Global ETD Search

1	Establishment of technical basis for the nuclear emergency planning and preparedness programme for the Pebble Bed Modular Reactor(PBMR) nuclear power station / Pulane Adelaide Moleme Moleme, Pulane Adelaide January 2003 (has links) In this work, description and explanation for the conditions of emergency preparedness is given. The aim was to establish the technical basis for emergency response plans in the event of a nuclear incident that might occur at the PBMR nuclear power station. The Koeberg Nuclear Power Station Technical Basis for Emergency Planning was basically used as a guideline and PC COSYMA, a programme that runs on PC, was used to help ,model the release as described or specified by the regulations set in [5] for South Africa. PC COSYMA runs were done to calculate doses to specific organs and to the whole body. These runs were done considering two weather categories (Pasquill stability category D and F) for neutral and moderately stable weather conditions. Wind speed of 2m/s was used for exposure dose integration periods 1, 10, 30 and 70 days. All dose results (organ and effective) were given by pathways: groundshine, cloudshine and inhalation, except thyroid doses whereby it was assumed that all doses are essentially committed through inhalation. A run for Thyroid including inhalation only as a pathway was also done and a run for iodine prophylaxis as protective action to see the effectiveness of the countermeasures. All the calculated doses were lower than the individual radiation dose limit (50mSv) set by the regulator (NNR) for PBMR [20]. The implementation of countermeasures prevented a further dose accrual. Therefore, there is no need for implementation of countermeasures at PBMR, since they are well below the set criterions. But it will be a good safety culture and Defense in Depth to make an allowance for an emergency plan. Since Defense in Depth is a principle that requires that there should be multiple layers of overlapping safety provision and good safety culture looks at the behaviour on doing things, / MSc. (ARTST) North-West University, Mafikeng Campus, 2003 Emergency management Nuclear power plants-Accidents
2	SBLOCA analysis for nuclear plant shutdown operations Wang, Yi 11 March 1994 (has links) A series of small break loss of coolant accident (SBLOCA) analyses in nuclear plant shutdown operations was simulated using the code RELAP5A,MOD3 version 8.0 to predict the SBLOCA phenomena in the Zion-l nuclear power plant The first objective is to study the impact of SBLOCA (1" and 2" breaks) on plant conditions while in the shutdown mode. In particular, to determine the time to "core uncovery" without operator interaction. The other objective is to study the effect of RHR heat exchanger elevation on natural circulation mass flow rate, fluid temperature and peak fuel pin temperature. Peak temperature and time to core uncovery were found for two small break LOCA cases. The natural circulation mass flow rate after break initiation was affected by varying the RHR heat exchanger elevation. The system pressure and temperature were not affected much by the elevation change in the RHR heat exchanger. The current version of RELAP5/MOD3 was found to be sensitive to the initial conditions in studies of low pressure,low temperature plant systems, especially for a large break LOCA. / Graduation date: 1994 Nuclear power plants -- Accidents
3	An investigation of coastal fumigation effects on nuclear accident consequences in Hong Kong Huang, Aiping, 黃愛平 January 1996 (has links) published_or_final_version / Mechanical Engineering / Master / Master of Philosophy Nuclear power plants - China - Daya Bay. Fumigation.
4	Nuclear emergency preparedness model based on Daya Bay Nuclear Power stations for educational purposes Cheng, Kit-yan, Ruby., 鄭潔茵. January 2005 (has links) published_or_final_version / abstract / Physics / Master / Master of Philosophy Nuclear accidents - Computer simulation.
5	Loss of normal feedwater ATWS for Vogtle Electric Generating Plant using RETRAN-02 Rader, Jordan D. 16 October 2009 (has links) With the ever advancing state of computer systems, it is imperative to maintain the most up-to-date and reliable safety evaluation data for nuclear power systems. Commonplace now is the practice of updating old accident simulation results with more advanced models and codes using today's faster computer systems. Though it may be quite an undertaking, the benefits of using a more advanced model and code can be significant especially if the result of the new analysis provides increased safety margin for any plant component or system. A series of parametric and sensitivity studies for the Loss of Normal Feedwater Anticipated Transient without Scram (LONF ATWS) for Southern Company's Vogtle Electric Generating Plant (VEGP) Units 1&2 located near Waynesboro, GA was performed using the best-estimate thermal-hydraulics transient analysis code RETRAN-02w. This thesis includes comparison to the results of a generic plant study published by Westinghouse Electric Corporation in 1974 using an earlier code, LOFTRAN, as well as Vogtle-specific analysis. The comparative analysis exposes and seeks to explain differences between the two codes whereas the Vogtle analysis utilizes data from the Vogtle FSAR to generate plant-specific data. The purpose of this study is to validate and update the previous analysis and gather more information about the plant actions taken in response to a LONF ATWS. As a result, now there is a new and updated evaluation of the LONF ATWS for both a generic 4-loop Westinghouse plant and VEGP using a more advanced code. Beyond the reference case analysis, a series of sensitivity and parametric studies have been performed to show how well each type of plant is designed for handling an ATWS situation. These studies cover a wide range of operating conditions to demonstrate the dependability of the model. It was found that both the generic 4-loop Westinghouse PWR system and VEGP can successfully mitigate a LONF ATWS throughout the core's operating cycle. Thermal-hydraulics Loss of feedwater Nuclear power plant safety Transient analysis ATWS LONF Nuclear power plants Safety measures
6	Assessing internal contamination levels for fission product inhalation using a portal monitor Freibert, Emily Jane 18 November 2010 (has links) In the event of a nuclear power plant accident, fission products could be released into the atmosphere potentially affecting the health of local citizens. In order to triage the possibly large number of people impacted, a detection device is needed that can acquire data quickly and that is sensitive to internal contamination. The portal monitor TPM-903B was investigated for use in the event of a fission product release. A list of fission products released from a Pressurized Water Reactor (PWR) was generated and separated into two groups--Group 1 (gamma- and beta-emitting fission products) and Group 2 (strictly beta-emitting fission products.) Group one fission products were used in the previously validated Monte Carlo N-Particle Transport Code (MCNP) model of the portal monitor. Two MIRD anthropomorphic phantom types were implemented in the MCNP model--the Adipose Male and Child phantoms. Dose and Risk Calculation software (DCAL) provided inhalation biokinetic data that were applied to the output of the MCNP modeling to determine the radionuclide concentrations in each organ as a function of time. For each phantom type, these data were used to determine the total body counts associated with each individual gamma-emitting fission product. Corresponding adult and child dose coefficients were implemented to determine the total body counts per 250 mSv. A weighted sum of all of the isotopes involved was performed. The ratio of dose associated with gamma-emitting fission products to the total of all fission products was determined based on corresponding dose coefficients and relative abundance. This ratio was used to project the total body counts corresponding to 250mSv for the entire fission product release inhalation--including all types of radiation. The developed procedure sheets will be used by first response personnel in the event of a fission product release. Internal contamination Fission products Inhalation Accident scenario TPM 903B Portal monitor Nuclear power plants Accidents Fission products
7	A safety and dynamics analysis of the subcritical advanced burner reactor: SABR Sumner, Tyler Scott January 2008 (has links) Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2008. / Committee Chair: Willem F.G. Van Rooijen; Committee Member: Ghiaasiaan, Seyed M; Committee Member: Weston M. Stacey
8	A safety and dynamics analysis of the subcritical advanced burner reactor: SABR Sumner, Tyler Scott 03 June 2008 (has links) As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing the nuclear fuel cycle has become the next major technical challenge for the nuclear energy industry. By separating the transuranics (TRU) from the SNF discharged by Light Water Reactors, it is possible to fuel Advanced Burner Reactors to minimize the amount of SNF that must be stored in High Level Waste Repositories. One such ABR concept is the Subcritical Advanced Burner Reactor (SABR) being developed at the Georgia Institute of Technology. SABR is a subcritical, sodium-cooled fast reactor with a fusion neutron source capable of burning up to 25% of the TRU fuel over an 8.2 year residence time. In the SABR concept an annular core with a thickness of 0.6 m and an active height of 3.2 m surrounds the toroidal fusion neutron source. Neutron multiplication varies during the lifetime of the reactor from keff = 0.95 at the beginning of reactor life to 0.83 at the end of an equilibrium fuel cycle. Sixteen control rods worth 9$ are symmetrically positioned around the reactor. This thesis describes the dynamic safety analysis of the coupled neutron source, reactor core and reactor heat removal systems. A special purpose simulation model was written to predict steady-state conditions and accident scenarios in SABR by calculating the coupled evolution of the power output from the fusion and fission cores and the axial and radial temperature distributions of a fuel pin in the reactor. Reactivity Feedback was modeled for Doppler and sodium coolant voiding. SABR has a positive temperature reactivity feedback coefficient. A series of accident scenarios were simulated to determine how much time exists to implement corrective measures during an accident before damage to the reactor occurs. SABR Fusion Neutron kinetics Thermal hydraulics Subcritical Sodium Reator fuel processing Waste disposal Sodium cooled reactors Nuclear power plants Accidents Nuclear reactors accidents Nuclear accidents Nuclear reactors Safety measures
9	Coverage of the Fukushima crisis in the two major English-language newspapers in Japan : a critical analysis Finn-Maeda, Carey 11 1900 (has links) This study uses a mixed-method approach to analyse the coverage of the 2011 Fukushima nuclear crisis in Japan’s two major English-language newspapers – The Japan Times and The Daily Yomiuri. Quantitative coding is combined with critical discourse analysis to determine whether the coverage was, overall, predominantly alarming, reassuring, or relatively balanced and neutral. This is done to ascertain whether the newspapers were sensationalising the crisis, echoing the official government and industry communication thereof, or reporting in a critical, responsible manner as the fourth estate. To answer the research question, key aspects of the coverage like foci, framing, sources, narratives, actors and agency, and criticisms are closely examined. It is revealed that the coverage was neither predominantly alarming nor reassuring, but was problematic in other ways. The implications of the complex findings, both for the Japanese media industry and international disaster reporting, are discussed. The study is situated in a broad literature framework that draws on agenda setting theory, research about the roles and responsibilities of the media, the field of risk communication and the reporting of radiation events in history. / Communication Science / M.A. (Communication) Fukushima Content analysis Critical discourse analysis The Japan Times The Daily Yomiuri Radiation Risk communication Disaster reporting Fourth estate Media 070.44936317990952117 Fukushima Nuclear Disaster, Japan, 2011 Content analysis (Communication)
10	Application of the Stimulus-Driven Theory of Probabilistic Dynamics to the hydrogen issue in level-2 PSA / Application de la Stimulus Driven Theory of Probabilistic Dynamics (SDTPD) au risque hydrogène dans les EPS de niveau 2. Peeters, Agnes 05 October 2007 (has links) Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles.<p>Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel.<p>Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre.<p><p>Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios.<p>These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient.<p>This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.<p> / Doctorat en Sciences de l'ingénieur / info:eu-repo/semantics/nonPublished Sciences de l'ingénieur Physique Nuclear energy Nuclear power plants -- Accidents Nuclear power plants -- Safety measures Nuclear reactors -- Safety measures Nuclear reactors -- Containment Energie nucléaire Centrales nucléaires -- Accidents Réacteurs nucléaires -- Confinement PSA STIMOCCO reliability fiabilité stimulus dynamic benchmark SARNET sûreté étude safety analysis probabiliste explosion deflagration niveau 2 severe accident accident grave cell-to-cell mapping combustion EPS cellules Petri dynamique probabilistic level-2 nuclear energy hydrogen énergie nucléaire centrale nucléaire nuclear power plant hydrogène délai delay

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