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[en] In complex and high-risk work conditions, especially such as in nuclear power plants, human understanding of the plant is highly cognitive and thus largely dependent on the effectiveness of the man-machine interface system. In order to provide more effective and reliable operating conditions for future nuclear power plants, developing more credible and easy to use evaluation methods will afford great help in designing interface systems in a more efficient manner. In this study, in order to analyze the human-machine interactions, I propose the Human-processor Communication(HPC) model which is based on the information flow concept. It identifies the information flow around a human-processor. Information flow has two aspects: appearance and content. Based on the HPC model, I propose two kinds of measures for evaluating a user interface from the viewpoint of these two aspects of information flow. They measure the communicative complexity of each aspect. In this study, for the evaluation of the aspect of appearance, I propose three complexity measures: Operation Complexity, Transition Complexity, and Screen Complexity. Each one of these measures has its own physical meaning. Two experiments carried out in this work support the utility of these measures. The result of the quiz game experiment shows that as the complexity of task context increases, the usage of the interface system becomes more complex. The experimental results of the three example systems(digital view, LDP style view and hierarchy view) show the utility of the proposed complexity measures. In this study, for the evaluation of the aspect of content, I propose the degree of informational coincidence, R (K, P) as a measure for the usefulness of an alarm-processing system. It is designed to perform user-oriented evaluation based on the informational entropy concept. It will be especially useful inearly design phase because designers can estimate the usefulness of an alarm system by short calculations instead of costly operator-based tests. The result of the experiments for validating the proposed measure shows that as the problem contains larger R (K, P), the correctness of identifying process deviations becomes lower. I also propose a procedure based on the analytic hierarchy process (AHP) for evaluating alarm-processing systems with regard to integrating a series of deviations. It provides a relative ranking among alternative alarm systems by integrating the evaluation results on the usefulness of identifying deviations. The conventional AHP determines weights using a subjective pairwise comparison between each pair of criteria. In this study, I reduce the number of pairwise comparisons by introducing the measure, R (K, P). In order to show the validityof the proposed integrating procedure, I present an exemplary application. Generally, this AHP-based evaluation procedure successfully integrates a series of evaluation results for deviations and clearly shows relative ranks. More investigation including the enlargement on various deviations will provide a valuable guideline for the alarm system evaluation
[en] A core internal vibration monitoring system which is particularly concerned on the core support barrel (CSB) in the nuclear power plant reactor vessel is developed in this work. The core or fuel damage accidents can be caused by the loose-jointed flange between the top of the CSB and the head of the vessel. The loose-jointed flange can be detected with the internal vibration monitoring system, which has conventionally used the signals from ex-core neutron detectors. In order to improve the accuracy of the CSB monitoring system, however, the signals from the piezoelectric accelerometers are used in this work. This thesis consists of two parts; one is the development of a suitable tool for detecting the hold down spring broken accident or wearing out of the CSB, and the other the generation of vibration signals to represent the abnormal states of CSB. In this thesis, the adaptive resonance theory (ART; a type of neural network) is used to develop the monitoring system. The monitoring system using the Fuzzy ARTMAP processes the signals from the accelerometers. On the other hand, in order to get the data sets of the CSB in abnormal (loose-jointed) states, the finite element method (FEM) is used to model the CSB in various loose-jointed states. The target CSB is the one which is placed in ULJIN nuclear power plant unit 1. A mock-up CSB is constructed and experiments are carried out to prove that the FEM analyses properly simulate the CSB frequency responses in various states. The results show that the CSB FEM analyses and mock-up experiments are in good agreement
[en] This paper quantitatively presents the results of a case study which examines the fault tree analysis framework of the safety of digital systems. The case study is performed for the digital reactor protection system of nuclear power plants. The broader usage of digital equipment in nuclear power plants gives rise to the need for assessing safety and reliability because it plays an important role in proving the safety of a designed system in the nuclear industry. We quantitatively explain the relationship between the important characteristics of digital systems and the PSA result using mathematical expressions. We also demonstrate the effect of critical factors on the system safety by sensitivity study and the result which is quantified using the fault tree method shows that some factors remarkably affect the system safety. They are the common cause failure, the coverage of fault tolerant mechanisms and software failure probability
[en] This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.
[en] This paper proposes a methodology to assess and reduce risks on Spent Nuclear Fuel (SNF) packages from ship collisions with a probabilistic approach. A method to estimate the impact energy upon a collision was elaborated in consideration that whether ships slide or stuck to each other during collision. Structural crash-worthiness of the SNF ship was obtained through a non-linear finite element analysis. Cask damage probability was calculated based on frontal and side cask impact scenarios. This cask damage probability was investigated over various transport parameters which include the number of transport casks carried per shipment, cask stowage configuration, cask loading direction, and ship's cruise velocity. Collision case studies are presented by using reference models of the SNF package, the struck and striking ship. Most and least vulnerable stowage locations on the SNF ship were discovered. The number of casks in a shipment and their stowage method which suppressed cask damage probabilities were identified. The proposed methodology was successful in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • A probabilistic risk approach on spent fuel transportation is proposed. • Cask impact scenarios were frontal and sideways impact. • Impact energy is estimated on whether ships stuck or slide against each other. • Ship's strength is estimated by a non-linear Finite Element Analysis (FEA). • A risk reduction method is outlined by optimizing shipment load and its stowage.
[en] A demonstration facility termed PRIDE (PyRoprocess Integrated inactive DEmonstration facility) has been built to study and prepare for the construction of the active facility. Prior to that, a radiological impact assessment must be conducted to establish a safe and secure facility design. Research have been done to identify possible accident scenarios and their impact thereof to the surrounding environment. However these studies were based on the notion of internal accidents which extent was defined by the process characteristics. There has yet a study on externally induced radiological consequences, for example by malicious acts launched towards the facility. This paper attempts to close the gap by analyzing a certain malicious attack scenario and its radiological consequences. It may provide support for identification of vital areas in the facility and to achieve a security-by design objective. Based on the results, an attack launched on a single transport cask fully loaded with spent fuel as described in the scenario did not cause a radiological impact which exceeds regulatory limits. The limits were surpassed when there were two or more spent fuel casks involved. This result might be used as a basis not to aggregate loaded transport casks either during transport or during handling in the laydown area
[en] Spent fuel transportation of South Korea is to be conducted through near sea because it is able to ship a large amount of the spent fuel far from the public comparing to overland transportation. The maritime transportation is expected to be increased and its risk has to be assessed. For the risk assessment, this study utilizes the probabilistic safety assessment (PSA) method and the notions of the combined event. Risk assessment of maritime transportation of spent fuel is not well developed in comparison with overland transportation. For the assessment, first, the transportation scenario should be developed and categorized. Categories are assorted into the locations, release aspects and exposure aspects. This study deals with accident that happens on voyage and concentrated on ship-ship collision. The collision accident scenario is generated with event tree analysis. The scenario will be exploited for the maritime transportation risk model which includes consequence and accident probability
[en] Research have been done to identify possible accident scenarios and their impact thereof to the surrounding environment. However these studies were based on the assumption of a short term release period and a constant meteorological conditions. There has yet a study which examines the environmental impact from a prolonged release of radioactivity from a pyroprocessing facility. This paper attempts to close the gap by analyzing a certain prolonged release scenario and its radiological consequences. The results showed an agreement with the previous research which concluded that the radiological impact to the surrounding environment of a pyroprocessing facility is below the prescribed regulatory limits. Furthermore this study showed that the transport path might be dynamic enough to reflect escaping puffs back to the geographical region of interest. It also showed the prominent effect of the wind vector and velocity changes at the source release point to the final distribution of radioactive material concentration. An effect which should not be neglected particularly in a prolonged release period scenario
[en] Spent nuclear fuel (SNF) management has been an indispensable issue in South Korea. Before a long term SNF solution is implemented, there exists the need to distribute the spent fuel pool storage loads. Transportation of SNF assemblies from populated pools to vacant ones may preferably be done through the maritime mode since all nuclear power plants in South Korea are located at coastal sites. To determine its feasibility, it is necessary to assess risks of the maritime SNF transportation. This work proposes a methodology to assess the risk arising from ship collisions during the transportation of SNF by sea. Its scope is limited to the damage probability of SNF packages given a collision event. The effect of transport parameters' variation to the package damage probability was investigated to obtain insights into possible ways to minimize risks. A reference vessel and transport cask are given in a case study to illustrate the methodology's application.
[en] This approach may reduce maintenance labor's costs and system unavailability. Experiment results confirmed that the proposed methodology successfully meet its designated objectives. The Condition Based Maintenance (CBM) approach has gained recognition for resolving this issue. The CBM attempts to perform maintenance only when it is required as suggested by the component's latest condition. This method implies the need for additional monitoring instrumentation costs. Furthermore, the CBM may require sudden, unplanned maintenance events and therefore the availability of a maintenance team at any time. This may be more costly compared to fixed maintenance schedules because it requires the maintenance team and equipment to be always available at any time. This paper proposes an integrated method to optimize the CBM by using prognostics methodology to manage the operation of an electromechanical component. The optimization is expected to prolong the component's service life and minimize CBM costs. A fault growth model was presented taking into account aleatoric and epistemic uncertainties. This model was coupled with an MPC to preserve the degrading component's performance within desired specification. The controller's actions were further optimized to coincide the under-performance-time with the planned maintenance schedule