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。

本文共壹冊，分柒章，約四萬言，章節目錄如下： 第一章：導論 一、前言，二、研究動機，三、研究目的，四、研究架構，五、研究限制。 第二章：緊急事故本質探討 一、名詞解釋，二、緊急事故性質，三、緊急事故影響與後果，四、面臨緊急事故時 之個人與組織行為，五、災變之防治。 第三章：緊急事故的管理 一、管理架構，二、管理之規劃與控制活動。 第四章：決策支援系統理論基礎 一、決策支援系統定義與特性，二、傳統EDP,MIS 與DSS 之比較，三、系統建立方法 。 第五章：緊急事故管理之決策支援系統設計 一、緊急事故下之決策程序，二、決策特徵，三、功能架構，四、系統建立。 第六章：個案實例 一、個案背景介紹，二、核能電廠緊急應變措施，三、系統需求，四、建立與實施。 第七章：結論與建議 一、結論，二、建議。 環繞人類四周環境中，常有許多不確定的災變隨時可能降臨。且發生災變時，如果資 訊缺乏或運用不當，常造成不必要的損失與傷亡。本研究係研究有關決策支援系統在 緊急事故管理上之應用，籍資訊的提供與利用，以支援緊急事故防治。 決策支援系統具有（一）易於使用，（二）模式庫與資料庫整合，（三）適於解決非 結構性問題，（四）具有良好彈性等特性，而緊急事故管理更明顯涉及決策者的價值 判斷，及對不確定環境的偏好，因此面臨這種結構程度低的問題，一個考慮完整的決 策支援系統將能提供管理人員更有效的支援。 本文之研究共分四部份，第一部份探討緊急事故的特性與本質，及面臨災變時人與群 體的行為與反應，第二部份探討緊急事故管理之決策程序與決策特性，第三部份則以 文獻分析方式探討決策支援系統相關文獻，以作為建造系統時指引，第四部份則以個 案研究法，針對核能電廠緊急事故之疏散掩蔽決策，提出應建立之決策支援系統。

A Agência Internacional de Energia Atômica (IAEA - \"International Atomic Energy Agency\") vem incentivando o desenvolvimento de sistemas passivos de refrigeração em plantas nucleares visando a simplificação e o incremento da confiabilidade em funções essenciais de segurança nos projetos de uma próxima geração de reatores nucleares refrigerados a água. O principal fundamento desses sistemas é o emprego da circulação natural como sistema de segurança aplicável em operações de desligamento do reator para manutenção ou na ocorrência de acidentes. A circulação natural é um fenômeno que surge em virtude do gradiente de temperatura em pontos diferentes do circuito de refrigeração. Em condições extremas de estabilidade têm-se o estabelecimento do escoamento bifásico gás/líquido podendo configurar-se segundo diferentes regimes. A fração de vazio é reconhecida como um dos parâmetros chave na predição da ocorrência de instabilidades do escoamento bifásico. Apresenta-se neste trabalho uma inovadora metodologia para estimativa da fração de vazio a partir de imagens digitais capturadas diretamente de circuitos experimentais que geram o escoamento bifásico. O método é baseado na aquisição de imagens, com controle da profundidade de campo, de uma seção do Circuito de Circulação Natural (CCN) presente no IPEN/CNEN-SP. A imagem é segmentada com base na inferência fuzzy de diferentes parâmetros de segmentação e ajustada ao foco utilizado na sua aquisição. Ela é varrida de um modo inédito e iterativo, utilizando máscaras de diferentes tamanhos integrando um conjunto de redes neurais com a Transformada Randomizada de Hough. Cada diferente tamanho de máscara é escolhido de acordo com os tamanhos das bolhas que são os objetos de interesse. O volume da bolha é estimado baseado em sua projeção plana capturada nas imagens digitais. O cálculo da fração de vazio considera o volume da seção geométrica do escoamento no tubo de vidro cilíndrico e a profundidade de campo utilizada e nos parâmetros geométricos inferidos para cada bolha detectada. Os resultados mostraram que a integração entre o conjunto de redes neurais e a Transformada Randomizada de Hough aumentaram a robustez das estimativas do sistema. / The International Atomic Energy Agency (IAEA) has been encouraging the use of passive cooling systems in new designs of nuclear power plants. Next nuclear reactor generations are intended to possess simpler and robust safety functions. Natural circulation based systems hold an undoubtedly prominent position among these. Natural circulation phenomenon occurrence depends only on the existence of refrigerant liquid temperature gradient in different sections of the plant refrigerator circuit. The study of limit conditions for these systems has led to instability behavior analysis where many different two-phase flow patterns are present. Void fraction is a key parameter in thermal transfer analysis of theses flow instability conditions. This works presents a new method to estimate void fraction from digital images captured at an experimental two-phase flow circuit. The method is primarily based on depth-of-field controlled image acquisition of a section of a closed loop of natural circulating water through cylindrical glass tubes. This loop is called Natural Circulation Facility (NCF) and is located at Nuclear and Engineering Research Institute in Brazil (IPEN/CNEN-SP). Image is segmented based on fuzzy inference of different segmentation parameters and adjusted to image acquisition focus. The image is then scanned in an inedited way using different-sized masks integrating a set of different artificial neural networks with a modified Randomized Hough Transform. Each different mask size is chosen in accordance to bubble sizes which are objects of interest. The bubble volume is estimated based on two-dimensional projection sizing based on digitally acquired images. Void fraction calculation takes into account the volume of the geometrical section of flow inside cylindrical glass tube considering used depth-of-field. It is also based on the summed bubble geometrical parameters inferred for each detected bubble. The results have shown that integration between artificial-neural-net sets and Randomized Hough Transforms increase system estimations robustness.

會計學教課書都教學生：決策之作成，應從成本效益出發，至於過去已投入的成本屬沉沒成本，非攸關，不應影響決策，讀者少有質疑，但事實究竟如何？實務上，決策者做決策時，考量之因素甚多，不會只有成本一項，不過教課書多只著重成本一項，這又與事實相左。本研究以「核四決策」為例，探討實務上決策之作成，在諸多因素中沉沒成本是否真的沒攸關性。 核四應否停建，不論在能源、社會政策等層面皆具高度爭議性。本研究結合會計領域與重大公共政策議題，藉支持/反對停建核四之觀點，辨認決策者或旁觀者所在乎之因素，如已投入興建成本、完工程度、預期尚需投入成本、經濟結果，以及核能安全、核電廠可能面臨的風險等，再探討決策者對停建核四的態度是否受沉沒成本或其他因素之影響；以及，決策者的知識背景（會計系、政治系）及政治理念是否影響決策者的上述認知。 本研究透過問卷搜集資料，發現已投入之核四興建成本（即沉沒成本）顯著影響受測者贊成停建與否的態度：當受測者越覺得沉沒成本重要，越傾向反對停建核四，有沉沒成本效果存在。本研究也發現受測者最重視的細項決策因素，為「臺灣若發生類似福島核災之核災損失」。對臺灣而言，日本福島本具時間、地理、文化上的相近性，再加上媒體敘事的渲染力，福島核災創造人民對於核災的公共想像及恐懼。在其他因素中，「非核家園」之理念亦顯著影響核四決策：當受測者越覺得「非核家園」之理念重要，越傾向贊成停建核四。本研究還發現受測者的知識背景影響其對沉沒成本的態度：評估核四決策時，政治系受測者較會計系受測者，易受沉沒成本影響，將沉沒成本視為攸關。再者，政治理念對於受測者贊成停建與否的影響，當受測者對候選人的偏好為江宜樺先生等人，其立場以反對核四停建居多；當受測者對候選人的偏好為蔡英文女士等人，其立場以贊成核四停建居多。 本研究發現臺電的溝通效果不佳，除數據之可靠性待加強外，傳遞的資訊並未針對問題的核心，例如核能安全很受受測者重視，然而臺電於說明核電廠之安全時，使用「複合式災害」來描述福島核災，複合來自地震及海嘯，而事實上福島核災係因海嘯而造成，海嘯因地震而引發；臺灣外海斷層之地理構造使臺灣核電廠區域縱有大地震時，也不會遭遇如福島般的大海嘯。 因此本研究建議臺電和行政機關，應針對核能安全及核能政策，辨認人民容易產生誤解的資訊為何，以正確的用詞精確澄清。此外，非會計背景之決策者，應加強關於沉沒成本的認知；會計背景之決策者，應抱持批判性思考的態度，質疑「沉沒成本不應該影響決策」的理論是否事實上成立。 / There is an undoubted rule always on accounting textbooks, which is, decision making should base on the cost-benefit analysis, as for the investment in the past belongs to sunk cost and is irrelevant to the decision making. However, is this legit? There are so many concerns beyond the cost when decision maker making decision, but most textbooks only focus on the cost, which is different from the turth. This research takes “the decision making in Lungmen nuclear power plant” as a case study, to examine whether sunk cost is truly irrelevant to decision making in reality. The issue that whether construction of Lungmen nuclear power plant should be suspended is highly controversial, no matter in aspect of energy, social policy and so on. First, across accounting and public issues, the researcher collects the viewpoints from pros and cons toward this issue, and identifies decision factors, for examples, the invested construction cost (which means sunk cost here), the degree of project completion, future expected cost, economic results, and nuclear safety, probable risk of nuclear power plant. Second, this research examines whether sunk cost or other factors influence decision makers’ attitude toward the issue. Third, whether knowledge backgrounds (major in accounting or political science) and political ideologies of decision makers influence their cognitions about sunk cost and the issue. By questionnaire survey, the results of this research found: First, sunk cost does influence receivers’ attitude toward the issue: the more important receivers think sunk cost is, the more they against the construction to be suspended; sunk cost effect has been observed. Second, among factors, receivers take Japan’s Fukushima nuclear disaster most serious. We argue that Fukushima nuclear disaster’s timing, geography and Japan’s culture are close to Taiwan, thus the disaster deepen public imagination and fear about the nuclear disaster. Third, the idea of nuclear-free homeland also influences decision making in the issue: the more important receivers think the idea is, the more they support the construction to be suspended. Forth, receivers’ knowledge backgrounds influence their cognitions about sunk cost: compared with receivers major in accounting, receivers major in political science take sunk cost as a relevant factor. Fifth, political ideologies of receivers influence their attitude toward the issue: when receivers vote to Jiang Yi-Huah etc., most of them against the construction to be suspended; when receivers vote to Tsai Ing-Wen etc., most of them support the construction to be suspended. Sixth, Taiwan Power Company doesn’t have effective communication with the public. Not only reliability of its data, but the covey of core information should be improved. For example, it describes Fukushima nuclear disaster as a “compound disaster” including earthquake and tsunami, but actually the disaster happened only due to tsunami caused by earthquake. The geological formation of offshore faults of Taiwan is different to Japan, thus there will be no Fukushima-like tsunami in Taiwan. This research suggest: First, Taiwan Power Company and executive administration should identify the information that public easily misunderstood, especially about nuclear safety and energy policy, and then clarify with proper words. Second, decision makers with non-accounting background should improve their knowledge about sunk cost. Third, decision makers with accounting background should have critical thinking about whether the theory that “sunk cost is irrelevant to decision-making” is found in reality.

In this thesis we look at one of the aging mechanisms that may have affected current aged Nuclear Power Plants (NPPs). Irradiation as an age-related degradation mechanism is studied for Reinforced Concrete (RC) in NPPs. This problem can be important for aged reactor buildings, radwaste buildings, spent nuclear fuel, research reactors, or accelerators that experience high levels of radiation close to existing thresholds. Mechanical properties of concrete are the most important parameters affected by radiation in NPPs. Compressive strength of concrete is reduced between 80 and 35 \% for radiation fluences between $2\times 10^{19}$ and $2\times 10^{21} n/cm^2$. Tensile strength reduction is more significant than compressive strength. It is reduced between 20 and 80 \% for a radiation fluence equal to $5\times 10^{19}$. We chose three radiation levels $2\times 10^{19}$, $2\times 10^{20}$, $2\times 10^{20}$ based on experimental results as the critical levels of radiation that RC structures in NPPs may be exposed to. Structures susceptible to the problem are mostly RC walls; so the RC panel is chosen as an appropriate representative scale element for the analysis. The effect of radiation on mechanical properties of concrete is considered to analyze degraded scale elements. Material properties, geometry, and loading scenarios of scale elements are selected to be close to actual quantities in existing nuclear power plant. Elements are analyzed under six types of loading combination of shear and axial loading conditions. A nonlinear finite element program, Membrane-2000, based on the Modified Compression Field Theory (MCFT) is used to solve scale elements numerically. Element behaviors are studied considering the factors influence ultimate strength capacity, failure mode, and structural ductility index of members. The results show that ultimate shear capacity of the elements subjected to combinations of shear and tension loading are reduced significantly for highly reinforced elements ($1.35<\rho<1.88$) in $2\times 10^{21} n/cm^2$ radiation. RC panels under shear-biaxial and uniaxial compression also show significant strength capacity reduction in radiation levels $2\times 10^{20} n/cm^2$ and $2\times 10^{21} n/cm^2$, respectively. Failure modes of the elements change from yielding of steel to shear failure by increasing level of degradation for the elements with reinforcement ratio between 0.9 and 1.88. Ductility of the RC panels is reduced significantly in the critical levels of radiation. Ductility of the elements became less than the allowable ductility value by increasing level of radiation.