• Title/Summary/Keyword: Thermal power plants

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Thermal Design Analysis of Triple-Pressure Heat Recovery Steam Generator and Steam Turbine Systems (삼중압 열회수 증기발생기와 중기터빈 시스템의 열설계 해석)

  • Kim, Dong-Seop;Lee, Bong-Ryeol;No, Seung-Tak;Sin, Heung-Tae;Jeon, Yong-Jun
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.26 no.3
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    • pp.507-514
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    • 2002
  • A computation routine, capable of performing thermal design analysis of the triple-pressure bottoming system (heat recovery steam generator and steam turbine) of combined cycle power plants, is developed. It is based on thermal analysis of the heat recovery steam generator and estimation of its size and steam turbine power. It can be applied to various parametric analyses including optimized design calculation. This paper presents analysis results for the effects on the design performance of heat exchanger arrangements at intermediate and high temperature parts as well as steam pressures. Also examined is the effect of steam sources for deaeration on design performance.

A Study on the Development of Plugging Margin Evaluation Method Reflected the Fouling of a Shell-and-Tube Heat Exchanger (다관원통형 열교환기의 파울링 현상을 고려한 관막음 여유 평가법 개발 연구)

  • Hwang, Kyeong-Mo;Jin,Tae-Eun
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.28 no.11
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    • pp.1384-1389
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    • 2004
  • As operating time of heat exchangers progresses, fouling generated by water-borne deposits and the number of plugged tubes increase and thermal performance decreases. Both fouling and tube plugging are known to interfere with normal flow characteristics and to reduce thermal efficiencies of heat exchangers. The heat exchangers of domestic nuclear power plants have been analyzed in terms of the heat flux and heat transfer coefficient at test conditions as a means of heat exchanger management. Except for the fouling level generated in operation of heat exchangers, also, all of the tubes of heat exchangers have been replaced when the number of plugged tubes exceeds the plugging criteria based on design performance sheet. This paper describes the plugging margin evaluation mettled reflected the fouling of shell-and-tube heat exchangers, which can evaluate the thermal performance for heat exchangers, estimate the future fouling variations, and reflect the current fouling level. To identify the effectiveness of the developed method, the fouling and plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power plant.

A Study on the Electrical Properties of Ethylene Propylene Rubber by Thermal Treatment and Irradiation (방사선 및 열처리에 의한 에틸렌프로필렌 고무의 전기적 특성에 관한 연구)

  • 이성일
    • Journal of the Korea Safety Management & Science
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    • v.4 no.4
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    • pp.137-146
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    • 2002
  • In order to investigate the effect of irradiation by $^{60}Co-\gamma$rays as well as the e thermal treatment on the dielectric deterioration in ethylene propylene rubber, insulating material for electric cables used in atomic power plants, charging discharging current, residual built- up voltage and dielectric properties are measu discussed in this study. Variance in the characteristic of relative dielectric constant as a function of tem was observed in relatively high dose of irradiation. Since glass transition tem appeared at tens of degree Celsius below zero, the characteristic is attributed orientation polarization. Dielectric loss is generally increased, with increasing d irradiation in the characteristic of dielectric loss as a function of temperature, No d loss by thermal treatment was observed. Dielectric resistance decreases with increa of irradiation in the characteristic of charging current as a function of temperature be considered that dielectric resistance seems to be recovered by thermal treatm characteristic of discharging current as a function of time in the specimen less ir become similar to that of the unirradiated, when thermal treated. A peak is shown residual built- up voltage as a function of time, and the corresponding time of the shorten as increasing dose of irradiation. It is also observed that the corresponding the peak is lengthened by thermal treatment.

Effect of multiple-failure events on accident management strategy for CANDU-6 reactors

  • YU, Seon Oh;KIM, Manwoong
    • Nuclear Engineering and Technology
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    • v.53 no.10
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    • pp.3236-3246
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    • 2021
  • Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

A Study on Generator Maintenance Scheduling Considering Renewable Energy Generators (신재생에너지 발전원을 고려한 발전기 예방정비계획수립에 관한 연구)

  • Lee, Yeonchan;Oh, Ungjin;Choi, Jaeseok;Jung, Myeunghoon
    • The Transactions of The Korean Institute of Electrical Engineers
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    • v.67 no.5
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    • pp.601-610
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    • 2018
  • The purpose of this paper is to establish a new optimum Generator Maintenance Scheduling(GMS) considering renewable energy generator. In this paper, the total renewable energy generation is estimated using hourly capacity factor of renewable energy generator. The GMS was optimized with the objective function of maximizing the minimum reserve rate, minimizing the probabilistic production cost, minimizing the loss of load expectation, and minimizing $CO_2$ emissions. In the case study of this paper, GMS considering renewable energy and GMS not considering renewable energy are shown by each objective function. And it shows scenarios of the reliability, the environmental and economical factors when two nuclear power plants inputted and ten coal thermal power plants shut downed respectively.

High-Temperature Stability Evaluation of Various Surface Treated Layers of Materials for Ultra-Super Critical Power Plants (초초임계압 발전용 소재의 표면처리층의 고온 안정성 평가)

  • Ryu, K.H.;Song, T.K.;Lee, J.H.;Kim, G.S.;Lee, S.H.;Urm, K.W.
    • Korean Journal of Materials Research
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    • v.16 no.5
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    • pp.329-335
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    • 2006
  • In order to improve thermal efficiency of the fossil fuel power plants, we need to develop advanced materials with superior durability in the ultra-super critical state, which requires surface modifications for superior surface properties. In this study, we coated the Incoloy 901 and 12-17Cr steels for turbine buckets and valves with nitriding, boriding, and $Cr_3C_2-NiCr$ HVOF(high velocity oxygen flow) method. Then the samples were heat treated at $650^{\circ}C$ for 100 hours in vacuum. We analyzed the evolution behaviors of nitrides such as $Fe_3N,\;Fe_4N$, and CrN and borides such as FeB and $Fe_2B$ with XRD and SEM/EDS by comparing hardnesses and compositions of the coated layers before and after the heat treatments.

A Study on Control Characteristic and Application of Optimal Modulation Controller for HVDC Transmission System (초고압 직류 송전 시스템에 대한 최적 변조 제어기의 적용 및 제어 특성에 관한 연구)

  • Lee, J.M.;Hur, D.R.;Chung, D.I.;Chung, H.H.
    • Proceedings of the KIEE Conference
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    • 1999.07c
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    • pp.1318-1320
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    • 1999
  • Recently, according to the growth of national economy and the improvement of lining conditions, electric power demand is increasing gradually. So it is being examined to construct large thermal power plants or nuclear plants. For the effective use of lands and for the economy of generations sites, the distance between generation and demand locations becomes farther and farther. At the same time, people desire higher quality or electric power. So in this paper, the optimal modulation controller for HVDC transmission system are designed by a recursive algorithm that determines the state weighting matrix Q of a linear quadratic performance. It means that the application of optimal modulation controller in HVDC transmission system can contribute the propriety to the improvement of the stability in HVDC transmission system.

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Evaluation of Post-LOCA Long Term Cooling Performance in Korean Standard Nuclear Power Plants

  • Bang, Young-Seok;Jung, Jae-Won;Seul, Kwang-Won;Kim, Hho-Jung
    • Nuclear Engineering and Technology
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    • v.33 no.1
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    • pp.12-24
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    • 2001
  • The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

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FATIGUE LIFE ASSESSMENT OF REACTOR COOLANT SYSTEM COMPONENTS BY USING TRANSFER FUNCTIONS OF INTEGRATED FE MODEL

  • Choi, Shin-Beom;Chang, Yoon-Suk;Choi, Jae-Boong;Kim, Young-Jin;Jhung, Myung-Jo;Choi, Young-Hwan
    • Nuclear Engineering and Technology
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    • v.42 no.5
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    • pp.590-599
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    • 2010
  • Recently, efficient operation and practical management of power plants have become important issues in the nuclear industry. In particular, typical aging parameters such as stress and cumulative usage factor should be determined accurately for continued operation of a nuclear power plant beyond design life. However, most of the major components have been designed via conservative codes based on a 2-D concept, which do not take into account exact boundary conditions and asymmetric geometries. The present paper aims to suggest an effective fatigue evaluation methodology that uses a prototype of the integrated model and its transfer functions. The validity of the integrated 3-D Finite Element (FE) model was proven by comparing the analysis results of individual FE models. Also, mechanical and thermal transfer functions, known as Green's functions, were developed for the integrated model with the standard step input. Finally, the stresses estimated from the transfer functions were compared with those obtained from detailed 3-D FE analyses results at critical locations of the major components. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system, and this combination, will enhance the continued operations of old nuclear power plants.

Real-time estimation of break sizes during LOCA in nuclear power plants using NARX neural network

  • Saghafi, Mahdi;Ghofrani, Mohammad B.
    • Nuclear Engineering and Technology
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    • v.51 no.3
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    • pp.702-708
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    • 2019
  • This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5% -100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.