• 기술사
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• 제44권6호
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• pp.40-44
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• 2011

Nuclear energy in Korea began in 1958, when the Korea's atomic energy act was formulated and the relevant organizations were founded. Since then, notwithstanding the two catastrophe like TMI and Chernobyl accident, Korea made a wise decision to expand the peaceful uses of the nuclear energy as well as to localize the essential nuclear design technology of fuel and nuclear steam supply system. This decision resulted in the success of export of nuclear power plants as well as research reactor in 2010s. The Korea's nuclear policy, which well utilized 'international crisis in nuclear business' as 'opportunity of Korea to get. nuclear technology', is believed nice policy as a role model of nuclear new-comer countries. Based upon the success story of localization of nuclear technology, Korea had an eye for a niche market, which was a basis of development of SMART, Korea-made integral PWR. The operation of a SMART plant can sufficiently provide not only electricity but also fresh water for 100,000 residents. Last two years, Korea's nuclear industry team led by the Korea Atomic Energy Research Institute completed the standard design of SMART and applied to the Korea's regulatory body for standard design approval. Now the Korea's licensing authority is reviewing the design with the relevant documents, and the design team is doing its best to realize its hope to get the approval by the end of this year. From next year, the SMART business including construction and export will be explored by the KEPCO consortium.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.784-791
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• 2019

In the present study, a series of tungsten austenitic stainless steel alloys have been developed by interchanging the molybdenum in standard SS316 by tungsten. This was done to minimize the long-life residual activation occurred in molybdenum and nickel after decommissioning of the power plant. The microstructure and mechanical properties of the prepared alloys are determined. For the sake of increasing multifunction property of such series of tungsten-based austenitic stainless steel alloys, gamma shielding properties were studied experimentally by means of NaI(Tl) detector and theoretically calculated by using the XCOM program. Moreover, fast neutrons macroscopic removal cross-section been calculated. The obtained combined mechanical, structural and shielding properties indicated that the modified austenitic stainless steel sample containing 1.79% tungsten and 0.64% molybdenum has preferable properties among all other investigated samples in comparison with the standard SS316. These properties nominate this new composition in several nuclear application domains such as, nuclear shielding domain.

• 한국지진공학회논문집
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• 제24권3호
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• pp.123-128
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• 2020

Analysis of the 2016 Gyeongju earthquake and the 2017 Pohang earthquake showed the characteristics of a typical high-frequency earthquake with many high-frequency components, short time strong motion duration, and large peak ground acceleration relative to the magnitude of the earthquake. Domestic nuclear power plants were designed and evaluated based on NRC's Regulatory Guide 1.60 design response spectrum, which had a great deal of energy in the low-frequency range. Therefore, nuclear power plants should carry out seismic verification and seismic performance evaluation of systems, structures, and components by reflecting the domestic characteristics of earthquakes. In this study, high-frequency amplification factors that can be used for seismic verification and seismic performance evaluation of nuclear power plant systems, structures, and equipment were analyzed. In order to analyze the high-frequency amplification factor, five sets of seismic time history were generated, which were matched with the uniform hazard response spectrum to reflect the characteristics of domestic earthquake motion. The nuclear power plant was subjected to seismic analysis for the construction of the Korean standard nuclear power plant, OPR1000, which is a reactor building, an auxiliary building assembly, a component cooling water heat exchanger building, and an essential service water building. Based on the results of the seismic analysis, a high-frequency amplification factor was derived upon the calculation of the floor response spectrum of the important locations of nuclear power plants. The high-frequency amplification factor can be effectively used for the seismic verification and seismic performance evaluation of electric equipment which are sensitive to high-frequency earthquakes.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.989-994
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• 2009

We at Mitsubishi have lined up Gen-III+ solutions for a wide variety of global customers: ATMEA1 of the 1100MWe class, and an APWR with the largest capacity of 1700MWe. In this paper, we would like to introduce the APWR. With an increased requirement for nuclear power generation as an effective countermeasure against global warming, we have established the APWR plant, a large-capacity Mitsubishi standard reactor combining our accumulated experience and technology as an integrated PWR plant supplier. The APWR plant has achieved high reliability, safety and enhanced economy based on a technology that has been developed with the support of the government and utilities through improvement and standardization programs of light water reactors. Currently, Tsuruga Units 3 and 4, the first two APWRs, are undergoing licensing, while we are making efforts to obtain the standard design certification (DC) of US-APWR and preparing for the European Utility Requirements (EUR) compliance assessment of EU-APWR. Mitsubishi Heavy Industries, Ltd. (MHI) positions the APWR as a core technology that will contribute to the prevention of global warming and meet worldwide requirements.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1231-1242
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• 2020

The Kori Unit 1 will be decommissioned after a permanent shutdown in June 2017. South Korea has a 0.1 mSv/yr exposure limit standard for limited or unlimited site release. This is South Korea's first commercial NPP; therefore, if the containment building is reused as a memorial hall, it will contribute to the improvement of public understanding and enhance the public's acceptance of NPPs. Also, existing Kori Unit 1 nuclear power plant manpower resources can be reused after decommissioning and resident staff and memorial hall visitors can activate nearby commercial areas. Therefore, such a reuse scenario may also prevent an economic recession. The exposure dose was calculated using the following scenarios: worker in the containment building, visitor in the containment building, and worker in buildings other than the containment building. The exposure dose in the buildings was calculated by the RESRAD-BUILD developed by the Argonne National Laboratory (ANL). The preliminary exposure dose and derived concentration guideline level (DCGL) were derived.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.590-599
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• 2010

Recently, efficient operation and practical management of power plants have become important issues in the nuclear industry. In particular, typical aging parameters such as stress and cumulative usage factor should be determined accurately for continued operation of a nuclear power plant beyond design life. However, most of the major components have been designed via conservative codes based on a 2-D concept, which do not take into account exact boundary conditions and asymmetric geometries. The present paper aims to suggest an effective fatigue evaluation methodology that uses a prototype of the integrated model and its transfer functions. The validity of the integrated 3-D Finite Element (FE) model was proven by comparing the analysis results of individual FE models. Also, mechanical and thermal transfer functions, known as Green's functions, were developed for the integrated model with the standard step input. Finally, the stresses estimated from the transfer functions were compared with those obtained from detailed 3-D FE analyses results at critical locations of the major components. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system, and this combination, will enhance the continued operations of old nuclear power plants.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2172-2194
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• 2023

A variational nodal method (VNM) with unstructured-mesh is presented for solving steady-state and dynamic neutron diffusion equations. Orthogonal polynomials are employed for spatial discretization, and the stiffness confinement method (SCM) is implemented for temporal discretization. Coordinate transformation relations are derived to map unstructured triangular nodes to a standard node. Methods for constructing triangular prism space trial functions and identifying unique nodes are elaborated. Additionally, the partitioned matrix (PM) and generalized partitioned matrix (GPM) methods are proposed to accelerate the within-group and power iterations. Neutron diffusion problems with different fuel assembly geometries validate the method. With less than 5 pcm eigenvalue (k_eff) error and 1% relative power error, the accuracy is comparable to reference methods. In addition, a test case based on the kilowatt heat pipe reactor, KRUSTY, is created, simulated, and evaluated to illustrate the method's precision and geometrical flexibility. The Dodds problem with a step transient perturbation proves that the SCM allows for sufficiently accurate power predictions even with a large time-step of approximately 0.1 s. In addition, combining the PM and GPM results in a speedup ratio of 2-3.

• 한국시뮬레이션학회:학술대회논문집
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• 한국시뮬레이션학회 1998년도 The Korea Society for Simulation 98 춘계학술대회 논문집
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• pp.86-90
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• 1998

The safety and reliability of nuclear power plant operations relies heavily on the plant operators ability to respond to various emergency situations. It has become standard industry practice to utilize simulators to improve the safety and reliability of nuclear power plants operations. The simulators built for Younggwang#3,4, which is the basic model of the Korean Nuclear Power Plant design, has been developed precisely for this purpose. Dynamic Model and Display Model are developed under US3(UNIX Simulation Software Support System) environment in simulator for Younggwang#3,4. Since these two models are developed under each own operating system, it is necessary to develop a method for transporting data between these two systems. This paper descirves communication environment between Dynamic Model and Display Model, and addresses a file generation method for the Display Model, which will be necessary for designing MMI of MCR(Main Control Room) in the furture.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.575-584
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• 2006

As the development of total management systems for sites along with site environmental information is becoming standard, the system known as the Site Information and Total Environmental database management System (SITES) has been developed over the last two years. The first result was a database management system for storing data obtained from facilities, and a site characterization in addition to an environmental assessment of a site. The SITES database is designed to be effective and practical for use with facility management and safety assessment in relation to Geographic Information Systems. SITES is a total management program, which includes its database, its data analysis system required for site characterization, a safety assessment modeling system and an environment monitoring system. It can contribute to the institutional management of the facility and to its safety reassessment. SITES is composed of two main modules: the SITES Database module (SDM) and the Monitoring & Assessment (M&A) module [1]. The M&A module is subdivided into two sub-modules: the Safety Assessment System (SAS) and the Site Environmental Monitoring System (SEMS). SAS controls the data (input and output) from the SITES DB for the site safety assessment, whereas SEMS controls the data obtained from the records of the measuring sensors and facilities. The on-line site and environmental monitoring data is managed in SEMS. The present paper introduces the procedure and function of the M&A modules.