Dissertations / Theses on the topic 'Windscale Nuclear Power Station'
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Boyes, Haydn. "Sensitivity analysis of the secondary heat balance at Koeberg Nuclear Power Station." Master's thesis, Faculty of Engineering and the Built Environment, 2021. http://hdl.handle.net/11427/33686.
Full textDinoko, Tshepo Samuel. "Modeling of the dispersion of radionuclides around a nuclear power station." Thesis, University of the Western Cape, 2009. http://etd.uwc.ac.za/index.php?module=etd&action=viewtitle&id=gen8Srv25Nme4_3451_1360933219.
Full textNuclear reactors release small amounts of radioactivity during their normal operations. The most common method of calculating the dose to the public that results from such releases uses Gaussian Plume models. We are investigating these methods using CAP88-PC, a computer code developed for the Environmental Protection Agency (EPA) in the USA that calculates the concentration of radionuclides released from a stack using Pasquill stability classification. A buoyant or momentum driven part is also included. The uptake of the released radionuclide by plants, animals and humans, directly and indirectly, is then calculated to obtain the doses to the public. This method is well established but is known to suffer from many approximations and does not give answers that are accurate to be better than 50% in many cases. More accurate, though much more computer-intensive methods have been developed to calculate the movement of gases 
using fluid dynamic models. Such a model, using the code FLUENT can model complex terrains and will also be investigated in this work. This work is a preliminary study to compare the results of the traditional Gaussian plume model and a fluid dynamic model for a simplified case. The results indicate that Computational Fluid Dynamics calculations give qualitatively similar results with the possibility of including much more effects than the simple Gaussian plume model.
程明錦 and Ming-kam Eric Ching. "A regional atmospheric dispersion model for Daya Bay Nuclear Power Station." Thesis, The University of Hong Kong (Pokfulam, Hong Kong), 1990. http://hub.hku.hk/bib/B31209634.
Full textRylands, Naasef. "Condition monitoring of induction motors in the nuclear power station environment." Master's thesis, University of Cape Town, 2018. http://hdl.handle.net/11427/29686.
Full textChing, Ming-kam Eric. "A regional atmospheric dispersion model for Daya Bay Nuclear Power Station /." [Hong Kong] : University of Hong Kong, 1990. http://sunzi.lib.hku.hk/hkuto/record.jsp?B12993104.
Full textSimons, Rowena Chrystal. "An exploratory analysis of quality management audit findings at a nuclear power station." Thesis, Cape Peninsula University of Technology, 2016. http://hdl.handle.net/20.500.11838/2382.
Full textThe quality assurance role is an essential function in high risk industries such as the nuclear power industry where process failures can potentially have catastrophic results. As part of mitigating the risk inherent in such industries, the need for reliable quality assurance cannot be over-emphasised. Underpinning a reliable quality assurance function, lies the need for effective identification of risk; as well as effective decision making processes by competent auditors. A nuclear quality assurance (QA) department has noted an increase in the variability of its audit outcomes, which has resulted in the value of the audit process being questioned by various stakeholders. The research endeavoured to: explore and describe the practice amongst auditors when rating audit findings; potentially identify reasons for inconsistencies amongst auditors when rating findings; and provide recommendations to improve both the consistency amongst auditors when rating audit finding and the overall performance of the audit process. An exploratory study using the Delphi technique was adopted to enable multiple iterations of qualitative and quantitative data collection and analysis, mimicking elements of a sequential exploratory strategy.
Leung, Wing-mo. "Age dependency of the radiological impact of the daya bay nuclear power station on the local population /." [Hong Kong] : University of Hong Kong, 1994. http://sunzi.lib.hku.hk/hkuto/record.jsp?B13597292.
Full textBezuidenhout, Jandré Albert. "Signature analysis of the primary components of the Koeberg nuclear power station / J.A. Bezuidenhout." Thesis, North-West University, 2010. http://hdl.handle.net/10394/4387.
Full textThesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.
Oudet, Alexandre. "Design and optimization of the HVAC system for a nuclear power plant demineralization station." Thesis, KTH, Energiteknik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-192184.
Full textDuring nuclear power plants shutdown many people could be deprived of electricity and it would have a negative impact both on the company’s image and on people activities. As a consequence, availability of equipments in the different buildings which compose the power plant needs to be assured. HVAC system (Heating, Ventilation and Air Conditioning) plays an important role on the reliability of these equipments as it makes sure that ambient conditions in the buildings fit the operating temperature range of the equipments. Consequently sizing a ventilation system is really important and it needs to be carried out seriously. This paper introduces the methodology to size and optimize a ventilation system for nuclear power plants’ building. This paper also develops the methodology used to size a smoke control system in a nuclear related building. Direct application of this methodology has been realised for a specific building which is the demineralization station of Hinkley Point C project.
Gumede, Nomfusi Leticia. "An investigation on the impact of procurement quality management in a nuclear power station." Thesis, Cape Peninsula University of Technology, 2011. http://hdl.handle.net/20.500.11838/2221.
Full textThis research project in Procurement Quality Engineering was conducted at a Nuclear Power Generation Company in the Western Cape, South Africa. During the past decade, quality management has become increasingly recognised as highly desirable for all organisations at all levels. All organisations, to varied degrees, can benefit from the application of quality management skills in some parts of their daily operations. The research project will investigate the impact or effect of late deliveries of spares on the operational cost of the organisation. The organisation is not aware what impact the delivery of spares has on operating costs. Against the above background, the problem to be researched within the ambit of this dissertation reads as follows: "Poor product and / or service delivery from Vendors and / or Suppliers have an adverse impact on the output of the Procurement Quality Department" .The primary research objectives of this study are the following: ~ To emphasise the importance of quality within the supply chain. ~ To investigate the impact of non-conforming items delivered to a Nuclear Power Plant. ~ To determine measures which can be put in place to improve communication between suppliers, vendors, buyers and procurement quality engineering. ~ To determine or investigate the cost of poor quality in the organisation. ~ To improve the quality of goods and services through the application of a quality management system within the supply chain. The research method used in this research project involved both qualitative and quantitative research processes. Questionnaires and statistical techniques were used to analyse the data, and to draw conclusions and recommend possible areas for improvement. The research methodology falls within the ambit of a case study.
Watt, Nicholas Robin. "Assessing the potential of phytoextraction to remediate land contaminated with 137Cs at nuclear power station sites." Thesis, University of the West of England, Bristol, 2004. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.409444.
Full textHuggett, Jenny A. "The effect of chlorine, heat and physical stress on entrained plankton at Koeberg Nuclear Power Station." Master's thesis, University of Cape Town, 1988. http://hdl.handle.net/11427/17079.
Full textThe large volume of seawater used for cooling at Koeberg Nuclear Power Station contains many planktonic organisms which are exposed to heat, chlorine and physical stress during their passage through the system. Phytoplankton biomass, measured as chlorophyll a, was reduced by an average of 55.32% due to entrainment, and productivity was decreased by 38.30% on average, mainly due to chlorination. Zooplankton mortality averaged 22.34% for all species and 30.52% for copepods, the dominant group. The copepod Paracartia africana was used in laboratory experiments designed to simulate entrainment. Latent mortality was monitored up to 60 hours after a 30-minute application of stress factors (physical stress was not simulated), and approximately 75% of the total mortality occurred within the 30-minute period. Male Paracartia experienced higher mortalities than females. Extrapolation of these results predicts an overall entrainment mortality (including latent mortality) of 40% for copepods and 29.04% for total zooplankton, although the latter cannot be substantiated. Plankton entrainment at Koeberg was not considered to be overly detrimental to the marine environment because of the very localised area affected, rapid dispersion of heat and chlorine, rapid regeneration times of phytoplankton and some zooplankton, low abundance of commercially important species and potential recruitment from the surrounding productive Benguela upwelling region.
Venables, Daniel. "Risk, trust and place : a mixed methods investigation into community perceptions of a nearby nuclear power station." Thesis, Cardiff University, 2011. http://orca.cf.ac.uk/8523/.
Full text梁榮武 and Wing-mo Leung. "Age dependency of the radiological impact of the daya bay nuclear power station on the local population." Thesis, The University of Hong Kong (Pokfulam, Hong Kong), 1994. http://hub.hku.hk/bib/B31211641.
Full textKliman, Douglas Hartley 1963. "Detection of phenological change in cultivated and uncultivated vegetation with multispectral video." Thesis, The University of Arizona, 1987. http://hdl.handle.net/10150/276600.
Full textAl-Sumait, Jamal. "Solving dynamic economic dispatch problems using pattern search based methods with particular focus on the West Doha Power Station in Kuwait." Thesis, University of Southampton, 2010. https://eprints.soton.ac.uk/165503/.
Full textSmith, Richard Angus. "Measuring quality management system performance using quantitative analyses." Thesis, Cape Peninsula University of Technology, 2013. http://hdl.handle.net/20.500.11838/1234.
Full textMany top performing businesses, which achieve superior levels of success and sustainability, have a sound, implemented, and well maintained, Quality Management System (QMS). The correlation between business success and an implemented management system has been shown in numerous papers. This research, which culminates in a quantitative measure of QMS performance, was conducted at Eskom’s Koeberg Nuclear Power Station (KNPS). The power station is the operating leg of the Koeberg Operating Unit (KOU). The researcher is a QMS lead auditor in the KNPS Quality Assurance Department. A program of audits is planned based on the KOU quality and safety manual and the national regulatory licencing requirements. The audit monitoring program is then implemented over a three year period and considers all the management system processes which impact on nuclear safety and business performance. The individual audits each consider ISO 9001 criteria in context of the business area audited. Each major business area (e.g. design, maintenance, etc.) within the power station adheres to all generic ISO 9001 QMS clauses and considerations, such as documentation management, records management, etc. Each process or business area audit is thus effectively a QMS audit. The audit results, when combined are therefore a representative measure of the overall organisational QMS performance. The potential value to be gained from the audit results and data accrued over the monitoring period has not been optimised to maximise the return on investment to Eskom. The research problem statement thus proposes that the performance measurement capability of the quality management system at Eskom's Koeberg Power Station is insufficient. This diminishes management's ability to identify business risk resulting from management system deficiencies, which impacts negatively on business performance. The research question seeks to determine how the performance measurement capability of the QMS can be improved to assist management in identifying business risk resulting from quality management system deficiencies in order to improve business performance. The research objectives are supported by the literature study, which identifies the quality management methods currently used in order to measure and subsequently improve business performance. It also shows how QMS performance measurement, when deconstructed and analysed can provide the required insight for supporting management decision making. The research approach is considered inductive in that a theory is developed based on the collection and the analysis of that data. Applied research, will thus serve as the basis of the research methodology as it is considered the most appropriate research approach, based on the need to answer practical questions around the measurement of QMS performance philosophy. The research shows that by introducing additional theming and severity data into the secondary audit findings data, it is possible over time to extract high level strategic direction information when analysing the additional metadata. The dimensions and value of the QMS Performance measuring instrument are: Ø A cause and effect theming philosophy of audit findings providing an additional context to business improvement advice to management. Ø The provision of a QMS process deficiency locator / identifier which targets management action areas for improvement. Ø The provision of a quantitative measure of the management system performance, providing a reference from which to improve. By providing a quantifiable measure of an organisations QMS performance, a reference point is provided to gauge QMS performance and also render a definitive measure to enable performance improvement of the business.
Sobotková, Monika. "Facelift EDU." Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414287.
Full textKratochvíl, Zdeněk. "Obnova hermetických potrubních průchodek." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2017. http://www.nusl.cz/ntk/nusl-318139.
Full textŠula, Vladimír. "Zajištění datové komunikace digitálních ochran a terminálů do monitorovacího systému jaderné elektrárny Dukovany." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2015. http://www.nusl.cz/ntk/nusl-221205.
Full textTomoryová, Bianka. "Facelift EDU." Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414291.
Full textKissler, Martin. "Modernizace Jaderné elektrárny Dukovany." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2015. http://www.nusl.cz/ntk/nusl-231807.
Full textRygl, Filip. "Výroba utahováku matice oběžného kola čerpadla." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2020. http://www.nusl.cz/ntk/nusl-417124.
Full textŽák, Tomáš. "Návrh schématu zajištěného napájení jaderného bloku pro řešení projektových i nadprojektových havárií." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2013. http://www.nusl.cz/ntk/nusl-220178.
Full textRokotianskaia, Kseniia. "Facelift EDU." Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414282.
Full textZhakupbekova, Rakhil. "Facelift EDU." Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414302.
Full textVacek, Tomáš. "Posouzení možnosti připojení kogenerační výrobny 138 MW v Prostějově." Master's thesis, Vysoké učení technické v Brně. Fakulta elektrotechniky a komunikačních technologií, 2011. http://www.nusl.cz/ntk/nusl-219083.
Full textRůžičková, Tereza. "Facelift EDU." Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414284.
Full textLi, Liang-Ying, and 李亮瑩. "Transient analysis of Lungmen Nuclear Power Station using RELAP5-RT." Thesis, 2009. http://ndltd.ncl.edu.tw/handle/87671786912834374015.
Full text國立清華大學
工程與系統科學系
97
The purpose of this thesis is using RELAP5-RT, a thermohydraulic system analysis program developed by INEL, to built an independent thermohydraulic analysis model for the simulation of power test transients of Lungmen Nuclear Power Station of Taiwan Power Company. The plant employs the Advanced Boiling Water Reactor (ABWR) designed by General Electric. The focuses of this research are the building of the control logics of the recirculation flow control system (RFCS), reactor protection system (RPS), and rod control & information system (RCIS). The control logics of the other two major control systems, feedwater control system (FWCS), and steam bypass and pressure control system (SBPC) are developed in a separated thesis work and not included in this report. The two separate RELAP5-RT thermohydraulic systems input decks, which model the reactor coolant system and balance of the plant of Lungmen Nuclear Power Plant, are combined into an integrated input deck. These input decks are parts of the Advanced Lungmen Plant Simulator (ALPS) developed by the Nuclear Power Plant Dynamic Simulation and Analysis Lab. of National Tsing Hua University. Then, the control logics of the Lungmen Nuclear Power Station’s control system are incorporated into theintergrated deck. The control logics are also adopted from ALPS. The input deck developed is used to simulate two power test transients of the plant, “trip of one reactor internal pump” and ”three reactor internal pumps trip”. The results are compared with the results of the GE’s STAR and the ALPS’s simulation. The comparisons show that the RELAP5-RT input deck of Lungmen Nuclear Power Station built in the present study can mimic.
Chen, Yu-Chen, and 陳宥辰. "Transient Analysis of Lungmen Nuclear Power Station using RELAP5-RT." Thesis, 2010. http://ndltd.ncl.edu.tw/handle/29564820682766731914.
Full text國立清華大學
工程與系統科學系
98
In the present study, an independent RELAP5-RT input deck for the LungMen Nuclear Power Station of Taiwan Power Company is developed. LungMen nuclear power station employs the Advanced Boiling Water Reactor (ABWR) designed by General Electric. RELAP5-RT is a thermohydraulic system analysis program developed by INEL. The work involved in the study includes: 1. combine the RELAP5-RT thermal hydraulic input decks of reactor vessel and balance of plant into an integrated deck. These input decks are parts of the Advanced Lungmen Plant Simulator (ALPS) developed by the Nuclear Power Plant Dynamic Simulation and Analysis Lab of National Tsing Hua University. 2. Implement the control logic of feedwater control system (FWCS), and steam bypass and pressure control system (SBPC) into the merged deck. These control logics are adopted from ALPS. Together with the control logics of recirculation flow control system (RFCS), reactor protection system (RPS), rod control and information system (RCIS), which have been developed in previous work, a transient analyses tool of LungMen NPS has been completed. The deck developed is used to simulate two power test transients of the plant-“one feedwater pump trip” and “Load Rejection” . The results are compared with the results of the GE’s STAR and the ALPS’s simulation. The comparisons show that the RELAP5-RT input deck of Lungmen Nuclear Power Station developed in the present study functions properly.
Cheng, Hsin, and 鄭欣. "Building MELCOR Input Deck of Chinshan Nuclear Power Station and Analyses of Station Blackout Sequence." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/61475608109741336931.
Full text國立清華大學
核子工程與科學研究所
102
In the present study, a MELCOR input deck for the Chinshan Nuclear Power Station of Taiwan Power Company is developed. Chinshan nuclear power station employs a Boiling Water Reactor (BWR IV) designed by General Electric and Mark I containment. The input deck is used to analyze the station blackout sequence, and the results will be compared with the MAAP5. The work involved in the study includes: (1) Use the MELCOR input deck from INER as the basis. Build a new MELCOR input deck of Chinshan nuclear power station according to the MAAP5 input deck and the corresponding calculation sheets from INER. (2) Initialize the new MELCOR input deck to staeday state. (3) Simulate the SBO event of the plant using MELCOR and MAAP5 codes with the assumption that the core melt occurs under high pressure and low pressure. (4) Compare the results of these two codes. The major focus are the timing of major events, the thermal hydraulic responses of reactor coolant system and containment, hydrogen generation, the radionuclide releases from core during the core melt and during the molten core concrete interactions, and the fraction of radionuclide releasing to the environment. Compared the results, it has been found that: (1) MELCOR has a more detailed modeling of core and vessel internal regions. It consists of 3 radial rings and 13 axial levels. MAAP5 treats the core as a single volume. (2) The reactor vessel bottom attack model amd mode of its failure of these two codes are also significantly different. (3) The amount of hydrogen generation during the core melt as predicted by these two codes are significantly different. The impacts of flow blockage on the prediction of hydrogen generation of these two codes are different. MAAP5 is more sensitive to the assumpation of flow blockage. (4) The classification of radionuclide groups is different. Due to the difference in the modeling of core region, the predicted in-vessel releases of radionuclide is different. The predicted ex-vessel releases are also significantly different due to difference in the modeling of core concrete interactions. The fraction of each radionuclide released to the environment is different. (5) The MELCOR results are very sensitive to the time step size. If the time step size has not been set properly, the code stops calculation prematurely.
Lai, Yu-Cheng, and 賴宥丞. "Building MELCOR Input Deck of Kuosheng Nuclear Power Station and Analyses of Station Blackout Sequence." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/62257925226519460538.
Full text國立清華大學
核子工程與科學研究所
102
In this study, the MELCOR input deck of Kuosheng Nuclear Power Plant is developed based the the plant data as specified in the MAAP5 input deck and calculation sheets of the plant, which are provided by Institute of Nuclear Energy Research. The plant is deployed with two Boiling Water Reactors (BWR VI) designed by General Electric and enclosed in Mark III containment. A high pressure station blackout (SBO) sequence of the plant is simulated using MELCOR and MAAP5. The results of the simulations are compared to assess the differences of these two codes. The comparisions are concentrated on the timing of major events, thermal hydraulic response of reactor coolant system and containment, debris relocations from one region to another, hydrogen production, in-vessel and ex- vessel release and environmental releases of radionuclides. The differences of the simulation results are very significant due to the differences in the severe accident phenomenological models adopted by these two codes. The amount of hydrogen generation within the reactor pressure vessel as predicted by MELCOR is 921 kg and that as predicted by MAAP5 code is 76 kg. Nevertheless, the amount of hydrogen production during molten core concrete interaction as predicted by MELCOR and MAAP5 code is 1,382 kg and 2,016 kg, respectively. The extent of in-vessel and ex-vessel releases of radionuclides as predicted by these two codes is also very different. In environment release, there are several fission products that MELCOR is bigger than MAAP5, including Cs, I, Te, Ru, Mo, Nb, U, Sn; And other fission products such as Xe, Ba, Zr, La, Ce, Cd, MAAP5 is larger than MELCOR. In the study, sensitivity study is performed to assess the impact of depressurization on the failure mode of reactor vessel bottom head. In a high pressure SBO sequence, the vessel failure is caused by the stress and strain produced in a high pressure environment. In a low pressure SBO sequence, the failure of vessel is caused by the melting of instrumentation tubes.
Chang, Ho-Yu, and 張賀嵎. "Building MELCOR Input Deck of Maanshan Nuclear Power Station and Analyses of Station Blackout Sequence." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/56254829322083637477.
Full textNordt, Kevin M. "MAAP/MELCOR comparison station blackout at the point beach nuclear power plant /." 1992. http://catalog.hathitrust.org/api/volumes/oclc/26109393.html.
Full textTypescript. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaves 93-94).
Huang, Meng Ting, and 黃孟婷. "Loss of Cooling Accident Simulation of Chinshan Nuclear Power Station Spent-fuel Pool." Thesis, 2015. http://ndltd.ncl.edu.tw/handle/equn6n.
Full text國立清華大學
核子工程與科學研究所
103
Spent fuel pool works as a temporary storage for fuel discharged from core, and relys on Spent Fuel Cooling System (SFPCS) to remove decay heat. When a loss of cooling event happens, the decay power of fuel can’t be removed from pool. The water level drops due to evaporation, and leads to fuel uncovery. After fuel is uncovered, the cladding temperature elevates due to deterioration of heat transfer. The oxidation of Zircaloy by the steam generated hydrogen and heat. This work aims to analyze a loss of cooling event of spent fuel pool of Chinshan Nuclear Power Station. In the present study, RELAP/MOD3 and MAAP5.02 are used to simulate the event. Chinshan Nuclear Power Station is operated by Taiwan Power Company, which employs BWR IV reactor and Mark I containment. The spent fuel pool of Chinshan Nuclear Power Station is divided into 14 storage region, and the hottest region is J region. This study uses ASB 9-2 formula to calculate decay power of spent fuels. The radiation heat transfer model and partial length fuel rods are built. The results of J region RELAP simulation indicate that spent fuel is uncovered at 6.75 days after event takes place. The spent fuel is uncoverd at 19.33 days in the whole pool simulation of RELAP5 simulation. The results of former simulation is too conservative. The results simulated by MAAP are closed to RELAP5’s results. It takes 19.25 days for fuel to uncover in MAAP simulation. Moreover, the fuel uncovers at 17.78 days after event happens by simple energy balance calculation. As predicted by RELAP5 core, the cladding temperature reaches 2200℉ at 22.92 days after event occurs. However, the corresponding time is 33.56 days in the MAAP5 simulation. Due to inconsistency in MAAP5 numerical calculation after fuel uncovery, the hydrogen generation rate doesn’t predict correctly. Therefore, cladding temperature after fuel uncover is not correct.
王靖雅. "Software Reliability Assessment of the Reactor Protection System for Lungmen Nuclear Power Station." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/73962376270000165021.
Full textLiang, Ching-Chun, and 梁景俊. "A Study on Severe Accident Sequence Analyses for Chin-Shan Nuclear Power Station." Thesis, 2002. http://ndltd.ncl.edu.tw/handle/42545787535599310968.
Full text中原大學
機械工程研究所
90
The purpose of this study is to evaluate the postulated severe accident scenarios - such as station blackout, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS) - for the Chin-Shan Nuclear Power Station using the Modular Accident Analysis Program (MAAP) version 4.0.4. For these accident scenarios, the behaviors of reactor core and containment, and the release of fission products were analyzed. In addition, the phenomena associated these scenarios were discussed. The station blackout scenario assumed that the plant lost all its on-site and off-site power, leading to loss of all coolant injection capabilities, except the reactor core isolation cooling (RCIC) system that is driven by the steam provided by the reactor. For the LOCA scenarios, all coolant injection systems were assumed to be lost and the break location was assumed to be at the piping connecting recirculation pump to the reactor vessel, with the break sizes of 0.1, 0.3, 0.5, 0.7, 1.0, and 2.1795 (double-ended, guillotine-type break) ft2. For the ATWS scenario, the reactor scram was assumed to be not available, due to the failures of automatic and manual control rod insertion as well as the stand-by liquid control system. In this scenario, the reactor core became degraded rapidly due to the elevated core power generated. For these types of scenarios, actions taken by the operators were analyzed to determine their impacts on the progression of the accidents. Without adequate core cooling and/or containment heat removal, the reactor core heated up, melted, and then relocated to the vessel bottom head. In the meantime, substantial amount of hydrogen resulting from the metal-water action in the core region was generated. Due to the decay heat associated with the core debris (or so-called corium), the molten corium continually heated up and melted through the bottom of the vessel. The molten corium that located at the lower drywell again heated up, interacted with the concrete, and generated additional non-condensable gases. The gases pressurized the wetwell gas space, leading to venting of the containment through the hard-pipe vent. Following containment venting, the fission products were released to the environment. Results of this study indicated that the progressions of the accident scenarios were affected by the availability of the coolant injection systems and the containment heat removal systems, and the reactions taken by the operators. In addition, the models implemented in the MAAP 4.0.4 compared to those of the MAAP 3B had significant effects on the timing of the failure of the core plate and the melt-through of the vessel bottom head. Furthermore, the values used in the decontamination factor had a major impact on the amount of the release of the fission products following containment venting.
劉璧銘. "The Thermohydraulic Analysis of the Containment System of Lungmen Nuclear Power Station Under LBLOCAs." Thesis, 2002. http://ndltd.ncl.edu.tw/handle/37073055147622412841.
Full text國立清華大學
工程與系統科學系
90
This study is aimed directly to the containment system of the Lungmen nuclear power station, which is an advanced boiling water reactor, constructed and operated by the Taipower company. A two-phase, multi-component, air-water-steam flowing model simulating different dynamic phenomena of containment system, during design basis accidents, is established. Importances and characteristics of this study are to establish an independent fluid thermohydraulic computer code, which is based on the fundamental theories, for the containment system of the Lungmen nuclear power station. It can analyze the thermodynamic properties and parameters of fluids in the containment during accidents or transients. Then we can provide the time-varying system response data as the boundary conditions for the detail analysis of hydrodynamic loading inside wetwell. Based on fluid dynamics and heat transfer processes, the model can be divided into several submodels, which includes drywell, water clearing, air clearing and wetwell. The mass and energy conservation of thermodynamics can be used to analyze the drywell and wetwell submodels. By adding the momentum equation to water clearing and air clearing submodels, the clearing velocity and parameters of flowing fluids can be calculated. In this investigation, two LOCAs of desian basis accidents are the targets to be analyzed and compared. One is the Feedwater line LBLOCA, and the other is the Main-Steam line LBLOCA. By analyzing two LBLOCAs, we can understand different effects created by these two DBAs. The thermodynamic porperties in the drywell or in the wetwell can compare with PSAR results of the Lungmen nuclear power station, to assure that the established computer program is correct and precious. Then the data of fluids flowing in the vents can act as the boundary conditions, which includes velocity of water clearing and mass of the fluid flowing, to enable the calculations of hydrodynamic loading on the structures submerged or above the suppression pool.
Huang, Li-Hua, and 黃立華. "A Study on the Kuosheng Nuclear Power Plant under the Station Blackout Accident Conditions." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/68882100057665084255.
Full text中原大學
機械工程研究所
102
Nuclear-related units have paid extraordinary attention to nuclear energy simulation. They have conducted a very large number of experiments and also developed several sets of nuclear power plant simulation software. The software used here are Modular Accident Analysis Program (MAAP5), developed by the Fauske &; Associates, Inc., a simulation program for analyzing nuclear power plant accidents, and MELCOR1.8.5, developed by the U.S. Department of Energy’s Sandia National Laboratory. These two programs are applied to study the cases of severe accidents of Kuosheng Nuclear Power Plant. This thesis analyzes the case for the Station Blackout, the time difference between the added Emergency Operation Procedures and the sustained Emergency Operation Procedures, and their impacts on the rescue operation logic for the plant and the time delay of firefighting water pouring into nuclear reactor cores. By comparing the analyses of different programs, the paper explores the serious accidents and verifies the accuracy of the simulation programs, which facilitates a quicker and more appropriate operation when emergencies occur, elevating the safety of the overall nuclear power plant. The analysis results show that in the case of the Station Blackout, MAAP5 appears to be conservative in calculating the amount of steam generated, which ends up with the smaller numbers of pressure and temperature peaks. On the other hand, MELCOR1.8.5 is conservative in calculating the amount of hydrogen generated. Generally speaking, these two models have similar simulation tendencies.
Tsai, Tseng-I., and 蔡正益. "An Evaluation on the Properment of Plant Modification Working Process in Nuclear Power Station." Thesis, 1994. http://ndltd.ncl.edu.tw/handle/61957685303806216485.
Full textFu, Yu-Feng, and 傅宇烽. "Building the Control Systems of RELAP5-3D Input Deck of KuoSheng Nuclear Power Station." Thesis, 2016. http://ndltd.ncl.edu.tw/handle/79221827744417434527.
Full text國立清華大學
核子工程與科學研究所
104
Abstract In the present study, the control system of RELAP5-3D input deck of Kuosheng Nuclear Power station is constructed. The plant employs a BWR6 (Boiling Water Reactor) reactor with Mark III containment and is operated by Taiwan Power Company. The rated power of the system is 2,894 MWth. The RELAP5-3D input deck is obtained from Institute of Nuclear Energy Research. The control system is formulated based on the control system embedded in the plant engineering simulator, which is developed on the 3 Key Master platforms. In the present study, the deck is initialized to a steady state condition with constant power. The feedwater control, pressure control, and recirculation flow control are incorporated sequentially. The three- element feewdwater control scheme is adopted. The elements are“narrow range water level”, “steam mass flow rate”, and “feed water mass flow rate”. After the deck is initialized to steady state with the feedwater control system, the Pressure Control is added to govern the valve area of Turbine Control Valves. The Recirculation Flow Control is incorporated to control the mass flow rate of Recirculation Flow. Finally, the Point Kinetic is adopted for the transient simulation. The results of the power tests "100% power load rejection test", "96% power main steam isolation valve closure test", and "68% power recirculation pump trip test," are used to test the validity of the control system of the input deck.
Trollope, Ian Douglas. "Derivation of Operational Intervention Levels for the early phase of radioactive material at Koeberg Nuclear Power Station." Thesis, 2015. http://hdl.handle.net/10539/16810.
Full textAn investigation was performed to look at a method to develop easy to use field survey measurements to assist decision makers in the process of deriving public protective actions. This method could be used at a nuclear power plant if certain accident conditions are known. International values for operational intervention levels (OIL’s) do exist and are recommended to be employed if station specific data has not been derived. No values exist specific to Koeberg Nuclear Power Station and as a result, this became an ideal opportunity to derive station specific values. It was firstly necessary to decide on a specific accident type and hence an applicable accident release fraction. A suitable accident software dispersion code was applied to calculate the organ doses for the selected accident type. It was also decided to use two different wind dispersion criteria to further refine the results. Due to the complexities of dose distribution within the body it was also necessary to look at the gamma dose in isolation as this would be the measurement radiation type utilised as a limit in the field either using installed radiation monitors or by physical measurement performed by station Radiation Protection staff. Comparisons were done with thyroid and lung dose versus gamma dose to arrive at ratios for this specific accident type. This would then be indicative of the total dose to each organ as a result of a single field measurement. Conclusions were drawn on the results obtained and recommendations were made for when this type of data may be suitable for use in the unlikely event of a nuclear accident.
Hsu, Keng-Hsien, and 許耕獻. "Transient Analysis of Advanced Boiling Water Reactor of Lungmen Nuclear Power Station using RELAP5-RT." Thesis, 2009. http://ndltd.ncl.edu.tw/handle/73439482143190315165.
Full text劉紹楷. "Transient Analyses of Advanced Boiling Water Reactor of Lungmen Nuclear Power Station using RELAP5/MOD3." Thesis, 2008. http://ndltd.ncl.edu.tw/handle/38498223830789580516.
Full text國立清華大學
工程與系統科學系
96
In this study, a reactor system thermohydraulic system analysis code, RELAP5/MOD3 is used to analyzed selected transients in the Final safety analysis report (FSAR) of Lungmen Nuclear Power Stations (NPS).The plant employs two general electric designed Advanced Boiling Water Reactor (ABWR) with rated power of 1350MWe. The Lungmen input deck of RELAP5/MOD3, models reactor pressure vessel (RPV) and banlance of plant (BOP), which includes major components such as turbines, heat ecxhangers , reheaters, and pumps. The input deck has been successfully initialize to a steady state condition. The internal pump trip and main steam line isolation valves closures transients in FSAR of Lungmen NPS are simulated. The results are compared with the data in FSAR. The comparsion of the simulated results with the results shown in FSAR are not satisfactory due to lack of modeling of the control system in the input deck.
Wu, Wei-lih, and 吳偉立. "Application of HHT to temperature variations at the thermal outlet of Third Nuclear Power Station." Thesis, 2005. http://ndltd.ncl.edu.tw/handle/87378891379691686032.
Full text國立中山大學
海洋物理研究所
93
Nan Wan is a half-closed embayment in the most southern part of Taiwan. While facing the Luzon Strait, it also connects to the Pacific Ocean in its southeast, and is adjacent the Taiwan Strait and the South China Sea . In view of general oceanic circulation, Nan Wan Bay happens to lie to the rim of South China Sea circumfluence and Kuroshio where a variety of water mass exchange has taken place, causing saline intrusion and mixed of water. Seasonal variation and tidal fluctuations also contribute to the exchange of water masses. The Third Nuclear Power Station of Taiwan Power Company is located in Nan Wan with its thermal discharge outlet adjacent to Maobitou to the west of the bay in order to minimize the effect of warm water discharge on the local marine ecology and coral . A long-term monitoring program on water temperature and other environmental factors has been set up implemented .this research report will first describe the archives regarding the hydrology in Nan Wan in support of monitoring the process in temperature variation . Previous research efforts are found somehow unable reveal precisely the physical mechanism leading to water temperature variations in the bay, due to limited facilities, short of information or poor analytical tools. This report adopts 14 records of water temperature at the thermal outlet of the Third Nuclear Power Station for signal analysis. As to non-linear and unstable data analysis, it is based on the Hilbert-Huang Transform. HHT includes Empirical Mode Decomposition, EMD which could decompose the raw data into numerous Intrinsic Mode Function, IMF. It is allowed to comprehend the main causes for the rising and dropping of water temperature based on the variation of spectroscopy by transferring through Hilbert and analyzing via IMF. Furthermore, the characteristic of each quantity could be developed according to the quantities acquired from the former method of HHT. The analytical report of water temperature covers 14 records dating from 1999 to 2003. In light of the analytical report, tide and wind account for the main cause of the temperature variation in waters while demanding information to ensure whether it is influenced by other factors like internal waves, water masses or landforms, etc. In addition, the report compares the difference in the same of data between FFT and HHT and moreover concludes the advantages and disadvantages as reference for researches.
Lin, Tser-Tung, and 林則棟. "An Analysis on the Indicator Meaning of Scram Frequency Statistics for Taipower Nuclear Power Station." Thesis, 1994. http://ndltd.ncl.edu.tw/handle/55370150926124320103.
Full textHU, ZHONG-QING, and 胡中清. "A study on the improvement of effectiveness of shift work group in nuclear power station." Thesis, 1991. http://ndltd.ncl.edu.tw/handle/08263813150688975493.
Full textYang, Shao-yu, and 楊韶彧. "The agenda building process of news source :the debate of the fourth nuclear power station." Thesis, 1993. http://ndltd.ncl.edu.tw/handle/28085525679751598255.
Full textYang, Jian-Hong, and 楊健鴻. "A Study on the Radiation Doses of Station Blackout Accident Scenariofor the Nuclear Power Plant." Thesis, 2012. http://ndltd.ncl.edu.tw/handle/06864006195934268565.
Full text中原大學
機械工程研究所
100
Abstract After Three Mile Island (TMI) accident, nuclear industry started to promote research on severe accidents. MAAP (Modular Accident Analysis Program) code is a severe accident analysis computer program developed by Fauske & Associates (FAI), sponsored by Electric Power Research Institute (EPRI), and is widely used in the nuclear industry. Now, MAAP has been advanced to MAAP5. In addition to including all the functions of MAAP 4.0.4, MAAP5 has a new function of dose calculation (MAAP5-DOSE). It includes on-site and off-site dose calculation. The Fukushima accident on March 11, 2011 was caused by earthquake and the subsequent tsunami, resulting in a station blackout (SBO) to the plant. The operators couldn’t inject water into core in time and caused core melt. Then, fission products were released into environment. The fission products covered substantial portion of the world by the ensuing air currents and ocean currents. The safety of nuclear power plants becomes the focus again in the world. The purpose of this study was to simulate the station blackout accident scenario for the Lungmen Nuclear Power Plant (NPP). As a result of loss of all cooling to the core, the core melted, reactor pressure vessel (RPV) melted-through, and containment overpressure protection system (COPS) activated. The release of fission products was then calculated by the MAAP5 code. Sensitivity analyses on RPV melt-through caused by delays of fire water injection were also studied. The results obtained from this study would be useful for the plant operators in evaluating the accident conditions and planning for the response procedures. With simulated results obtained from dose analysis, the highest dose at low population zone (300 m) was determined to be 1.56 Sv at the end of calculation (70- hr). The direct effect on the human body was nausea. From atmospheric instability analysis, factors such as wind speed, wind direction, and air current were connected with the concentration of radioactive doses and the magnitude of offsite doses. The highest integrated dose was calculated to be 11.08 Sv. If a human body receives such acute exposure (11.08 Sv), the probability of death is above 90%. The results obtained from severe accident analyses clearly showed that the various mitigation measures would help prevent the melt-through of RPV and the subsequent activation of the containment overpressure protection system, resulting in the offsite doses being released. The Fukushima accident was the turning point to the added concerns regarding the safety of nuclear power plants. It helps us understand natural disaster and its resulting severe nuclear accidents. To enforce current regulation will give a great contribution to the nuclear safety.
林冠佑. "Loss of Cooling Accident Simulation of Chinshan Nuclear Power Station Spent-fuel Pool Using RELAP5." Thesis, 2012. http://ndltd.ncl.edu.tw/handle/18663723662513539502.
Full text國立清華大學
核子工程與科學研究所
100
In the present study, a RELAP5/Mod3 input deck for the spent fuel pool of the Chinshan Nuclear Power Station of Taiwan Power Company is developed. Chinshan nuclear power station employs a Boiling Water Reactor (BWR IV) designed by General Electric and Mark I containment. The input deck is used to analyze the loss of cooling event of spent fuel pool. The work involved in the study includes:(1)Use the ASB-92 formula to calculate decay power of the spent fuels. The spent fules of the latest discharged cycle are calculated in detail. The decay power generated in these spent fuel depends on the final power of the spent fuels during operation. (2) The lumped parameter approach is adopted to model the spent fuel racks within the pool. The rack which contains maxmum numbers of the latest cycle of spent fuels is model in detail. (3) The radiation heat transfer model is built. (4) The impact of counter current flow limit (CCFL) and radation heat transfer model is assessed. (5) Sensitivity studies of the cooling effect of water spray on the heat up of spent fules are performed. The results indicate that spent fuel is uncovered at 6.75 days after accident takes place and the cladding temperature rises above 2200℉at 8.1 days after accident takes place 8.1 days. The time is about 13.2 hours less that the results predicted using simple energy balance method. The results also show that the impact of CCFL and radiation heat transfer model is marginal. The results indicate that the earlier take spray action, the shorter time to recover the spent fuel will take.If the capacity of spray is bigger, the cladding temperature will be decreased more effectively.