hydrogen generation.” Subsequently, a number of passive autocatalytic recombiners (PARs) were implemented to mitigate the hydrogen risk during the first three days of the accident. The Shapiro diagram is used to assess the flammability condition of the containment atmosphere based on MAAP analysis. The results show that the gas mixture composition is acceptable in the majority of the containment compartments and only marginally acceptable in the cavity. Even under the conservative conditions of the accident, the simulation results confirmed the sufficiency of recombiners alone without igniters in the low hydrogen concentration zones, while for compartments close to the sources, additional mitigation may be needed.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42403020","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In view of the multimetal barrier environment of nuclear power plant, by considering the factors such as transmission power, transmission position, and multipath interference, based on the simulation of metal pipes and equipment, this paper carries out the barrier area channel modeling in logarithmic fading mode and makes quantitative analysis on the channel transmission, path loss, channel power characteristics, and so on under the metal barrier environment. Based on the channel modeling, this paper optimizes the coverage of the network in the obstacle area by using the improved teaching and learning group intelligent algorithm. The simulation results show that the improved teaching and learning algorithm can optimize the network coverage of the obstacle area well, and under the four obstacle modules, 14 nodes can cover the whole area by more than 99%. This provides a solution to the problem of network coverage in the practical application of wireless sensor networks.
{"title":"Research on Channel Modeling and Communication Coverage of Wireless Sensor Networks in Barrier Area of Nuclear Power Plants","authors":"Zhiguang Deng, Qian Wu, Xin Lv, Bi-Wei Zhu, M. Xiang, Xue-Mei Wang, Jia-Liang Zhu","doi":"10.1155/2022/5223755","DOIUrl":"https://doi.org/10.1155/2022/5223755","url":null,"abstract":"In view of the multimetal barrier environment of nuclear power plant, by considering the factors such as transmission power, transmission position, and multipath interference, based on the simulation of metal pipes and equipment, this paper carries out the barrier area channel modeling in logarithmic fading mode and makes quantitative analysis on the channel transmission, path loss, channel power characteristics, and so on under the metal barrier environment. Based on the channel modeling, this paper optimizes the coverage of the network in the obstacle area by using the improved teaching and learning group intelligent algorithm. The simulation results show that the improved teaching and learning algorithm can optimize the network coverage of the obstacle area well, and under the four obstacle modules, 14 nodes can cover the whole area by more than 99%. This provides a solution to the problem of network coverage in the practical application of wireless sensor networks.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47236389","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}
Yubiao Zhang, H. Xue, Shun Zhang, Shuai Wang, Yuman Sun, Yonggang Zhang, Yongjie Yang
The interaction between the mechanical heterogeneity and the residual stress in dissimilar metal welded joints (DMWJs) leads to a complex mechanical field of crack tips, which strongly affects stress corrosion cracking (SCC) behaviors. A dual-field coupling model was established by using the user-defined field (USDFLD) and the predefined stress field method based on the elastoplastic finite element method in this study. Thus, the mechanical heterogeneity and the residual stress of the DMWJ are realized. The influence of the interaction between the mechanical heterogeneity and the residual stress on the mechanical field of crack tips at different locations was investigated. The results show that the mechanical heterogeneity causes the stress and strain distribution on both sides of the crack tip asymmetry. And the residual stress affects the magnitude of the stress and strain around the crack tip. The variation trend of the stress and strain along the crack propagation with crack length is basically the same as that of the residual stress. However, the stress and strain distributions are slightly lagging behind the residual stress distribution due to the redistribution of the residual stress caused by the crack propagation. In addition, the stress and strain range of cracks at different positions with crack length are also different.
{"title":"Interaction of Mechanical Heterogeneity and Residual Stress on Mechanical Field at Crack Tips in DMWJs","authors":"Yubiao Zhang, H. Xue, Shun Zhang, Shuai Wang, Yuman Sun, Yonggang Zhang, Yongjie Yang","doi":"10.1155/2022/7462200","DOIUrl":"https://doi.org/10.1155/2022/7462200","url":null,"abstract":"The interaction between the mechanical heterogeneity and the residual stress in dissimilar metal welded joints (DMWJs) leads to a complex mechanical field of crack tips, which strongly affects stress corrosion cracking (SCC) behaviors. A dual-field coupling model was established by using the user-defined field (USDFLD) and the predefined stress field method based on the elastoplastic finite element method in this study. Thus, the mechanical heterogeneity and the residual stress of the DMWJ are realized. The influence of the interaction between the mechanical heterogeneity and the residual stress on the mechanical field of crack tips at different locations was investigated. The results show that the mechanical heterogeneity causes the stress and strain distribution on both sides of the crack tip asymmetry. And the residual stress affects the magnitude of the stress and strain around the crack tip. The variation trend of the stress and strain along the crack propagation with crack length is basically the same as that of the residual stress. However, the stress and strain distributions are slightly lagging behind the residual stress distribution due to the redistribution of the residual stress caused by the crack propagation. In addition, the stress and strain range of cracks at different positions with crack length are also different.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42490934","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}
Radioactivity of gross alpha/beta is an index of water quality detection, which can reflect the radioactivity intensity of water. However, the traditional detection method of these parameters, thick source method, has problems of cumbersome and time consumption in sample preparation and cannot realize the rapid detection on-site. Based on this, this paper studies the enrichment method based on reverse osmosis membrane to accurately and quickly determine the gross α and gross β in water by using the reverse osmosis membrane as the carrier and enriching the radionuclides in water to the high-pressure side of the reverse osmosis membrane to replace the sample preparation process in traditional thick source method, so as to shorten the sample processing time in the detection process and avoid the cumbersome sample preparation process. The reverse osmosis membrane enrichment method for the determination of gross in 241Am and 40KCl standard solutions was used to study gross alpha/beta radioactivity, and the results showed that the average recoveries of radioactivity of gross alpha/beta were 95.0% and 93.6%, respectively. At the same time, the results of the thick source method and the reverse osmosis membrane method on the gross alpha/beta of actual water samples in 5 different regions were compared. It showed that the thick source method and the reverse osmosis membrane method had a good consistency in the detection results of total α and total β radioactivity, and the reverse osmosis membrane method had better stability than the thick source method. The average relative standard deviations (RSD) of the gross alpha and gross beta activity obtained by the thick source method are 11.9% and 7.3%, respectively, while RSD of the gross alpha and gross beta radioactivity obtained by the reverse osmosis membrane method were 6.9% and 4.7%, respectively. The preparation time of single sample was reduced by 75.7%, and the overall detection cycle time was reduced by 68.1%.
{"title":"Rapid Determination of Gross Alpha/Beta Activity in Water Based on Reverse Osmosis Membrane Enrichment Pretreatment","authors":"Ruiqi Zhang, Jianye Wang, Nanjing Zhao, Xuyun Huang","doi":"10.1155/2022/2868792","DOIUrl":"https://doi.org/10.1155/2022/2868792","url":null,"abstract":"Radioactivity of gross alpha/beta is an index of water quality detection, which can reflect the radioactivity intensity of water. However, the traditional detection method of these parameters, thick source method, has problems of cumbersome and time consumption in sample preparation and cannot realize the rapid detection on-site. Based on this, this paper studies the enrichment method based on reverse osmosis membrane to accurately and quickly determine the gross <i>α</i> and gross <i>β</i> in water by using the reverse osmosis membrane as the carrier and enriching the radionuclides in water to the high-pressure side of the reverse osmosis membrane to replace the sample preparation process in traditional thick source method, so as to shorten the sample processing time in the detection process and avoid the cumbersome sample preparation process. The reverse osmosis membrane enrichment method for the determination of gross in <sup>241</sup>Am and <sup>40</sup>KCl standard solutions was used to study gross alpha/beta radioactivity, and the results showed that the average recoveries of radioactivity of gross alpha/beta were 95.0% and 93.6%, respectively. At the same time, the results of the thick source method and the reverse osmosis membrane method on the gross alpha/beta of actual water samples in 5 different regions were compared. It showed that the thick source method and the reverse osmosis membrane method had a good consistency in the detection results of total <i>α</i> and total <i>β</i> radioactivity, and the reverse osmosis membrane method had better stability than the thick source method. The average relative standard deviations (RSD) of the gross alpha and gross beta activity obtained by the thick source method are 11.9% and 7.3%, respectively, while RSD of the gross alpha and gross beta radioactivity obtained by the reverse osmosis membrane method were 6.9% and 4.7%, respectively. The preparation time of single sample was reduced by 75.7%, and the overall detection cycle time was reduced by 68.1%.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"179 2","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-02-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138513441","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}
By analyzing the recorded operation data of a nuclear power plant (NPP), its results can serve the fault detection or operation experience feedback. Data missing exists in the recorded operation data. It may lower the data quality and affect the accuracy of the analysis results. In order to improve the data quality, two parts of researches are carried on. Firstly, to locate the missing data accurately the detecting algorithm for missing data of the NPP operation parameters based on wavelet analysis. Different judging basis is proposed for discrete and continuous missing respectively. Then, the filling method based on the hot deck algorithm are studied. As the dynamic properties of the parameters are closely related to the operating state of NPP, the similarity of the operation parameter vectors are formed to express the similarity of the operating states, so as to fulfill the requirements of the hot deck algorithm. To improve the accuracy of the measuring results, taken the differences between the characteristics of the analog parameters and the switch parameters into consideration, the similarity measurements using Mahalanobis distance for the analog parameter vectors and the matching measure for the switch parameter vectors are studied respectively. Finally, the operation data is taken to build the experiment data set for the algorithm verification. The results shows that the designed algorithm performs much better than the mean interpolation method and LSTM.
{"title":"Study on Missing Data Filling Algorithm of Nuclear Power Plant Operation Parameters","authors":"Tianshu Wang, Ren Yu, Qiao Peng","doi":"10.1155/2022/4172622","DOIUrl":"https://doi.org/10.1155/2022/4172622","url":null,"abstract":"By analyzing the recorded operation data of a nuclear power plant (NPP), its results can serve the fault detection or operation experience feedback. Data missing exists in the recorded operation data. It may lower the data quality and affect the accuracy of the analysis results. In order to improve the data quality, two parts of researches are carried on. Firstly, to locate the missing data accurately the detecting algorithm for missing data of the NPP operation parameters based on wavelet analysis. Different judging basis is proposed for discrete and continuous missing respectively. Then, the filling method based on the hot deck algorithm are studied. As the dynamic properties of the parameters are closely related to the operating state of NPP, the similarity of the operation parameter vectors are formed to express the similarity of the operating states, so as to fulfill the requirements of the hot deck algorithm. To improve the accuracy of the measuring results, taken the differences between the characteristics of the analog parameters and the switch parameters into consideration, the similarity measurements using Mahalanobis distance for the analog parameter vectors and the matching measure for the switch parameter vectors are studied respectively. Finally, the operation data is taken to build the experiment data set for the algorithm verification. The results shows that the designed algorithm performs much better than the mean interpolation method and LSTM.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-02-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47647916","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTS-induced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a two-loop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. The material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. The results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. This study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.
{"title":"Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System","authors":"M. Annor-Nyarko, Hong Xia, A. Ayodeji","doi":"10.1155/2022/5886583","DOIUrl":"https://doi.org/10.1155/2022/5886583","url":null,"abstract":"The damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTS-induced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a two-loop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. The material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. The results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. This study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46884654","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Pavliuk, Evgeniy V. Bespala, S. Kotlyarevskiy, I. Novoselov, V. N. Kotov
The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors. The analysis of heat release processes inside storages containing irradiated nuclear graphite, representing a potential hazard due to the possible heating and, accordingly, the release of long-lived radionuclides during oxidation was carried out. The following factors were considered as the main factors that can lead to an increase in the temperature inside the storage facility: corrosion of metallic radioactive waste, the presence of fuel fragments, and also the random exposure of irradiated graphite to local sources of thermal energy (spark, etc.). It was noted in the work that the combined or separate influence of some factors can lead to an increase in the temperature of the onset of the initiation of Wigner energy release in graphite radwaste (Tin ≈ 90–100°C for the “Worst-case” graphite). The model of heat generation in the storage was developed based on the analysis of the features of graphite radioactive waste storage and Wigner energy release. The layered location of different types of waste (graphite and aluminum) and the local character of the distribution of heat sources were adopted in this model. The greatest heating is achieved if graphite radioactive waste is located near the concrete walls of the storage facility, as well as in direct contact with irradiated aluminum radioactive waste, which was shown in this paper.
{"title":"Analysis of Heat Release Processes inside Storage Facilities Containing Irradiated Nuclear Graphite","authors":"A. Pavliuk, Evgeniy V. Bespala, S. Kotlyarevskiy, I. Novoselov, V. N. Kotov","doi":"10.1155/2022/2957310","DOIUrl":"https://doi.org/10.1155/2022/2957310","url":null,"abstract":"The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors. The analysis of heat release processes inside storages containing irradiated nuclear graphite, representing a potential hazard due to the possible heating and, accordingly, the release of long-lived radionuclides during oxidation was carried out. The following factors were considered as the main factors that can lead to an increase in the temperature inside the storage facility: corrosion of metallic radioactive waste, the presence of fuel fragments, and also the random exposure of irradiated graphite to local sources of thermal energy (spark, etc.). It was noted in the work that the combined or separate influence of some factors can lead to an increase in the temperature of the onset of the initiation of Wigner energy release in graphite radwaste (Tin ≈ 90–100°C for the “Worst-case” graphite). The model of heat generation in the storage was developed based on the analysis of the features of graphite radioactive waste storage and Wigner energy release. The layered location of different types of waste (graphite and aluminum) and the local character of the distribution of heat sources were adopted in this model. The greatest heating is achieved if graphite radioactive waste is located near the concrete walls of the storage facility, as well as in direct contact with irradiated aluminum radioactive waste, which was shown in this paper.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46561240","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The transportation of heavy equipment in nuclear engineering has always been the focus of engineers, especially those transfer devices with the characteristics of small geometric size and heavy load. According to this kind of compact heavy-load transfer device and its engineering tasks, the core problems caused by excessive vertical and horizontal forces in the design process were analyzed. By introducing the theory of inventive problem solving (TRIZ) design method, these problems were creatively solved by the contradiction theory and substance-field model in TRIZ, and an innovative design scheme of the compact heavy load-independent transfer device was obtained. Through the analysis of the design scheme and the stability and rapidity of its hydraulic system, some key parameters were determined. The power of the transfer device was all from the hydraulic system, and it can carry up to 300 t weight of reactor equipment, while its geometric size was only 1600 × 400 × 500 mm. It was of great significance to improve the efficiency of the nuclear engineering system.
{"title":"Innovative Design of Compact Heavy-Load Independent Transfer Device for Nuclear Engineering System","authors":"Hao Wan, Songfeng Weng, Hua Du, Dailin Dong, Bingyan Wang, Tianda Yu","doi":"10.1155/2022/5256808","DOIUrl":"https://doi.org/10.1155/2022/5256808","url":null,"abstract":"The transportation of heavy equipment in nuclear engineering has always been the focus of engineers, especially those transfer devices with the characteristics of small geometric size and heavy load. According to this kind of compact heavy-load transfer device and its engineering tasks, the core problems caused by excessive vertical and horizontal forces in the design process were analyzed. By introducing the theory of inventive problem solving (TRIZ) design method, these problems were creatively solved by the contradiction theory and substance-field model in TRIZ, and an innovative design scheme of the compact heavy load-independent transfer device was obtained. Through the analysis of the design scheme and the stability and rapidity of its hydraulic system, some key parameters were determined. The power of the transfer device was all from the hydraulic system, and it can carry up to 300 t weight of reactor equipment, while its geometric size was only 1600 × 400 × 500 mm. It was of great significance to improve the efficiency of the nuclear engineering system.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47122539","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm ∗ 60 mm and 2.8 mm ∗ 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that Δ T ONB (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, Δ T ONB in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.
在质量通量为100 ~ 300 kg/(m2·s)，进水温度为70 ~ 100℃，热流密度为10 ~ 70 kW/m2，局部压力为0.145 MPa的条件下，对两个垂直矩形通道内的ONB(核沸腾起始)进行了实验研究。截面尺寸分别为1.8 mm∗60 mm和2.8 mm∗60 mm。利用三种沸腾起始判断方法对ONB点进行了定位，发现Δ T ONB (ONB点壁面过热度)随入口温度的降低而增大，随质量通量的增大而增大。结果还表明，尽管在ONB位置，窄通道内的气泡大小和行为与非窄通道内的气泡大小和行为不同，但传热没有受到明显影响。此外，通过Thom提出的相关性可以在±30%的误差范围内预测两个通道中的Δ T ONB。
{"title":"Experimental Study of Onset of Nucleate Boiling in Vertical Rectangular Channels with Different Flow Path Heights","authors":"N. Cheng, Shuwen Yu, J. Xiao, C. Peng","doi":"10.1155/2022/7760569","DOIUrl":"https://doi.org/10.1155/2022/7760569","url":null,"abstract":"A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm \u0000 \u0000 ∗\u0000 \u0000 60 mm and 2.8 mm \u0000 \u0000 ∗\u0000 \u0000 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that \u0000 \u0000 Δ\u0000 \u0000 \u0000 T\u0000 \u0000 \u0000 ONB\u0000 \u0000 \u0000 \u0000 (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, \u0000 \u0000 Δ\u0000 \u0000 \u0000 T\u0000 \u0000 \u0000 ONB\u0000 \u0000 \u0000 \u0000 in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41486924","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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