• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.397-416
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• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part II, based on the verified finite element (FE) models of aircrafts Airbus A320 and A380, as well as the NPP containment and auxiliary buildings in Part I of this paper, the whole collision process is reproduced numerically by adopting the coupled missile-target interaction approach with the finite element code LS-DYNA. The impact induced damage of NPP plant under four impact locations of containment (cylinder, air intake, conical roof and PCS water tank) and two impact locations of auxiliary buildings (exterior wall and roof of spent fuel pool room) are evaluated. Furthermore, by considering the inner structures in the containment and raft foundation of NPP, the structural vibration analyses are conducted under two impact locations (middle height of cylinder, main control room in the auxiliary buildings). It indicates that, within the discussed scenarios, NPP structures can withstand the impact of both two aircrafts, while the functionality of internal equipment on higher floors will be affected to some extent under impact induced vibrations, and A380 aircraft will cause more serious structural damage and vibrations than A320 aircraft. The present work can provide helpful references to assess the safety of the structures and inner equipment of NPP plant under commercial aircraft impact.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.139-150
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• 2005

A robust control design procedure for a nuclear reactor has been developed and experimentally validated on the Penn State TRIGA research reactor. The utilization of the robust controller as a component of an autonomous control system is also demonstrated. Two methods of specifying a low order (fourth-order) nominal-plant model for a robust control design were evaluated: 1) by approximation based on the 'physics' of the process and 2) by an optimal Hankel approximation of a higher order plant model. The uncertainty between the nominal plant models and the higher order plant model is supplied as a specification to the ,u-synthesis robust control design procedure. Two methods of quantifying uncertainty were evaluated: 1) a combination of additive and multiplicative uncertainty and 2) multiplicative uncertainty alone. The conclusions are that the optimal Hankel approximation and a combination of additive and multiplicative uncertainty are the best approach to design robust control for this application. The results from nonlinear simulation testing and the physical experiments are consistent and thus help to confirm the correctness of the robust control design procedures and conclusions.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1661-1668
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• 2020

A physical protection system (PPS) is used for the protection of critical facilities. This paper proposes a structure analytic hierarchy approach (SAHA) for the hierarchical evaluation of the PPS effectiveness in critical infrastructure. SAHA is based on the traditional analysis methods "estimate of adversary sequence interruption, EASI". A community algorithm is used in the building of the SAHA model. SAHA is applied to cluster the associated protection elements for the topological design of complicated PPS with graphical vertexes equivalent to protection elements.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.235-246
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• 2014

The automatic seismic trip system (ASTS) continuously monitors PGA (peak ground acceleration) from the seismic wave, and automatically generates a trip signal. This work presents how the system can be designed by using a systems engineering approach under the given regulatory criteria. Overall design stages, from the needs analysis to design verification, have been executed under the defined processes and activities. Moreover, this work contributes two significant design areas for digitalized ASTS. These are firstly, how to categorize the ASTS if the ASTS has a backed up function of the manual reactor trip, and secondly, how to set the requirements using the given design practices either in overseas ASTS design or similar design. In addition, the methodology for determining the setpoint can be applied to the I&C design and development project which needs to justify the error sources correctly. The systematic approach that has been developed and realized in this work can be utilized in designing new I&C (instrument and control system) as well.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.1026-1031
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• 1998

A licensing approach for new fuel design is propose. It includes the possible licensing options to be taken according to the level of significance of design changes. This approach can be applied on legal grounds of the AEA, Article 44.3 and Article 21, but implies that the related ERAEAs should be revised or other alternative regulatory guidelines should be prepared. However, it is not intended to invoke additional requirements but rather to streamline and formulate the current practice by using one of supplementary provisions of the AEA, Article 104.2 which is recently enacted for approval of the technical report on a special topic.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.363-368
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• 2012

A feedwater controller targeted for an ABWR has been implemented using a modern field programmable gate array (FPGA), and verified using the full scope simulator at Taipower's Lungmen nuclear power station. The adopted control algorithm is a rule-based fuzzy logic. Point to point validation of the FPGA circuit board has been executed using a digital pattern generator. The simulation model of the simulator was employed for verification and validation of the controller design under various plant initial conditions. The transient response and the steady state tracking ability were evaluated and showed satisfactory results. The present work has demonstrated that the FPGA based approach incorporated with a rule-based fuzzy logic control algorithm is a flexible yet feasible approach for feedwater controller design in nuclear power plant applications.

• Journal of Information Technology Applications and Management
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• v.11 no.4
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• pp.35-47
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• 2004

The nuclear design analysis requires time-consuming and erroneous model-input preparation. code run. output analysis and quality assurance process. To reduce human effort and improve design quality and productivity. Innovative Design Processor (IDP) is being developed. Two basic principles of IDP are the document-oriented desigll and the web-based design. The document-oriented design is that. if the designer writes a design document called active document and feeds it to a special program. the final document with complete analysis. table and plots is made automatically. The active documents can be written with Microsoft Word or created automatically on the web. which is another framework of IDP. Using the proper mix-up of server side and client side programming under the LAMP (Linux/Apache/MySQL/PHP) environment. it e design process on the web is modeled as a design wizard style so that even a novice designer makes the design document easily. This automation using the IDP is now being implemented for all the reload design of Korea Standard Nuclear Power Plant (KSNP) type PWRs. The introduction of this process will allow large reduction in all reload design efforts of KSNP and provide a platform for design and R&D tasks of KNFC.

• Journal of the Korean Society of Systems Engineering
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• v.7 no.2
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• pp.1-5
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• 2011

The purpose of the Environment Design (ED) for Main Control Room (MCR) of Nuclear Power Plant (NPP) is to provide and create an optimal working space to be free from physical, physiological and mental stress as well as environmental discomfort, based on the previous environment design experiences and to recommend the best ED including the color, the lighting and the interior design. Environment Design consists of three main areas: Human factor engineering design, Interior design with color design, and Lighting design. These design areas have been interactively cooperated in a way that each design specialist would share the objectives and concepts of the Environment Design for MCR. The specialists for Human Factors Engineering design had a corporative role in such a way to provide the guidelines for MCR design suitability of Interior and Lighting design considering the Human System Interface (HSI) safety concerns. This paper describes fruitful efforts to create the best fit for MCR ED among several design proposals with the design recommendations, impacts, and contributions to NPP environment.