International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第3页

Drop analysis and numerical simulation of spent fuel transfer basket 乏燃料输送篮跌落分析及数值模拟

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.012

Jiaojiao Dong, Chen Wang, Xiaogang Xu, Siqi Zhu

{"title":"Drop analysis and numerical simulation of spent fuel transfer basket","authors":"Jiaojiao Dong, Chen Wang, Xiaogang Xu, Siqi Zhu","doi":"10.1016/j.jandt.2025.04.012","DOIUrl":"10.1016/j.jandt.2025.04.012","url":null,"abstract":"<div><div>The spent fuel transfer basket is responsible for the transfer and storage of the assemblies, and the fall condition of the basket needs to be considered to ensure the critical safety of the spent fuel assemblies during the operation process. According to the actual process, two working conditions are assumed, i.e. vertical fall of the basket at a height of 1 m from the bottom of the pool in the air condition and tipping over underwater. The key parameters in the numerical simulation are verified by carrying out experiments to obtain numerical simulation results with high accuracy. Aiming at the tipping process which is greatly influenced by the fluid resistance, the underwater tipping test of the 1:2 model of the basket is carried out to obtain the influence of the fluid on the tipping of the basket and to obtain the equivalent fall angle of the tipping fall in the air of the basket. The equivalent stiffness of the assemblies in the model is corrected by the 1:1 model 1-m vertical drop test, and the corrected numerical simulation agrees well with the experimental results, which indicates that the drop calculation model obtained in this study can be applied to the actual design.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 55-62"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143950464","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}

引用次数: 0

Numerical study on PWR fuel rod cladding ballooning and burst behavior with the thermo-mechanical coupling finite element method 采用热力耦合有限元法对压水堆燃料棒包壳气胀爆行为进行数值研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.003

Xin Zhang , Gen Li , Runhua Gao , Xinhai Zhao , Peitao Yao

{"title":"Numerical study on PWR fuel rod cladding ballooning and burst behavior with the thermo-mechanical coupling finite element method","authors":"Xin Zhang , Gen Li , Runhua Gao , Xinhai Zhao , Peitao Yao","doi":"10.1016/j.jandt.2025.04.003","DOIUrl":"10.1016/j.jandt.2025.04.003","url":null,"abstract":"<div><div>The ballooning behavior of fuel rod cladding is an important issue of nuclear reactor, which may initiate a severe nuclear reactor accident. Based on COMSOL multi-physical module, this research established a thermo-mechanical coupling finite element model, which can accurately simulate the distribution and evolution of temperature, strain, and stress of the fuel cladding under high temperature and pressure. The validation results indicated that the maximum deviations of the predicted temperature and strain are 8 % and 11 %, respectively, in comparison to the experimental data. Moreover, the burst data predicted by the model were consistent with the results of Yadav experiment. These tests encompassed a heating rate of 2 K/s to 8 K/s and an internal overpressure range of 1 MPa–9 MPa. The burst temperature predicted by the model decreased with the increase of internal overpressure, from 1315 K to approximately 1014 K. This trend was consistent with that observed in the experimental data, thereby verifying the model's accuracy and reliability. The effects of key parameters such as cladding heating rate and cladding internal pressure on cladding ballooning and burst behavior were further analyzed. The results indicated that as the heating power increased, the cladding burst temperature rose from 1055 K to 1290 K, while the burst strain decreased from 68 % to 27 %. When the internal overpressure rose from 3 MPa to 7 MPa, the burst strain and stress increased by 93 % and 174 %, respectively. This study elucidated thermo-mechanical cladding response, facilitating predictive safety analysis and optimal fuel rod design to mitigate loss-of-coolant accident risks in nuclear reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 7-18"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143859354","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}

引用次数: 0

Preliminary study on the postulated siting accident source term of integrated small reactor 关于综合小型反应堆选址事故源假设条件的初步研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.006

Dawei Sun, Shengqin Gao, Liang Zhai, Qiliang Mei, Yaru Fu, Jie Mao

{"title":"Preliminary study on the postulated siting accident source term of integrated small reactor","authors":"Dawei Sun, Shengqin Gao, Liang Zhai, Qiliang Mei, Yaru Fu, Jie Mao","doi":"10.1016/j.jandt.2025.04.006","DOIUrl":"10.1016/j.jandt.2025.04.006","url":null,"abstract":"<div><div>In accordance with the characteristics of integrated small reactor, a study was conducted on the postulated siting accident source term scheme. The principles for selecting accident types were explored for both the design basis accident condition and the design extension condition. For the deterministic design basis accident condition, a representative accident, such as a fuel handling accident, was identified from the perspective of envelope; for the design extension condition, a representative accident was identified based on the likelihood of accident occurrence, such as when the screening frequency is 10<sup>−7</sup>/(reactor·year). When the probability of core damage is extremely low, it is considered to construct at least one entire core fuel cladding damage accident as a representative case. Focusing on the radioactive production, migration, removal, and release, a study was conducted on the source term analysis models and key parameter values of various representative accidents, thus simulating the radioactive activity released into the environment by the accident and identifying <sup>131</sup>I as the dominant nuclide that should be given special attention. Taking a coastal plant site as the object, the dose to the public staying at 16 locations along the site boundary after the accident was further evaluated, and it was demonstrated that even reducing the site boundary to 100 m around the reactor can meet the dose limit requirements of the “small reactor review principle”, and the suitability of the source term schemes was certificated.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 1-6"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143859353","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}

引用次数: 0

Thermomechanical and radiological investigations of storage casks for radioactive spent fuel assemblies 放射性乏燃料组件贮存桶的热机械和放射学研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.005

M. Mehdizadeh, M. Aghaie, M. Sharifi

{"title":"Thermomechanical and radiological investigations of storage casks for radioactive spent fuel assemblies","authors":"M. Mehdizadeh, M. Aghaie, M. Sharifi","doi":"10.1016/j.jandt.2025.04.005","DOIUrl":"10.1016/j.jandt.2025.04.005","url":null,"abstract":"<div><div>In this research, the Monte Carlo particle transport method and COMSOL Multiphysics were utilized to simulate and analyze the TK-13 cask, which contains spent fuel assemblies of the WWER-1000 reactor. Initially, the cask was modeled as being filled with water, and its criticality was assessed by evaluating the effective multiplication factor (Keff) while considering various neutron shielding materials, including polyethylene, borated polyethylene, ethylene glycol, and water solutions. The Keff value was calculated by varying the boron percentage in the holding basket of spent fuel assemblies and adjusting the pitch between these assemblies. Additionally, the total neutron and gamma dose rates on the cask's outer wall were measured. A thick steel wall, 34 cm in thickness, was modeled around the spent fuel assemblies, and a concrete wall was also considered, with total dose rates calculated for both cases. The uncertainty in these measurements was also quantified. The cask was further simulated in COMSOL Multiphysics to evaluate thermal and mechanical parameters. Thermal analysis was conducted in both steady and unsteady states; for steady-state analysis, the minimum and maximum fuel burnup scenarios were examined, along with the effects of fins on the outer wall. In the unsteady thermal analysis, temperature variations at different points within the cask and their distribution over a hundred-year period were investigated. The mechanical analysis focused on Von Mises stress and displacement values at various locations within the cask. Finally, the results were compared with findings from other studies, leading to comprehensive conclusions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 29-42"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143877037","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}

引用次数: 0

The impact of natural removal on the release of nuclides into the containment after accident 事故发生后，自然移除对核素释放到安全壳的影响

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.002

Haiying Chen, Xinli Gao, Pingfei Du, Yuan Gao, Fudong Liu

{"title":"The impact of natural removal on the release of nuclides into the containment after accident","authors":"Haiying Chen, Xinli Gao, Pingfei Du, Yuan Gao, Fudong Liu","doi":"10.1016/j.jandt.2025.04.002","DOIUrl":"10.1016/j.jandt.2025.04.002","url":null,"abstract":"<div><div>The natural removal coefficient directly affects the amounts of radioactive iodine and aerosols released into the containment after the accident. The radioactivity calculation models in LOCA were established, and the impact of natural removal coefficients on the release of elemental iodine and aerosols into containment was quantitatively analyzed. The results showed that due to the comprehensive influence of factors such as core release type, release time, natural removal effect, nuclide decay, and containment leakage, the radioactivity of each nuclide in the containment reached its maximum value after 40 min of the accident, and then gradually decreased over time. During the effective natural removal time, there was a significant difference in the radioactivity of elemental iodine in the containment under different natural removal coefficients. Taking <sup>131</sup>I as an example, the radioactivity ratio of elemental <sup>131</sup>I in containment corresponding to the two natural removal coefficients decreased first and then increased over time. Finally, the radioactivity of elemental <sup>131</sup>I under different natural removal coefficients was basically the same. The change of aerosol radioactivity in the containment was obviously affected by the value of natural removal coefficients. Under two different natural removal coefficients, the maximum radioactivity ratio of the aerosol nuclides in the containment was about 2.3.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 43-47"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143891342","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}

引用次数: 0

Debris filtering efficiency assessment of the fuel assembly bottom nozzle 燃油组件底部喷嘴碎屑过滤效率评估

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.004

Zheng Mei-yin, Jiao Yong-jun, Li Zheng-yang, Ren Quan-yao, Xiao Zhong, Su Min

{"title":"Debris filtering efficiency assessment of the fuel assembly bottom nozzle","authors":"Zheng Mei-yin, Jiao Yong-jun, Li Zheng-yang, Ren Quan-yao, Xiao Zhong, Su Min","doi":"10.1016/j.jandt.2025.04.004","DOIUrl":"10.1016/j.jandt.2025.04.004","url":null,"abstract":"<div><div>The debris-induced fretting wear is a key factor to cause the fuel failure in pressurized water reactor (PWR). Assessing the debris filtering efficiency of bottom nozzle is an important way to analyze the reliability of fuel assembly. The debris filtering efficiency assessment method of bottom nozzle was proposed, which concludes experimental research method and numerical calculation method. The experimental method concludes A-types debris (ball, hex nut, spring, gasket, and pin) and B-types debris (cylindrical metal wire, helical metal cutting, and metal strip). The experimental method demonstrates that the bottom nozzle with hemispherical surface structure presents the filtering efficiency of 100 % for B-types debris. The numerical simulation method is a coupling process of CFD and DEM, which mainly analyze the movement and collision of debris. The bottom nozzle shows better filtering efficiency for cylindrical debris, and the filtering efficiency in the numerical simulation is in good accordance with the experimental results. The filtering efficiency assessment method is suitable for evaluating the debris filtering behavior of bottom nozzle and helpful for the design and verification of bottom nozzle.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 48-54"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143899108","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}

引用次数: 0

Analysis of natural circulation characteristics for pool-type fast reactor system 池式快堆系统自然循环特性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 DOI: 10.1016/j.jandt.2025.04.007

Yufeng Lv, Zhiwei Zhou, Xingmin Liu

{"title":"Analysis of natural circulation characteristics for pool-type fast reactor system","authors":"Yufeng Lv, Zhiwei Zhou, Xingmin Liu","doi":"10.1016/j.jandt.2025.04.007","DOIUrl":"10.1016/j.jandt.2025.04.007","url":null,"abstract":"<div><div>In the inherent safety design of the pool-type fast reactor, when the normal heat transfer routine of the primary system is lost during accident conditions, the residual heat of the reactor core is transferred to the ultimate heat sink through the natural circulation in the reactor vessel. The natural circulation phenomenon in the pool-type reactor system under accident transient is complicated. For the purpose to improve the knowledge of natural circulation characteristics of the pool-type reactor system and validate the applicability and accuracy of system analysis code for simulating natural circulation in pool systems, the RELAP5 MOD3.2 code was adopted to calculate the station blackout (SBO) test conducted on a pool-type test facility. The analysis results show that in the early stage of the transient process, the flow re-distribution phenomenon occurs in the pool-type vessel. There exists natural circulation flow path both inside core assembly simulators and outside core assembly simulators. Due to the influence of the hot fluid at the outlet of the core and the cold fluid at the outlet of the DHX, the thermal stratification phenomenon occurs in hot pool. The main thermal-hydraulic parameters calculated by the code (such as the fluid outlet temperature in the core assembly simulator, the flow rate of the primary circuit and the fluid inlet and outlet temperature of the DHX primary side) are in good agreement with the experimental results. Through the conduct of this work, the feasibility of the RELAP5 code calculation model and modeling method is verified, which can be further adopted to the natural circulation simualtion of pool-type reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 19-28"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143865130","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}

引用次数: 0

Study on post-accident aerosol retention behavior in microchannels of steel containment in passive nuclear power plants 无源核电站钢壳微通道事故后气溶胶滞留行为研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-06-01 DOI: 10.1016/j.jandt.2024.12.006

Yaru Fu, Zeyu Chen, Qiliang Mei, Dawei Sun, Shengqin Gao, Meng Li

{"title":"Study on post-accident aerosol retention behavior in microchannels of steel containment in passive nuclear power plants","authors":"Yaru Fu, Zeyu Chen, Qiliang Mei, Dawei Sun, Shengqin Gao, Meng Li","doi":"10.1016/j.jandt.2024.12.006","DOIUrl":"10.1016/j.jandt.2024.12.006","url":null,"abstract":"<div><div>During an accident at a nuclear power plant, even if the containment is not severely damaged or fails, radioactive aerosols can still leak into the environment through microchannels that potentially exist in the containment. Unlike ordinary gases, radioactive aerosols deposit and get retained in these microchannels through various mechanisms, thereby reducing their leak into the environment. Passive nuclear power plants rely on passive removal mechanisms to eliminate radioactive aerosols, which are highly reliable and effective. The retention of aerosols in microchannels of the containment is also one of the essential passive removal mechanisms. In this work, an analysis model for aerosol retention behavior in microchannels of passive nuclear power plant containment establishes, an analysis software is independently developed, and computational analysis is conducted. The analysis results indicate that when the flow is laminar, the primary removal mechanism of aerosols is inertial impaction at the entrance. When the fluid is turbulent, the removal effect of aerosols is superior to that in laminar flow. In passive nuclear power plants, due to the external spray, most areas are covered by a water film and therefore constituted as wet regions during post-accident stage. When microchannels are in these wet regions, compared to dry regions, aerosol removal effectiveness is enhanced because of the extra removal mechanisms thermophoresis and coagulation.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 158-163"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100591","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Investigation of sodium fire aerosol diffusiophoresis deposition in concrete gaps during the processes within nuclear power plants 核电厂内混凝土间隙中钠火气溶胶扩散电泳沉积的研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-06-01 DOI: 10.1016/j.jandt.2024.12.005

Yichen Li, Tiancai Liu, Huanjun Zhu, Xiangyu Yan

{"title":"Investigation of sodium fire aerosol diffusiophoresis deposition in concrete gaps during the processes within nuclear power plants","authors":"Yichen Li, Tiancai Liu, Huanjun Zhu, Xiangyu Yan","doi":"10.1016/j.jandt.2024.12.005","DOIUrl":"10.1016/j.jandt.2024.12.005","url":null,"abstract":"<div><div>Sodium fire accidents are among the high-risk incidents that nuclear power plants with sodium-cooled fast reactors pay close attention to. Sodium fire reactions can generate a considerable amount of heat and aerosols within a short period. To analyze the influence, brought about by the aerosol migration after the sodium fire accident in the process of nuclear power plants, considering the migration process of aerosol particles within the concrete cracks of the walls in it and the concentration changes caused by the steam condensation due to the low-temperature walls. This research establishes a diffusiophoresis model of sub-micron particles under the condition of mixed gas condensation. Simulations were carried out for particles of different sizes under different steam content conditions. The results demonstrate that the diffusiophoresis effect caused by gas condensation within the concrete cracks in the process areas of nuclear power plants has played a positive role in reducing sodium fire aerosol leakage. The penetration coefficient of aerosols within the concrete cracks can be reduced by 6 %. On the one hand, it is discovered that the particle size and steam content have a significant impact on the diffusiophoresis. On average, for every 5 % increase in the steam content in the environment, the penetration coefficient of particles decreases by approximately 2 %. On the other hand, it is found that the aerosol particles generated in the initial stage of the sodium fire reaction and the end of the combustion are more susceptible to the diffusiophoresis effect.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 171-176"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100593","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Machine learning for forecasting factory concentrations of nitrogen oxides from univariate data exploiting trend attributes 利用趋势属性从单变量数据预测工厂氮氧化物浓度的机器学习

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2024-06-01 DOI: 10.1016/j.jandt.2024.12.002

Jiaxin Liu, Shuo Yang, Qichao Li, Leiming Ji, Xuefeng Hou, Liudong Hou, Jing Ma

{"title":"Machine learning for forecasting factory concentrations of nitrogen oxides from univariate data exploiting trend attributes","authors":"Jiaxin Liu, Shuo Yang, Qichao Li, Leiming Ji, Xuefeng Hou, Liudong Hou, Jing Ma","doi":"10.1016/j.jandt.2024.12.002","DOIUrl":"10.1016/j.jandt.2024.12.002","url":null,"abstract":"<div><div>The development of post-processing technology for spent nuclear fuel is essential to ensuring the sustainable growth of nuclear energy. However, post-processing facilities release copious amounts of emissions with high concentrations of nitrogen oxides (NO<sub>x</sub>), making the accurate measurement of their concentrations in radioactive settings greatly challenging. The application of machine learning strategies to predict NO<sub>x</sub> emissions offers a promising approach for improving the measurement and management of NO<sub>x</sub> in post-processing facilities, owing to their potential for cost reduction and operational expediency compared to conventional methods. Therefore, this study presents the outcomes of predictive activities for NO<sub>x</sub> emissions using machine learning. We employed a vector autoregression (VAR) model that considers the influence of other pollutants on NO<sub>x</sub> emissions. The results confirm that the VAR model sufficiently predicts NOx emissions. Furthermore, this study reveals the intricate interplay and feedback loops among various pollutants, thereby providing guidance for formulating comprehensive pollution control strategies. Finally, a lightweight and precise NO<sub>x</sub> forecasting model was developed by extracting the primary features affecting NO<sub>x</sub> predictions. This model has substantial significance for elevating the precision of pollutant emission forecasts and offers substantive support for the development and sustainable growth of the nuclear chemical industry.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 117-122"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100892","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0