Kerntechnik最新文献_第7页

A review on optimal UPFC device placement in electric power systems 电力系统中UPFC器件最优放置的研究进展

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-10-03 DOI: 10.1515/kern-2022-0063

Yasser Ammar, A. Elbaset, A. Adail, Sayed. M. S. El Araby, A. Saleh

{"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}

引用次数: 1

CFD simulation on flow boiling in full scale 5 × 5 rod bundle 全尺寸5 × 5棒束流动沸腾的CFD模拟

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-09-30 DOI: 10.1515/kern-2022-0031

Bing Ren, F. Gan, Ping Yang

{"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}

引用次数: 0

Optimization of PID controller for water level control of the nuclear steam generator using PSO and GA 基于粒子群算法和遗传算法的核蒸汽发生器水位PID控制器优化

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-09-15 DOI: 10.1515/kern-2021-1038

O. Safarzadeh, Amir Tizdast

{"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}

引用次数: 1

Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor TRIGA mark II型研究堆乏燃料安全评价与管理

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-08-29 DOI: 10.1515/kern-2022-0016

Beya Heritier, Rowayda F. Mahmoud, A. El Saghir, M. Shaat, A. Badawi

{"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}

引用次数: 0

Computational fluid dynamics simulation of material testing reactor spent fuel cooling in wet storage 材料试验堆乏燃料湿贮存冷却计算流体动力学模拟

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-08-17 DOI: 10.1515/kern-2022-0039

S. Abdel-Latif, S. Elnaggar

{"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}

引用次数: 0

Evaluation and integral analysis of ADS and CMT failures during AP1000 SBLOCA with ASYST VER 3 simulation code AP1000 SBLOCA过程中ADS和CMT故障的评估与积分分析

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-08-12 DOI: 10.1515/kern-2022-0010

O. E. Osman, A. Badawi, Ayah E. Elshahat

{"title":"Evaluation and integral analysis of ADS and CMT failures during AP1000 SBLOCA with ASYST VER 3 simulation code","authors":"O. E. Osman, A. Badawi, Ayah E. Elshahat","doi":"10.1515/kern-2022-0010","DOIUrl":"https://doi.org/10.1515/kern-2022-0010","url":null,"abstract":"Abstract This research focuses on verifying the importance of the ADS and the CMT, by using the ASYST code. We evaluated the role of these two components by postulating the failure of the ADS as a single failure approach and the failure of the CMT with ADS failure as multiple failures approach during hypothetical SBLOCA conditions. These accidents acted as confounding factors distorting the AP1000 PSS. We investigated the reactor and safety system behavior during the SBLOCA. We evaluated the importance and effectiveness of two components in reducing and mitigating the consequences of the accident. We checked the effectiveness of these components by comparing the importunity-related issues with and without these components during the accidents. We found that the ADS decreased the pressure, allowing natural circulation to quench the reactor core during the LOCA. During the failure of ADS, the vapor bubbles formed in the reactor vessel covering the fuel rods increased their temperature. The CMT borated water feeding quenched the actinides decay heat. The non-existence of the CMT resulted in decreasing the RCS. ASYST was compared to NOTRUMP to validate it capability to analyze thermal phenomena during accidents. It was found that in the AP1000, the ADS and CMT were considered as the overall importunity of the others PSS.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"61 1","pages":"556 - 569"},"PeriodicalIF":0.5,"publicationDate":"2022-08-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73861440","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Transient analysis of MTR research reactor during fast and slow loss of flow accident MTR研究堆快、慢失流事故瞬态分析

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-08-10 DOI: 10.1515/kern-2022-0052

H. Selim

{"title":"Transient analysis of MTR research reactor during fast and slow loss of flow accident","authors":"H. Selim","doi":"10.1515/kern-2022-0052","DOIUrl":"https://doi.org/10.1515/kern-2022-0052","url":null,"abstract":"Abstract The main objective for reactor safety is to keep the fuel in a safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accidents (DBAs), one of them is the loss of flow accident (LOFA), is required for assessing reactor safety. In this research, the safety aspects of 22 MW MTR research reactor under steady state and during loss of flow accident is studied. The flow transients considered include fast loss of flow accident (FLOFA) and slow loss of flow accident (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The analysis is done using PARET, a neutronics-hydrodynamics-heat transfer code. The transients were initiated from a full power with a flow trip point at 85% nominal. The calculated parameters are the temperatures of different components (fuel, clad and coolant) as a function of time for the hot channel. The results indicate that in both accidents the calculated maximum cladding surface temperature for the hottest channel of the reactor core does not exceed the allowable safety limit and the fuel integrity is maintained.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"1 1","pages":"529 - 534"},"PeriodicalIF":0.5,"publicationDate":"2022-08-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83975345","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Experimental study on direct contact condensation of saturated steam at low mass flux in subcooled quiescent water 过冷静水中低质量通量饱和蒸汽直接接触冷凝实验研究

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-08-10 DOI: 10.1515/kern-2021-1059

M. A. Kaleem, Ajmal Shah, M. Iqbal, A. Quddus, Atif Mehmood, A. Riaz, M. K. Ayub

{"title":"Experimental study on direct contact condensation of saturated steam at low mass flux in subcooled quiescent water","authors":"M. A. Kaleem, Ajmal Shah, M. Iqbal, A. Quddus, Atif Mehmood, A. Riaz, M. K. Ayub","doi":"10.1515/kern-2021-1059","DOIUrl":"https://doi.org/10.1515/kern-2021-1059","url":null,"abstract":"Abstract The phenomenon of saturated steam jet injection in subcooled quiescent water has many practical applications including in heat exchangers, steam jet pumps, steam dumping systems in nuclear plants, etc. The experimental setup is designed and fabricated indigenously to investigate this phenomenon at lower mass fluxes ∼120 and 150 kg/m2 s. The steam jet of conical shape has been observed for all the test conditions. The recorded axial temperature distribution showed that near the nozzle region, the temperature is governed by the saturated condition of steam while the later region is dependent on the water pool temperature. The maximum temperature is observed to be at the center of the jet. It has been found that the dimensionless penetration length of the steam jet in water is directly dependent on both the temperature of the water pool and the mass flux of steam. The dimensionless jet length has been found in the range ∼1.54–2.02 and 2.07–2.19 for mass fluxes ∼120 and 150 kg/m2 s, respectively. The average heat transfer coefficient has been found in the range ∼1.97–2.37 MW/m2 K.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"60 233 1","pages":"547 - 555"},"PeriodicalIF":0.5,"publicationDate":"2022-08-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83300357","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Determination of heat flux leading to the onset of flow instability in MTR reactors MTR反应器中引起流动不稳定的热通量的测定

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-07-14 DOI: 10.1515/kern-2022-0046

S. E. El-Morshedy

{"title":"Determination of heat flux leading to the onset of flow instability in MTR reactors","authors":"S. E. El-Morshedy","doi":"10.1515/kern-2022-0046","DOIUrl":"https://doi.org/10.1515/kern-2022-0046","url":null,"abstract":"Abstract The prediction of heat flux leading to the Onset of Flow Instability (OFI) phenomena is an important consideration in the design of Material Testing Reactors (MTR) due to the possibility of flow excursion during postulated accident. From the thermal-hydraulic point of view, OFI is the critical phenomenon limiting MTR reactor power. In a previous work, an empirical correlation is developed to predict the subcooling at OFI in narrow vertical rectangular channels simulating a coolant channel of MTR. In the present work, an innovative model to determine the heat flux leading to OFI in MTR reactors is introduced based on the previous correlation. The developed model gives a very low deviation of only 1.65% from the experimental data of Whittle & Forgan that covers a wide range of MTR operating conditions. The heat flux leading to OFI is also predicted by both Whittle & Forgan and Fabrega correlations for comparison. The present model is then applied on the IAEA 10 MW MTR generic reactor to predict the Best-Estimate (BE) and Best-Estimate-Plus-Uncertainty (BEPU) Onset of Flow Instability Ratio (OFIR) and the power leading to OFI as well as the bubble detachment parameter under different coolant velocities and inlet temperatures. The model is also used to predict both the OFIR and bubble detachment parameter in the reactor under unprotected Loss-of-Flow transient for exponential flow decay with a time constant of 1.0 s (fast LOFA), 10, 15 and 25 s (slow LOFA) from a power level of 10 MW. For BEPU calculation, a combined statistical method with direct propagation of errors is adapted to treat the uncertainty factors for fuel fabrication and measured parameters in the BEPU calculation. The model results is analyzed and discussed.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"7 1","pages":"535 - 546"},"PeriodicalIF":0.5,"publicationDate":"2022-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81433846","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The role of advanced nuclear reactors in non-electrical and strategic applications, producing sustainable energy supplies and reducing the greenhouse gasses 先进核反应堆在非电力和战略应用中的作用，产生可持续能源供应和减少温室气体

IF 0.5 4区工程技术

Kerntechnik Pub Date : 2022-07-14 DOI: 10.1515/kern-2022-0029

A. Hedayat

{"title":"The role of advanced nuclear reactors in non-electrical and strategic applications, producing sustainable energy supplies and reducing the greenhouse gasses","authors":"A. Hedayat","doi":"10.1515/kern-2022-0029","DOIUrl":"https://doi.org/10.1515/kern-2022-0029","url":null,"abstract":"Abstract Nowadays, nuclear reactors became extremely fascinating not only for most of the nuclear communities but also for the prominent energy suppliers to fix the global warming effects worldwide. This paper presents a review of the last advances, applications, and challenges of nuclear reactors. Different types and classifications are introduced. Advantages and disadvantages are discussed for best decision-making. Next, nuclear safety is also discussed as the most important challenging subject to develop nuclear reactors worldwide. They are specially mentioned to find the key solution for the future of nuclear energy. A brief review of nuclear roadmaps is compared with other clean green technologies as well. Estimated prospects for projects timelines and progressions of new nuclear reactors are also presented and discussed briefly. Studies confirmed that nuclear reactors are not only required for developing non-electrical applications or even high-tech systems but also they are extremely profitable to restrict global warming effects. Finally, the solution is to enhance the markets of the nuclear reactors, especially the matured Gen III+ Pressurized Water Reactors (PWRs) to resolve short-term problems as well as advanced futuristic developing Small Modular Reactors (SMRs) for the mid-term and long-term strategies. Moreover, research reactors especially advanced Multi-Purpose Research Reactors (MPRR) are necessary tools to develop both nuclear power plants and other advanced technologies as well as the modern Micro Modular Reactors.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"190 1","pages":"579 - 596"},"PeriodicalIF":0.5,"publicationDate":"2022-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77100037","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0