Abstract The nuclear power industry is developing rapidly toward intelligence and scale, the digital twin was combined with the industrial interconnection technology to solve the key problems in the application of the digital twin, such as the three-dimensional model presentation, real-time data docking, and the improvement of intelligence degree. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0). Firstly, the three-dimensional twin of nuclear power equipment is constructed and the real-time update of twin data is realized based on the Node-EPICS event driver and Websocket communication protocol; Then, the communication interface with MySQL database is developed to realize the storage and management of data; Finally, the PID control system of molten salt circuit pipeline is integrated with back propagation neural network algorithm, and the efficiency and precision of temperature control system are improved by self-modification of weight. The results show that this system has the functions of three-dimensional display, network communication, data storage, and parameter optimization, and the data update cycle is raised to 100 ms, which can provide a certain reference value for the digital transformation of the nuclear monitoring field.
{"title":"Design and optimization of molten salt reactor monitoring system based on digital twin technology","authors":"Wen-qing Liu, Lifeng Han, Li Huang","doi":"10.1515/kern-2022-0055","DOIUrl":"https://doi.org/10.1515/kern-2022-0055","url":null,"abstract":"Abstract The nuclear power industry is developing rapidly toward intelligence and scale, the digital twin was combined with the industrial interconnection technology to solve the key problems in the application of the digital twin, such as the three-dimensional model presentation, real-time data docking, and the improvement of intelligence degree. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0). Firstly, the three-dimensional twin of nuclear power equipment is constructed and the real-time update of twin data is realized based on the Node-EPICS event driver and Websocket communication protocol; Then, the communication interface with MySQL database is developed to realize the storage and management of data; Finally, the PID control system of molten salt circuit pipeline is integrated with back propagation neural network algorithm, and the efficiency and precision of temperature control system are improved by self-modification of weight. The results show that this system has the functions of three-dimensional display, network communication, data storage, and parameter optimization, and the data update cycle is raised to 100 ms, which can provide a certain reference value for the digital transformation of the nuclear monitoring field.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"207 1","pages":"651 - 660"},"PeriodicalIF":0.5,"publicationDate":"2022-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76201292","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 Nuclear Facility (NF), during shutdown and startup, are in the essential need for reliable electric power that should be delivered by electric power grid to NF. Safe operation of NF needs a limited variation in both frequency and voltage.The reduction of power losses, improving voltage profile, and frequency in electric grid connected with NF can be achieved by optimally distributed generators (DGs) placement. This paper presents a mathematical model for multible types of DGs placement in electric grid feeding NF. Also, it proposes artificial intelligence solution methodology for active and reactive power DGs placement problem. The trained Adaptive Neuro-Fuzzy Inference System (ANFIS) with Cat Swarm Optimization algorithm (CSO) is used for optimal solution. The optimization technique is tested and validated by using different sizes of electric grid. Test results showed a more reliable and efficient approach compared with other approachs.
{"title":"Improving of electric network feeding nuclear facility based on multiple types DGs placement","authors":"A. Saleh, A. Adail","doi":"10.1515/kern-2022-0068","DOIUrl":"https://doi.org/10.1515/kern-2022-0068","url":null,"abstract":"Abstract Nuclear Facility (NF), during shutdown and startup, are in the essential need for reliable electric power that should be delivered by electric power grid to NF. Safe operation of NF needs a limited variation in both frequency and voltage.The reduction of power losses, improving voltage profile, and frequency in electric grid connected with NF can be achieved by optimally distributed generators (DGs) placement. This paper presents a mathematical model for multible types of DGs placement in electric grid feeding NF. Also, it proposes artificial intelligence solution methodology for active and reactive power DGs placement problem. The trained Adaptive Neuro-Fuzzy Inference System (ANFIS) with Cat Swarm Optimization algorithm (CSO) is used for optimal solution. The optimization technique is tested and validated by using different sizes of electric grid. Test results showed a more reliable and efficient approach compared with other approachs.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"21 1","pages":"13 - 20"},"PeriodicalIF":0.5,"publicationDate":"2022-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79645723","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}
Yasser Ammar, A. Elbaset, A. Adail, Sayed El. Araby
Abstract The dependently of the electrical grid is critical key point to safety of the nuclear research reactor (NRR) operation. This paper provides an optimization approach relying on optimal allocation of UPFC device to obtain higher electrical power quality of such nuclear facilities. The particle swarm optimization (PSO) technique was used to address the optimal UPFC allocation problem. The suggested approach is applied to the IEEE 33-bus test system, and results reveal that the suggested PSO is more efficient in minimizing total power losses and enhancing voltage profile using only one of UPFC device. The results show the technique is good method in this case.
{"title":"A sustainable solution to ensure the dependently and safety of electrical grid relying on optimal allocation of UPFC for research reactor","authors":"Yasser Ammar, A. Elbaset, A. Adail, Sayed El. Araby","doi":"10.1515/kern-2022-0057","DOIUrl":"https://doi.org/10.1515/kern-2022-0057","url":null,"abstract":"Abstract The dependently of the electrical grid is critical key point to safety of the nuclear research reactor (NRR) operation. This paper provides an optimization approach relying on optimal allocation of UPFC device to obtain higher electrical power quality of such nuclear facilities. The particle swarm optimization (PSO) technique was used to address the optimal UPFC allocation problem. The suggested approach is applied to the IEEE 33-bus test system, and results reveal that the suggested PSO is more efficient in minimizing total power losses and enhancing voltage profile using only one of UPFC device. The results show the technique is good method in this case.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"53 1","pages":"683 - 696"},"PeriodicalIF":0.5,"publicationDate":"2022-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83521587","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 This study looks for innovative methods to improve the overall performance of the U-20% Zr metallic fuel. The first solution is to swap out the helium gap for a ternary liquid metal bonded gap while the second involves minimizing the helium gap’s thickness to 0.04 mm in order to minimize its thermal resistance. The proposed solutions have been subjected to neutronic, thermal-hydraulic, and solid structure investigations, and their performance has been contrasted with that of a typical U-20% Zr metallic alloy with a 0.08 mm He-gap. According to neutronic analysis, the investigated fuel materials have almost identical neutronic performance. After using the LM bonded gap, both thermal-hydraulic and solid structure performance improved significantly. The performance of the U-20% Zr with 0.04 mm He-gap was moderate and unattractive to be used since it was deduced that its drawbacks outweighed its benefits.
摘要本研究旨在探索提高U-20% Zr金属燃料整体性能的创新方法。第一种解决方案是将氦隙换成三元液态金属键合隙，第二种解决方案是将氦隙的厚度减小到0.04毫米，以最小化其热阻。对所提出的溶液进行了中子、热水力和固体结构研究，并将其性能与典型的U-20% Zr金属合金的性能进行了对比，其he间隙为0.08 mm。根据中子分析，所研究的燃料材料具有几乎相同的中子性能。采用LM键合间隙后，热工性能和固相结构性能均有显著提高。0.04 mm he间隙的U-20% Zr的性能一般，不适合使用，因为它的缺点大于它的优点。
{"title":"Untraditional solution for enhancing the performance of U-20 % Zr metallic alloy as an ATF using liquid metal bonded gap","authors":"M. M. Mohsen, M. Abdel‐Rahman, A. Galahom","doi":"10.1515/kern-2022-0065","DOIUrl":"https://doi.org/10.1515/kern-2022-0065","url":null,"abstract":"Abstract This study looks for innovative methods to improve the overall performance of the U-20% Zr metallic fuel. The first solution is to swap out the helium gap for a ternary liquid metal bonded gap while the second involves minimizing the helium gap’s thickness to 0.04 mm in order to minimize its thermal resistance. The proposed solutions have been subjected to neutronic, thermal-hydraulic, and solid structure investigations, and their performance has been contrasted with that of a typical U-20% Zr metallic alloy with a 0.08 mm He-gap. According to neutronic analysis, the investigated fuel materials have almost identical neutronic performance. After using the LM bonded gap, both thermal-hydraulic and solid structure performance improved significantly. The performance of the U-20% Zr with 0.04 mm He-gap was moderate and unattractive to be used since it was deduced that its drawbacks outweighed its benefits.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"39 1","pages":"640 - 650"},"PeriodicalIF":0.5,"publicationDate":"2022-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79871193","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 pressurizer system of pressurized water reactor (PWR) maintains the reactor coolant system pressure during steady-state operation and limits pressure changes during transients. The in/out surge transients will cause pressure variations and they are controlled by either the spray system or the heater system. The spray system is actuated when the pressure exceeds a preset value. The heater system is initiated when the pressure falls below a preset value. The fundamental understanding and a reliable modeling of the pressurizer system behavior under steady state and transient conditions are needed to simulate overall nuclear power plant behavior. In the present study, an algorithm using Python 3.7 is developed to represent the dynamic behavior of the pressurizer system under steady-state and during in/out surge transients. Moreover, RELAP5 code is used to simulate the pressurizer system during the prescribed transients. The analysis and assessment results demonstrate satisfactory control performance during the in/out surge transients that guarantee the safety of PWR during operation. Also, the comparison between Python algorithm and RELAP5 model illustrates the capability and effectiveness of the Python algorithm for dynamic simulation and control.
{"title":"Pressurizer system dynamic model for transient control in PWR","authors":"H. Selim, N. El-Sahlamy","doi":"10.1515/kern-2022-0038","DOIUrl":"https://doi.org/10.1515/kern-2022-0038","url":null,"abstract":"Abstract The pressurizer system of pressurized water reactor (PWR) maintains the reactor coolant system pressure during steady-state operation and limits pressure changes during transients. The in/out surge transients will cause pressure variations and they are controlled by either the spray system or the heater system. The spray system is actuated when the pressure exceeds a preset value. The heater system is initiated when the pressure falls below a preset value. The fundamental understanding and a reliable modeling of the pressurizer system behavior under steady state and transient conditions are needed to simulate overall nuclear power plant behavior. In the present study, an algorithm using Python 3.7 is developed to represent the dynamic behavior of the pressurizer system under steady-state and during in/out surge transients. Moreover, RELAP5 code is used to simulate the pressurizer system during the prescribed transients. The analysis and assessment results demonstrate satisfactory control performance during the in/out surge transients that guarantee the safety of PWR during operation. Also, the comparison between Python algorithm and RELAP5 model illustrates the capability and effectiveness of the Python algorithm for dynamic simulation and control.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"56 1","pages":"672 - 682"},"PeriodicalIF":0.5,"publicationDate":"2022-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89870396","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}
Yasser Ammar, A. Elbaset, A. Adail, Sayed. M. S. El Araby, A. Saleh
Abstract UPFC device is discussed in this paper along with their models and functions. Moreover, the suggested and the complementally approaches in the current research study. As a result, the methods are divided into three divisions, which are sensitivity analysis based methods, conventional optimization based methods and artificial intelligence (AI) based methods. In addition, artificial intelligence based methods plays a major role in reducing the search space region. However, to optimize the resulting benefits, the placement, sizing and parameter of UPFC device should be determined. This paper presents and discusses in depth an overall review of the last two decades’ studies, including proposed and comparative methods and strategies, approaches, objective functions, UPFC device tools utilized, limitations, contingency situations and all parameters evaluated and simulated. This paper also provides an analysis of UPFC’s various benefits and uses of power flow studies, such as, power loss mitigation, system load ability improvement, power system security, enhancement of voltage stability, cost of generation and UPFC installation and utilizing specific optimization techniques. Therefore, a more weighted overview of the proposed method is presented focused on artificial intelligence optimization methods.
{"title":"A review on optimal UPFC device placement in electric power systems","authors":"Yasser Ammar, A. Elbaset, A. Adail, Sayed. M. S. El Araby, A. Saleh","doi":"10.1515/kern-2022-0063","DOIUrl":"https://doi.org/10.1515/kern-2022-0063","url":null,"abstract":"Abstract UPFC device is discussed in this paper along with their models and functions. Moreover, the suggested and the complementally approaches in the current research study. As a result, the methods are divided into three divisions, which are sensitivity analysis based methods, conventional optimization based methods and artificial intelligence (AI) based methods. In addition, artificial intelligence based methods plays a major role in reducing the search space region. However, to optimize the resulting benefits, the placement, sizing and parameter of UPFC device should be determined. This paper presents and discusses in depth an overall review of the last two decades’ studies, including proposed and comparative methods and strategies, approaches, objective functions, UPFC device tools utilized, limitations, contingency situations and all parameters evaluated and simulated. This paper also provides an analysis of UPFC’s various benefits and uses of power flow studies, such as, power loss mitigation, system load ability improvement, power system security, enhancement of voltage stability, cost of generation and UPFC installation and utilizing specific optimization techniques. Therefore, a more weighted overview of the proposed method is presented focused on artificial intelligence optimization methods.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"18 1","pages":"661 - 671"},"PeriodicalIF":0.5,"publicationDate":"2022-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78389148","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 paper presents a Computational Fluid Dynamics (CFD) methodology to model gas-liquid boiling flow in a full scale 5 × 5 rod bundle with spacer grid typical in Pressurized Water Reactor (PWR) fuel rod bundle. The CFD modeling method is developed based on the STAR-CCM+ CFD code, including the Eulerian-Eulerian two-fluid model and the improved wall heat partitioning model. The OECD/NRC PWR Sub-channel and Bundle Tests (PSBT) are used as a numerical benchmark to assess the simulation quantitatively. The simulated geometry is a full scale of 5 × 5 fuel rod bundle with 17 spacers, including 7 mixing vane spacers (MV), 8 simple spacers (SS) and 2 non-mixing vane spacers (NMV). The present simulated results are in good agreement with the experimental results, the average error of the simulated cross-section void fraction is less than 20%. Based on the simulations, the axial distributions of second flow intensity, the rod surface temperature, bulk fluid temperature, and the void fraction are discussed. The results show that the spacer grid structures, especially the mixing vane, play an essential part in spreading the bubbles, reducing the void fraction and the rod surface temperature.
{"title":"CFD simulation on flow boiling in full scale 5 × 5 rod bundle","authors":"Bing Ren, F. Gan, Ping Yang","doi":"10.1515/kern-2022-0031","DOIUrl":"https://doi.org/10.1515/kern-2022-0031","url":null,"abstract":"Abstract The paper presents a Computational Fluid Dynamics (CFD) methodology to model gas-liquid boiling flow in a full scale 5 × 5 rod bundle with spacer grid typical in Pressurized Water Reactor (PWR) fuel rod bundle. The CFD modeling method is developed based on the STAR-CCM+ CFD code, including the Eulerian-Eulerian two-fluid model and the improved wall heat partitioning model. The OECD/NRC PWR Sub-channel and Bundle Tests (PSBT) are used as a numerical benchmark to assess the simulation quantitatively. The simulated geometry is a full scale of 5 × 5 fuel rod bundle with 17 spacers, including 7 mixing vane spacers (MV), 8 simple spacers (SS) and 2 non-mixing vane spacers (NMV). The present simulated results are in good agreement with the experimental results, the average error of the simulated cross-section void fraction is less than 20%. Based on the simulations, the axial distributions of second flow intensity, the rod surface temperature, bulk fluid temperature, and the void fraction are discussed. The results show that the spacer grid structures, especially the mixing vane, play an essential part in spreading the bubbles, reducing the void fraction and the rod surface temperature.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"99 1","pages":"627 - 639"},"PeriodicalIF":0.5,"publicationDate":"2022-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77219366","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 water level control system implicated in the nuclear steam generator has played an essential role in unexpected shutdowns of the power plant. According to reports, about 25% of the emergency power blackouts are caused by improper level control systems. The effectiveness of optimization methods in designing a controller is currently proved in different disciplines. The novelty of this paper is the proportional integral derivative (PID) controller tuning of nuclear steam generator by particle swarm optimization (PSO) and genetic algorithm (GA) for the lowest steady-state error, overshoot, undershoot, and settling time. Different types of the cost function are employed to obtain the controller gains. The integral of the absolute error (IAE), square error (ISE), time-weighted average error (ITAE), time-weighted square error (ITSE), and a weighted function based on overshoot, undershoot, and settling time are used. The gain scheduling of optimized PIDs is used to have an entire operating range control system. The desired load-following and stability of the optimized PID controller are investigated under both time and frequency domains using trajectory tracking, disturbance rejection, and Nichols chart criterion.
{"title":"Optimization of PID controller for water level control of the nuclear steam generator using PSO and GA","authors":"O. Safarzadeh, Amir Tizdast","doi":"10.1515/kern-2021-1038","DOIUrl":"https://doi.org/10.1515/kern-2021-1038","url":null,"abstract":"Abstract The water level control system implicated in the nuclear steam generator has played an essential role in unexpected shutdowns of the power plant. According to reports, about 25% of the emergency power blackouts are caused by improper level control systems. The effectiveness of optimization methods in designing a controller is currently proved in different disciplines. The novelty of this paper is the proportional integral derivative (PID) controller tuning of nuclear steam generator by particle swarm optimization (PSO) and genetic algorithm (GA) for the lowest steady-state error, overshoot, undershoot, and settling time. Different types of the cost function are employed to obtain the controller gains. The integral of the absolute error (IAE), square error (ISE), time-weighted average error (ITAE), time-weighted square error (ITSE), and a weighted function based on overshoot, undershoot, and settling time are used. The gain scheduling of optimized PIDs is used to have an entire operating range control system. The desired load-following and stability of the optimized PID controller are investigated under both time and frequency domains using trajectory tracking, disturbance rejection, and Nichols chart criterion.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"35 1","pages":"597 - 606"},"PeriodicalIF":0.5,"publicationDate":"2022-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80420513","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}
Beya Heritier, Rowayda F. Mahmoud, A. El Saghir, M. Shaat, A. Badawi
Abstract Democratic Republic of Congo (DRC) has a TRIGA mark II research reactor called TRICO II, its design power is 1.00 MW. The reactor was in extended shutdown state since November 2004. The DRC government has decided to resume its operation using the last uploaded core. One of the safety features to be determined before putting the spent fuel into the reactor core is the calculation of its excess reactivity, radionuclide inventories as well as its discharge burn-up. The spent fuel was modeled and simulated by using Monte Carlo software, MCNPX code. The input data and the horizontal and vertical modeling for the fuel pins, control rods and moderator were done. The model results were validated by calculating the effective delayed neutron fraction (β eff) and the worth of the control rods. The results of the criticality and fuel burn-up were compared with the reference design parameters and with the experimental measurements and it were found in good agreement. The calculations showed that the last uploaded core has 47.00 g of 235U which represent only 2% of fissile materials. The depletion analysis results showed that the highest radio-activities come from 151Sm, 137Cs, 90Y, 90Sr and 85Kr.
刚果民主共和国(DRC)有一座TRIGA mark II型研究堆，称为TRICO II，其设计功率为1.00 MW。该反应堆自2004年11月以来一直处于长时间关闭状态。刚果民主共和国政府决定使用最后上传的核恢复其运行。在将乏燃料放入反应堆堆芯之前，需要确定的安全特性之一是计算其过度反应性、放射性核素库存以及排放燃耗。利用蒙特卡罗软件MCNPX代码对乏燃料进行了建模和仿真。完成了燃料销、控制棒和慢化剂的输入数据和水平、垂直建模。通过计算有效延迟中子分数(β eff)和控制棒的价值，对模型结果进行了验证。将临界值和燃耗值与参考设计参数和实验测量值进行了比较，结果吻合较好。计算表明，最后上传的堆芯含有47.00克235U，仅占裂变材料的2%。损耗分析结果表明，151Sm、137Cs、90Y、90Sr和85Kr的放射性活性最高。
{"title":"Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor","authors":"Beya Heritier, Rowayda F. Mahmoud, A. El Saghir, M. Shaat, A. Badawi","doi":"10.1515/kern-2022-0016","DOIUrl":"https://doi.org/10.1515/kern-2022-0016","url":null,"abstract":"Abstract Democratic Republic of Congo (DRC) has a TRIGA mark II research reactor called TRICO II, its design power is 1.00 MW. The reactor was in extended shutdown state since November 2004. The DRC government has decided to resume its operation using the last uploaded core. One of the safety features to be determined before putting the spent fuel into the reactor core is the calculation of its excess reactivity, radionuclide inventories as well as its discharge burn-up. The spent fuel was modeled and simulated by using Monte Carlo software, MCNPX code. The input data and the horizontal and vertical modeling for the fuel pins, control rods and moderator were done. The model results were validated by calculating the effective delayed neutron fraction (β eff) and the worth of the control rods. The results of the criticality and fuel burn-up were compared with the reference design parameters and with the experimental measurements and it were found in good agreement. The calculations showed that the last uploaded core has 47.00 g of 235U which represent only 2% of fissile materials. The depletion analysis results showed that the highest radio-activities come from 151Sm, 137Cs, 90Y, 90Sr and 85Kr.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"55 1","pages":"615 - 624"},"PeriodicalIF":0.5,"publicationDate":"2022-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77596294","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 Safe and efficient storage of spent fuel elements is an important aspect of the safety and economy of nuclear reactors. The present work investigates the thermal-hydraulic behaviour of the cooling process for the nuclear spent fuel stored in the material testing reactor auxiliary pool. The parameters affected by the spent fuel cooling accuracy, the decay power of spent fuel and the initial temperature of the coolant pool are studied. These parameters are simulated by developing a model using thermal-hydraulic computational fluid dynamics, ANSYS FLUENT 17.2 Code. The developed model is evaluated by the previous measurements; an experimental test rig is designed and constructed to investigate the thermal-hydraulic behaviour of the natural circulation cooling of the nuclear spent fuel. The present study uses the validated model to simulate numerically the forced convection heat transfer for spent fuel pools. Various coolant velocities and decay powers are examined. Also, the thermal-hydraulic behaviour of the nuclear spent fuel is studied in transient mode; the initial temperature is raised to 338K. The results show the spent fuel cooling improves as the coolant velocity increases. A good agreement was identified after comparing experimental results with the investigated model.
{"title":"Computational fluid dynamics simulation of material testing reactor spent fuel cooling in wet storage","authors":"S. Abdel-Latif, S. Elnaggar","doi":"10.1515/kern-2022-0039","DOIUrl":"https://doi.org/10.1515/kern-2022-0039","url":null,"abstract":"Abstract Safe and efficient storage of spent fuel elements is an important aspect of the safety and economy of nuclear reactors. The present work investigates the thermal-hydraulic behaviour of the cooling process for the nuclear spent fuel stored in the material testing reactor auxiliary pool. The parameters affected by the spent fuel cooling accuracy, the decay power of spent fuel and the initial temperature of the coolant pool are studied. These parameters are simulated by developing a model using thermal-hydraulic computational fluid dynamics, ANSYS FLUENT 17.2 Code. The developed model is evaluated by the previous measurements; an experimental test rig is designed and constructed to investigate the thermal-hydraulic behaviour of the natural circulation cooling of the nuclear spent fuel. The present study uses the validated model to simulate numerically the forced convection heat transfer for spent fuel pools. Various coolant velocities and decay powers are examined. Also, the thermal-hydraulic behaviour of the nuclear spent fuel is studied in transient mode; the initial temperature is raised to 338K. The results show the spent fuel cooling improves as the coolant velocity increases. A good agreement was identified after comparing experimental results with the investigated model.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"6 1","pages":"570 - 578"},"PeriodicalIF":0.5,"publicationDate":"2022-08-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88843400","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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