The goal of this thesis is to simulate the course of events during Dukovany Nuclear Power Plant blackout and to determine the optimal process of cooling down the cold branches of the primary circuit loops to secure the maximum amount of the primary fuel needed for the residual heat outlet so that the operating staff has as much time as possible for renewing the electric power supply. The first part of the thesis describes nuclear power plants built in the Czech Republic and in the world as well as reactor blocks whose construction is being considered. The detailed description of Dukovany’s steam generator is shared in chapter three. Following chapter summarizes blackouts that occurred at power plants around the world, events that might have led to blackouts in the Czech Republic, and it also describes current approach to blackout problematics at Dukovany Nuclear Power Plant. Chapters six and seven contain the core of the thesis. That includes detailed description of a mathematical model explaining the behaviour of a reactor block during blackout as well as the analysis of the results found.

The master's thesis deals with the analysis of source term of the spent nuclear fuel of the Dukovany Nuclear Power Plant in order to determine the proposals for the transfer of spent nuclear fuel to a deep geological repository in the Czech Republic. To introduce the reader into the issue are briefly described the main aspects, such as the development of the nuclear fuel used in the history of the Dukovany Nuclear Power Plant. These aspects have an influence on the final draft of the timetable. One of the important partial tasks is the processing of an estimate of the future range of spent nuclear fuel, which is based on the current ideas of company ČEZ, a.s. for the future direction of the fuel cycle at the Dukovany nuclear power plant. For the purposes of this work, the key data are the time dependencies of radioactivity and the development of residual heat in individual fuel assemblies. This data are calculated by the software called PAL440_R4, based on the prepared estimates of the spent nuclear fuel assortment. The calculated data are then edited and sorted by MS Excel. For the sake of completeness, the characteristic values and the time dependencies of the radioactivity and the development of residual heat in fuel assemblies. Final timetables for the transfer of spent nuclear fuel to a deep geological repository are processed in several variants, and their selection and application options are justified. For illustration are important parameters given in the form of tables and charts.

Viscous dampers are widely used to ensure seismic resistance of pipelines and equipment in nuclear power plants. Damping characteristics of these dampers are nonlinearly frequency dependent and thus causing complications in computational modelling of seismic response. Engineers commonly use two ways to deal with this nonlinearity: The first option is to consider damper by means of “snubber”. This is essentially linear spring element that is active for dynamic load and does not resist static loads. Snubber behaviour during seismic event is described by a equivalent stiffness (sometimes called pseudostiffness). The equivalent stiffness could be defined by the iterative calculations of piping natural frequencies and mode shapes taking into account seismic excitation. However, in complicated structures such as the main circulation loop of nuclear power plant the iterative calculation is difficult and could bring significant inaccuracies. On the other hand, the benefit of such modelling is a possibility to apply the commonly used linear response spectrum method for a solution. The second option is to describe damping characteristics using suitable rheological model. The seismic response is than determined by direct integration of the equations of motion. The behaviour of dampers is described exactly enough but the calculation and post-processing, especially nodal stresses time-histories, are time consuming. The goal of this work was to find a methodology for determining the seismic response of complex pipe systems with viscous dampers. Methodology allows a sufficiently accurate determination of the seismic response of piping systems and also allows obtaining of the results in effective time. The procedure is as follows. Firstly, specialized piping program (AutoPIPE) is used for the development of computational model. Next step is to determine a static response of structure and its verification with experimental measurements, if possible. Using script in Python language a computational model is converted from AutoPIPE into general finite element model in ANSYS system. Four-parameter Maxwell rheological model is used to describe behaviour of viscous dampers. Seismic load is represented by synthetic accelerograms. Newmark algorithm of direct integration of the equation of motion is used to obtain seismic response (only reactions and displacements in nodes of interest are necessary). Than is the equivalent stiffness is than gained from displacements and reactions as median value of their ratios. Received stiffness are subsequently transferred to AutoPIPE program where the seismic solution is performed using response spectra method. Finally, the dynamic response is combined with the static response and stress assessment according standards is done. The created methodology was applied in the seismic resistance calculation of the main circulation piping and piping of pressurizer in nuclear power plants type VVER 440 and type VVER 1000.

This thesis is focused on the design of passive cooling system for a nuclear reactor VVER- 1000.This type of reactor is located in the Czech Republic in the location of Nuclear power plant Temelín. The thesis states an overview of the different cooling systems of nuclear power plants. The thesis is focused on passive safety system especially on passive cooling system, so there was done an overview of currently used passive safety system. In the work is discussed nuclear safety and the maximum projected accident of VVER-1000, which is called LOCA accident. In the design part of the thesis was done thermal calculation of heat exchangers. Exchangers are designed as condensers with a natural flow, where cooling of system is provided by outside airflow in case an accident. The results are evaluated at the end of the thesis.

The flow distribution in the primary circuit of the pressurized water reactor was studied with experiments and Computational Fluid Dynamics (CFD) simulations. The main focus was on the flow field and mixing in the downcomer of the pressure vessel: how the different factors like the orientation of operating loops, the total loop flow rate and the asymmetry of the loop flow rates affect the outcome. In addition to the flow field studies the overall applicability of CFD methods for primary circuit thermal-hydraulic analysis was evaluated based on the CFD simulations of the mixing experiments of the ROCOM (Rossendorf Coolant Mixing Model) test facility and the mixing experiments of the Paks NPP. The experimental part of the work in work package 3 included series of steady state mixing experiments with the ROCOM test facility and the publication of results of Paks VVER-440 NPP thermal mixing experiments. The ROCOM test facility models a 4-loop KONVOI type reactor. In the steady-state mixing experiments the velocity field in the downcomer was measured using laser Doppler anemometry and the concentration of the tracer solution fed from one loop was measured at the downcomer and at the core inlet plane. The varied parameters were the number and orientation of the operating loops, the total flow rate and the (asymmetric) flow rate of individual loops. The Paks NPP thermal mixing experiments took place during commissioning tests of replaced steam generator safety valves in 1987-1989. It was assumed that in the reactor vessels of Paks VVER-440 NPP equipped with six loops the mixing of the coolant is not ideal. For the realistic determination of the active core inlet temperature field for the transients and accidents associated with different level temperature asymmetry a set of mixing factors were determined. Based on data from the online core monitoring system and a separate mathematical model the mixing factors for loop flows at the core inlet were determined. In the numerical simulation part of the work package 3 the detailed measurements of ROCOM tests were used for the validation of CFD methods for primary circuit studies. The selected steady state mixing experiments were simulated with CFD codes CFX-4, CFX-5 and FLUENT. The velocity field in the downcomer and the mixing of the scalar were compared between CFD simulations and experiments. The CFD simulations of full scale PWR included the simulation of Paks VVER-440 mixing experiment and the simulation of Loviisa VVER-440 downcomer flow field. In the simulations of Paks experiments the experimental and simulated concentration field at the core inlet were compared and conclusions made concerning the results overall and the VVER-440 specific geometry modelling aspects like how to model the perforated elliptic bottom plate and what is the effect of the cold leg bends to the flow field entering to the downcomer. With Loviisa simulations the qualitative comparison was made against the original commissioning experiments but the emphasis was on the CFD method validation and testing. The overall conclusion concerning the CFD modelling of the flow field and mixing in the PWR primary circuit could be that the current computation capacity and physical models also in commercial codes is beginning to be sufficient for simulations giving reliable and useful results for many real primary circuit applications. However the misuse of CFD methods is easy, and the general as well as the nuclear power specific modelling guidelines should be followed when the CFD simulations are made.

The function of safety reviews (here understood as expert judgements on proposals for design modifications and redesign of technical systems in commercial Nuclear Power Plants, supported by formalised safety review processes) plays a fundamental role for safety in nuclear installations. The primary aims of the presented case studies includes: critically examining and identifying the main areas for improvement of the existing technical safety review process as it is conducted at a Swedish nuclear power plant, developing a new process, and evaluating whether any improvements were accomplished. By using qualitative methods, observation/participation and interviews, data has been gathered on how the safety review process is perceived and conducted by experts involved in the safety review process, and ways to improve this process have been developed. This area is neglected in the larger safety literature. The novel approach here is to gather data directly from those involved in the safety review process, analysis of safety review reports as well as from inspection reports by the regulatory authority. The study presented in paper I shows that the partition between primary and independent review is positive, having supplementary roles with different focus and staff with different skills and perspectives making the reviews. The study identifies a number of areas for improvement, such as: - a tendency to put too much resource on minor assignments - a clearer prioritization would improve focus on the most critical projects - there is a need for improved guidance and direction for how to structure the work It is argued that future applications of safety review processes should focus more on communicating and clarifying the process and its adherent requirements, and improve the feedback system within the process. It is also recommended that the NPPs create introductory training for new reviewers The study presented in paper II concluded that grading of the primary safety review reports facilitates improved experience feedback by providing easier access to good examples for reviewers. Improvements identified by implementing the revised process are primarily linked to the independent safety review function, including better planning and means for resource allocation as well as clearer and more unambiguous supporting instructions. Introduction of formalized independent review meetings provides increased exchange of knowledge and strengthened the independent safety review function in the organization. / <p>QC 20130305</p>

From 1976 to 1992 Portland General Electric (PGE) -- a private utility based in Portland, Oregon -- operated the Trojan Nuclear Plant near Rainier, Oregon, on the bank of the Columbia River. Trojan was the first commercial nuclear facility in the Pacific Northwest and was the largest such facility in U.S. history. From its origins, Trojan was the focus of growing conflict over atomic energy facilities and their environmental effects, risks, and costs. This thesis traces the history of Trojan, including the conditions in which PGE decided to build the plant as well as the changing conditions in which the environmental movement in Oregon worked to impact the operation of Trojan and the development of further atomic energy facilities in the region. Two sets of values, largely endemic to the region, came into conflict in the debate over Trojan: one which valued preservation of vital natural systems over all else, and another that elevated technological progress to supreme importance in achieving the ultimate social good. Supporters of Trojan and anti-nuclear activists both viewed misinformation about nuclear power as one of the central problems in the way that Oregon residents viewed nuclear power. Although there were many loyal supporters of Trojan, particularly in Columbia County, there were also a great number who viewed the technology cautiously. While both PGE and nuclear opponents worked diligently to sway public opinion, many activists did so by attempting to uncover and publicize hidden information about the design and operation of Trojan, and the nuclear fuel cycle in general. This included efforts throughout the plant's lifetime to develop opportunities for intervention in administrative proceedings, government hearings, and other arenas which often discourage citizen involvement. Related to the public debate over Trojan were ongoing operational difficulties and changing economic conditions, which contributed to the decision PGE announced in 1993 that Trojan would be permanently shut down. This study is based primarily on coverage from newspapers and periodicals, new and extant oral history interviews, documents from the personal files of activists, as well as various archival materials associated with PGE, activist groups, and government agencies.

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

本文共乙冊，約八萬言，分為五章十節。 第壹章 緒論。首先闡釋本文題目之意義；其次陳述本文之研究動機、目的方法與限 制；最後說明本文研究之架構。 第貳章 假設分析的運作背景。從組織環境與組織問題的分析中，以瞭解運用假設分 析方法之緣由。 第參章 假設分析的運作。首先論述假設分析運作的理論基礎；其次解析其運作之程 序與方法。 第肆章 假設分析的應用。首先說明政策論證的內涵，模式及其與假設分析方法之關 係；其次試圖運用本文所論述之理論與方法，以檢視和評析台灣電力公司興建核能四 廠一案的政策過程。 第伍章 結論。發表本文研究之心得並提出假設分析方法成功運作的要件。