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1
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.
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At Koeberg Nuclear Power Station, the reactor thermal power limit is one of the most important quantities specified in the operating licence, which is issued to Eskom by the National Nuclear Regulator (NNR). The reactor thermal power is measured using different methodologies, with the most important being the Secondary Heat Balance (SHB) test which has been programmed within the central Koeberg computer and data processing system (KIT). Improved accuracy in the SHB will result in a more accurate representation of the thermal power generated in the core. The input variables have a significant role to play in determining the accuracy of the measured power. The main aim of this thesis is to evaluate the sensitivity of the SHB to the changes in all input variables that are important in the determination of the reactor power. The guidance provided by the Electric Power Research institute (EPRI) is used to determine the sensitivity. To aid with the analysis, the SHB test was duplicated using alternate software. Microsoft Excel VBA and Python were used. This allowed the inputs to be altered so that the sensitivity can be determined. The new inputs included the uncertainties and errors of the instrumentation and measurement systems. The results of these alternate programmes were compared with the official SHB programme. At any power station, thermal efficiency is essential to ensure that the power station can deliver the maximum output power while operating as efficiently as possible. Electricity utilities assign performance criteria to all their stations. At Koeberg, the thermal performance programme is developed to optimize the plant steam cycle performance and focusses on the turbine system. This thesis evaluates the thermal performance programme and turbine performance. The Primary Heat Balance (PHB) test also measures reactor power but uses instrumentation within the reactor core. Due to its location inside the reactor coolant system, the instrumentation used to calculate the PHB is subject to large temperature fluctuations and therefore has an impact on its reliability. To quantify the effects of these fluctuations, the sensitivity of the PHB was determined. The same principle, which was used for the SHB sensitivity analysis, was applied to the PHB. The impact of each instrument on the PHB test result was analysed using MS Excel. The use of the software could be useful in troubleshooting defects in the instrumentation. A sample of previously authorised tests and associated data were used in this thesis. The data for these tests are available from the Koeberg central computer and data processing system.
2
Dinoko, 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.
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Nuclear 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.
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程明錦 y 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.
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Rylands, 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.
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The induction motor is a highly utilised electrical machine in industry, with the nuclear industry being no exception. A typical nuclear power station usually contains more than 1000 motors, where they are used in safety and non-safety application. The efficient and fault-free operation of this machine is critical to the safe and economical operation of any plant, including nuclear power stations. A comprehensive literature review was conducted that covered the functioning of the induction machine, its common faults and methods of detecting these faults. The Condition Based Maintenance framework was introduced in which condition monitoring of induction machines is an essential component. The main condition monitoring methods were explained with the main focus being on Motor Current Signature Analysis (MCSA) and the various methods associated with it. Three analysis methods were selected for further study, namely, Current Signature Analysis, Instantaneous Power Signature Analysis (IPSA) and Motor Square Current Signature Analysis (MSCSA). Essentially, the methodology used in this dissertation was to study the three common motor faults (bearings, stator and rotor cage) in isolation and compare the results to that of the healthy motor of the same type. The test loads as well as fault severity were varied where possible to investigate its effect on the fault detection scheme. The data was processed using an FFT based algorithm programed in MATLAB. The results of the study of the three spectral analysis techniques showed that no single technique is able to detect motor faults under all tested circumstances. The MCSA technique proved the most capable of the three techniques as it was able to detect faults under most conditions, but generally suffered poor results in inverter driven motor applications. The IPSA and MSCSA techniques performed selectively when compared to MCSA and were relatively successful when detecting the mechanical faults. The fact that the former techniques produce results at unique points in the spectrum would suggest that they are more suitable for verifying results. As part of a comprehensive condition monitoring scheme, as required by a large population of the motors on a nuclear power station, the three techniques presented in this study could readily be incorporated into the Condition Based Maintenance framework where the strengths of each could be exploited.
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Ching, 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.
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Simons, 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.
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Thesis (MTech (Quality))--Cape Peninsula University of Technology, 2016.The 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.
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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.
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Bezuidenhout, 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.
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In line with its commitment to safe nuclear power generation, the Koeberg Nuclear
Power Station (KNPS) replaced the outdated vibration monitoring system with a modern
on-line vibration monitoring system. This will allow plant personnel to monitor
components on a continuous basis which will provide faster response time in the
scenario of excessive vibrations of the primary components.
This study focuses on the analysis of the vibration of the primary components of the
KNPS by analysing the frequency spectra of the vibration signals of the primary
components and comparing these to reference signatures obtained during similar
operating conditions. The condition of the vibration sensors will also be evaluated.
In order to obtain a deeper understanding of the vibration behaviour and hence vibration
signatures of the KNPS primary reactor components, a simplified mathematical model
of the primary components is developed, based on the system of elasto-dynamic
equations. The equations are solved numerically and used to simulate the KNPS
vibration monitoring system. The mechanical system is modelled. Time series are
generated and Fast Fourier Transforms (FFT) are calculated to simulate the new KNPS
monitoring system. In the simulation mechanical degradation of the primary
components as well as sensor degradation is simulated.
The purpose of this study is to indicate whether mechanical degradation has occurred in
the primary components of the plant and to validate the vibration signals. At the same
time the study aims to lay a foundation for future monitoring and interpretation of
vibration signatures by simulating the vibration and the monitoring signals.
It was found that the primary components had not been affected by mechanical
degradation as no deviations in resonances were detected in the frequency signatures.
A small number of vibration sensors were found to have deteriorated; hence
replacement / maintenance was proposed.
The mechanical model and the simulation of the monitoring signals proved to be useful
to understand and interpret the vibration of the KNPS primary components.Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.
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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.
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Avstängda kärnkraftverk berövar många människor av elektricitet och det skulle ha en negativ inverkan både på företagets framtoning och mänskliga aktiviteter. På grund av detta behöver tillgängligheten av utrustningen i alla byggnaderna som kärnkraftverken består ses till. HVAC-system (Heating, Ventilation and Air Conditioning) spelar en viktig roll när det gäller tillgänglighet av utrustning eftersom dessa system ser till pålitligheten är på topp genom att anpassade omgivningsförhållanden till utrustningen. Att designa ventilationssystemet rätt är därför mycket viktigt och måste göras noggrant. Denna rapport introducerar metodologin för att designa och optimera ett ventilationssystem för en av byggnaderna i ett kärnkraftverk. Utöver detta utvecklas och beskrivs en metodologi för att designa ett rökkontrollssystem för en byggnad som ingår i kärnkraftverket. Dessa metodologier har implementerats för en byggnad i en demineraliseringsstation, Hinkley Point C project.During 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.
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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.
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Thesis (MTech (Quality))--Cape Peninsula University of Technology, 2011.This 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.
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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.
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The intended endpoint for Magnox Electric plc's reactor decommissioning strategy is site clearance and de-licensing, which may require remediation of any 137CS contaminated land on these sites. Phytoextraction might provide a practical and environmentally acceptable method of reducing radioactive waste volumes for disposal. Field trials conducted at Bradwell Nuclear Power Station, Essex, UK and experiments in controlled conditions, using Beta vulgaris, showed that soil-to-plant transfer factors increase as soil 137CS activity concentration decreases, implying that constant 137CS removal rates are possible during site remediation. It was shown that 133CS might be responsible for this effect. Short time-interval multiple croppings were not found to increase the rate of 137 Cs removal. For successful implementation of 137CS phytoextraction plant species need to be identified that can accumulate higher concentrations of 137 Cs than those identified at present. The large variation in 137CS uptake between species tested here suggests that extensive species screening programmes will be required. Soil amendments appear essential to the phytoextraction of aged 137CS and NH/, 133CS+ and K+ were shown to be capable of extracting up to 25 % of the 137CS from soil at a field site. In a 14 month simulated 137CS phytoextraction trial in controlled conditions, over 60 % of 137CS applied to an organic soil was removed using a 20 mg kg-1 CsCl soil amendment suggesting that, where 137CS is sufficiently plant available, phytoextraction might be a useful soil remediation technology. In a desk based BPEO study, composting was identified as the most appropriate option for conditioning the biomass arising from 137 Cs phytoextraction allowing it to be stored at the UK's LLW repository at Drigg, Cumbria. It was concluded that the risk associated with soil amendments of 137 Cs leaching off sites and the poor plant growth conditions likely to be found at field sites would require the use of a greenhouse with soil contained in lysimeters during phytoextraction.
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Huggett, 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.
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Bibliography: pages 112-138.The 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.
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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/.
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Recent UK government policy advocates the expansion of nuclear power, and indicated that any new nuclear power stations will be built mostly at existing 'nuclear' sites where it is apparently assumed that broad community acceptance will be encountered. This thesis investigated community perceptions of an existing nearby nuclear power station at three locations, through a mixed-methods design incorporating a Q-Method study (n=84) and a household study (n=1,327), and with additional reference to an existing qualitative dataset. The thesis aimed to provide a detailed description of how such communities live with nuclear power. Specifically, it investigated (a) the main community points of view on the nearby nuclear power station; (b) the dimensionality of trust between communities and the power station; (c) the associations between risk perceptions, trust, sense of place, and residential proximity to the power station, and (d) the factors associated with community support for new nuclear build in the nearby area. Four points of view were identified. These were broadly consistent across study locations but also reflected some site-specific concerns. The dimensionality of trust between the nuclear power station and nearby communities was found to comprise separate Affective and Cognitive components. It was concluded, however, that the primary influences, both on public perceptions of the risks associated with the existing nuclear power station, and on community attitudes towards the building of a new one, were related to perceptions of place. This thesis provides a contemporary insight into some of the ways that communities live in close proximity to a nuclear power station. Its theoretical and applied implications are discussed in the context of psychological theory and recent UK energy policy.
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梁榮武 y 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.
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Kliman, 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.
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Multispectral video (MSV) images were used to measure phenological changes in cultivated and uncultivated vegetation communities surrounding the Palo Verde Nuclear Generating Station (PVNGS). Multispectral video imagery was acquired from aircraft on seven dates between the middle of June and the end of September, 1986. Images representing three sites near the PVNGS were selected to calculate Ratio Vegetation Index (RVI) values for seven surface cover types. Mean RVI values were tested sequentially for change, plotted as a function of time, and then compared to a moisture index and the crop calendar. MSV detected changes in cultivated vegetation corresponding to the crop calendar. Changes in natural vegetation and the non-vegetated cover types were also detected, but did not correlate to the moisture index. There is insufficient evidence to determine if detected changes in uncultivated vegetation were the result of phenological changes or electronic noise.
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Al-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/.
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This thesis is concerned with Dynamic Economic Dispatch (DED) problems, in particular in the context of the current and future needs of the electrical power system in the State of Kuwait. General Economic Dispatch (ED) issues are addressed, under both static and dynamic conditions, with valve-point effects accounted for. Improvements have been achieved in terms of lower fuel costs, but also more efficient and reliable simulation algorithms. The existing ED/DED models have been improved in various ways and enhanced by developing and incorporating two renewable energy sources; namely wind energy and solar energy. These two have been identified as most relevant to the power system investigated. The models developed are general and can be adjusted to represent many practical systems. The Economic Dispatch problem had been formulated and solved as a constrained optimisation and a particular technique selected for this purpose – not explored before – was a Pattern Search (PS) algorithm. For illustrative purposes, the proposed PS technique had been applied to various test systems to validate its effectiveness. Furthermore, convergence characteristics and robustness of the method had been assessed through comparison with results reported in literature. The PS technique was found to be very competitive in terms of its overall performance. Variations of the technique have also been explored, in particular a hybrid formulation exploiting Genetic Algorithm (GA), Pattern Search (PS) and Sequential Quadratic Programming, and advantages of such a combined technique reported. A DED model for the West Doha Power Station (WDPS) in Kuwait has been developed and the penetration of renewable energy resources to this model has been discussed. The DED model was then solved using the PS method developed in this thesis to achieve the optimal dispatch with the aim to minimise fuel costs in WDPS. Considerable potential savings in electric power production of WDPS have been identified and thus the benefits of deploying renewable energy in Kuwaiti electric system demonstrated.
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Smith, 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.
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Dissertation submitted in fulfilment of the requirements for the degree
Master of Technology: Quality
in the Faculty of Engineering
at the Cape Peninsula University of Technology, 2013Many 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.
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Sobotková, Monika. "Facelift EDU". Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414287.
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The master thesis deals with the facelift of the forecourt of the Dukovany Nuclear Power Station . This space consists of supporting functions for the main working such as administrative, metrology, stocks, cloakrooms and services. There is also the station for the buses, which carry employees to and from work, the regular bus station and parking lots. The forecourt of the power station is now inconvenient from a functional point of view, because the capacity of the existing buildings and parking lots is insufficient; and also from aesthetic point of view, because the buildings from 70’s don’t look good anymore, the parking consists of the large asphalt areas and there is no representative anteroom in front of the main entry, which should be there in view of the significance of the power station. In the study I deal with these problems by the complex reconstruction of the area, I replace the huge asphalt parking lots with the parking houses, create the administrative zone with the public place in front of the main entry to the plant, extend the capacities for the particular functions and add the new required functions (kindergaten, other services). The result of this conversion is the area with the particular functional zones with the representative forecourt in front of the main entry and with the enough space for each of the functions. This work follows up the last semester urbanist-architectonic study. The diploma thesis focus on the improvements of the weaker parts of the prime study and on the elaboration of the architectural study of the adminstrative zone with services.
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Kratochvíl, Zdeněk. "Obnova hermetických potrubních průchodek". Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2017. http://www.nusl.cz/ntk/nusl-318139.
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The topic of this work is the renewal of hermetic pipe penetrations within the primary loop of a nuclear power station. A description of the Dukovany nuclear power plant is included at the beginning, ranging from basic description to details of the renewed component and the reasoning for its replacement. Following is a description of the technologies, which are applied in order to restore the hermetic pipes, the choice of the placement of the heterogeneous weld and the compilation of a workflow for the renewal in the environment of the primary loop while maintaining the quality standards and the strict safety conditions. The conclusion includes a calculation of the expenses tied with the chosen variant the reasoning behind the renewal regarding the length of the outage.
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Šula, 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.
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This thesis describes the provision of the data communication of the new protection relays via 6 kV switch gears to the electro monitoring system at the Dukovany Nuclear Power Station. These protection relays will replace the current analog protection. The replacement process will start in 2015 and will finish in 2018. Given the overall complexity of the project, this thesis deals only with the following 6 kV switch gears 9BB and 9BD. The thesis is divided into a theoretical and a practical part. The theoretical part concerns the actual consumption and output of the Dukovany Nuclear Power Station. The following chapters describe the electro monitoring system and GRAF and LOGA software programs that are closely connected with it and that were used in the practical part. The last chapter of the theoretical part deals with the actual replacement of the protection relays of the 6 kV switch gears. Also, it describes a new optical network, which will be set up at the Dukovany Nuclear Power Station as a part of the process of replacement. The segment of the practical part concerns laboratory verification of the data communication of the new protection relays. Next, it describes actions that had to be taken in order to ensure and verify correct functionality of the data communication. This part is followed up by a closely related practical activity, which describes examination and verification of the protection functions of each of the 9BB and 9BD switch gears protection relays.
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Tomoryová, Bianka. "Facelift EDU". Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414291.
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The subject of the diploma thesis is an architectural study of the forecourt of the Dukovany Nuclear Power Station. The project is based on a pre-diploma project which aimed to develop an urban concept. At the moment the buildings in the area are in an unsuitable condition from a construction - technical and architectural perspective. The urban-architectural study was intended to create a new concept for the forecourt and simultaneously enable its efficient operation in case of the completion of a new nuclear unit. The concept of the solution is based on the removal of outdated buildings and their replacement with new buildings which will meet the current requirements of the Dukovany Nuclear Power Station. From the current layout the main roads, the gatehouse, the information center and the greenery will be preserved. After removing the selected buildings, the renovation area will be divided into the main axes on which the main road network will be designed. The unsuitable connection to public road infrastructure will be solved by minimizing the entrances to the area to two entrances and by moving the suburban bus station from the main road. After connecting the whole area by roads, the area will be divided into functional areas. In front of the gatehouse there will be a square surrounded by public services, an administrative building and by the preserved information center building. From this space, the main representative space of the entire complex will be created. Warehouses and workshops will be located behind the square. Transport areas will be located at the rear of the complex. Thanks to this location, there will be a good connection with the current factory and also with the planned completion of the new nuclear unit on the west of the area. There will also be educational and sport facilities in the complex. The kindergarten will be oriented towards the preserved greenery.
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Kissler, Martin. "Modernizace Jaderné elektrárny Dukovany". Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2015. http://www.nusl.cz/ntk/nusl-231807.
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Thesis focuses on a detailed technical description of all important parts of secondary circuit in Dukovany power plant and its connection to other systems of power plant. In thesis are analyzed significant adjustments which have been made during the entire operation of power plant including in particular the actions associated with project called Utilization of project reserves of units EDU. In the main part of the thesis were carried out calculations of the power plant's power for states before and after the modernization and there is also analyzed the impact of individual changes on the whole power plant. These changes are with the entire secondary circuit drawn in the T-s diagram.
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Rygl, 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.
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This thesis solves the problem of design and production of a special mounting device for the impeller nut of the main circulation pump GCN-317. The proposed device is intended to solve the problems associated with the mounting and dismounting of this nut. The work provides basic information about the Dukovany Nuclear Power Plant and VVER-440 systems in general, including their brief history. It also introduces the operational and legislative environment of the primary circuit of the nuclear power plant. It deals with the task and technical description of the main circulation pump and its overhaul. The following sections present the basic principles of construction of the device and an overview of its components with a description of their purpose and method of production. The last part of the thesis describes the function of the device and its verification and reports on its deployment.
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Žák, 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.
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The present Master´s thesis aims at designing an evolutionary scheme of secured power supply at nuclear power plants with VVER 440 reactors during design basis accidents as well as design extension conditions. In the first part of this thesis, concepts relating to the defence in depth of nuclear reactors, operating modes of the blocks as well as types and possibilities of electrical power supply and electric circuits of the block are defined. Although the present thesis deals with PWR 440 in general, special emphasis is put on the Czech NPP in Dukovany, where there are four PWR 440 reactors in operation, and on the possibilities for enhancing the defence in depth in this area. The second part of the thesis deals with the difference in station blackout definitions before and after Fukushima; not only the differences in situation evaluation are dealt with, but a solution is also proposed to make the system of secured power supply system during design basis accidents as well as design extension conditions more robust. This option has been selected out of a number of possibilities based on the evaluation of reliability, availability and cost-effectiveness of the proposal.
25
Rokotianskaia, Kseniia. "Facelift EDU". Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414282.
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The subject of the diploma thesis is the elaboration of an architectural study of the reconstruction of the pre-plant zone of the Facelift of the Dukovany power plant. The construction site is an area that belongs to the village of Dukovany and borders the village of Rouchovany. As a whole, this area is in poor technical and architectural condition. However, its location gives potential for new uses. The solved area belongs to the ČEZ nuclear power plant.
26
Zhakupbekova, Rakhil. "Facelift EDU". Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414302.
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The subject of the diploma thesis is the elaboration of an architectural study of the reconstruction of the pre-plant zone of the Facelift of the Dukovany power plant. The construction site is an area that belongs to the village of Dukovany and borders the village of Rouchovany. As a whole, this area is in poor technical and architectural condition. However, its location gives potential for new uses. The solved area belongs to the ČEZ nuclear power plant.
27
Vacek, 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.
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The goal of this project is to test the possibility to connect the Cogenerational generation of power 138 MW (still in the development stage) to the control room 110 kV in Prostejov production. This merge would product the electrical energy as well as the heat energy for all local area. In this dissertation we will be considering the solution of the steady state (stationary state) of system with the voltage level of 110kV, as well as the influence of the generation of power on this system, there by the suggesting a connection. The Congenerational production indicates higher effectiveness in the transformation of energy during primary production process due to the production of heat energy as well as the electrical energy from the primary power sources. In our country, as well as around the world, commonly used fuels are fossil fuels- coal, crude oil, and gas. As the demand for energy grows, those supplies are slowly running out. Not to mention that those fuels have a negative environmental impact. They are a source of carbon, which causes damage to the atmosphere and leads to global warming. Power plants which do not produce carbon are much safer for the environment, and much more productive. However, the residue of this energy is challenging to dispose of. Nuclear energy has common attributes with renewing the sources of energies that are extremely friendly to our environment. Nuclear power plants also produce enough energy and with the usage of Fourth generation reactors, they will be able to recycle the nuclear fuels. Today, more importance is put on renewing sources which are more gentle for the environment. In the near future, CEZ Company, the largest producer of electric energy is planning to use water energy. Water energy comes from water plants or dams. Other ecological forms of energy include geothermal and solar energies. These two types of energy are not as applicable for our geographical position. Geothermal energy is commonly used on islands where there is an abundance of natural hot springs. The most discussed source of energy is bioenergy. It uses natural wood sources, recycled wood products, and applies bioenergy as a main source for thermal power plants.
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Růžičková, Tereza. "Facelift EDU". Master's thesis, Vysoké učení technické v Brně. Fakulta stavební, 2020. http://www.nusl.cz/ntk/nusl-414284.
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The theme of the diploma thesis is the solution of the pre-plant zone of the Dukovany nuclear power station. Creating a vision of how this space could develop further in the next 50 years and how it could work in a transitional phase during the construction of a new nuclear power plant unit. The subject of the thesis is the elaboration of an architectural study, which is based on an urban study. Urban study was processed within the framework of the pre-diploma thesis and solved mainly the overall problems of this area, the new transport connection, and the functional division of the whole area. The area was divided into three functional units, namely the transport zone, the administrative zone and the sports and education zone. At present, there are a lot of small structurally and functionally unsuitable buildings in this area. The diploma thesis deals with the design of new buildings with a clear functional use in the administrative zone, in the area in front of the main gatehouse. A new representative square was created, and three buildings are designed around it. The dominant feature of the whole area was a high-rise office building near the gatehouse. In front of the office building, towards the main road, the service building I was designed, which contains business and healthcare services. On the other side of the square was located the service building II, where there are technical services, such as workshops, warehouses and metrology. The last building solved within the diploma thesis is the building of sports and locker rooms of suppliers, which is designed behind the square near the greenery. New building copies mass of the only preserved building in this area, namely the building of the information centre. The designed building has a fitness centre and locker rooms of external workers of the power plant.
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Li, Liang-Ying y 李亮瑩. "Transient analysis of Lungmen Nuclear Power Station using RELAP5-RT". Thesis, 2009. http://ndltd.ncl.edu.tw/handle/87671786912834374015.
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碩士國立清華大學
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.
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Chen, Yu-Chen y 陳宥辰. "Transient Analysis of Lungmen Nuclear Power Station using RELAP5-RT". Thesis, 2010. http://ndltd.ncl.edu.tw/handle/29564820682766731914.
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碩士國立清華大學
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.
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Cheng, Hsin y 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.
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碩士國立清華大學
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.
32
Lai, Yu-Cheng y 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.
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碩士國立清華大學
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.
33
Chang, Ho-Yu y 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.
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Nordt, Kevin M. "MAAP/MELCOR comparison station blackout at the point beach nuclear power plant /". 1992. http://catalog.hathitrust.org/api/volumes/oclc/26109393.html.
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Thesis (M.S.)--University of Wisconsin--Madison, 1992.Typescript. eContent provider-neutral record in process. Description based on print version record. Includes bibliographical references (leaves 93-94).
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Huang, Meng Ting y 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.
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碩士國立清華大學
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.
36
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.
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Liang, Ching-Chun y 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.
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碩士中原大學
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.
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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.
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碩士國立清華大學
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.
39
Huang, Li-Hua y 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.
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碩士中原大學
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.
40
Tsai, Tseng-I. y 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.
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Fu, Yu-Feng y 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.
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碩士國立清華大學
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.
42
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.
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A research report submitted to the Faculty of Science, University of the Witwatersrand, Johannesburg, in partial fulfilment of the requirements for the degree of Master of Science, Joannesburg, 2014.An 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.
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Hsu, Keng-Hsien y 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.
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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.
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碩士國立清華大學
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.
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Wu, Wei-lih y 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.
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碩士國立中山大學
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.
46
Lin, Tser-Tung y 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.
Texto completo
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HU, ZHONG-QING y 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.
Texto completo
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Yang, Shao-yu y 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.
Texto completo
49
Yang, Jian-Hong y 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.
Texto completoResumen
碩士中原大學
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.
50
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.
Texto completoResumen
碩士國立清華大學
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.