• Journal of the Korean Institute of Intelligent Systems
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• v.15 no.6
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• pp.673-681
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• 2005

As it is the most important to make sure security and reliability for nuclear Power Plant, it's considered the most crucial issues to develop a fault detective and diagnostic system in spite of multiple hardware redundancy in itself. To develop an algorithm for a fault diagnosis in the nuclear PWR steam generator, this paper proposes a method based on ART2(adaptive resonance theory 2) neural network that senses and classifies troubles occurred in the system. The fault diagnosis system consists of fault detective part to sense occurred troubles, parameter estimation part to identify changed system parameters and fault classification part to understand types of troubles occurred. The fault classification part Is composed of a fault classifier that uses ART2 neural network. The Performance of the proposed fault diagnosis a18orithm was corroborated by applying in the steam generator.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.1-7
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• 2009

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU.

• Nuclear Engineering and Technology
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• v.12 no.4
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• pp.274-285
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• 1980

The TDCORE and the RELOAD-II codes are adopted in an attempt to establish a simplified computational system for the in-core fuel management decisions of a PWR. The TDCORE is being used to simulate the power and burnup behavior of the Gori unit 1 PWR during the fuel cycle 1 through the cycle 5. The validity of the TDCORE code is also presented by comparing the TDCORE prediction with the in-core measurements. The RELOAD-II code is used to determine the optimum fuel loading pattern which is one of the most important decisions of the Gori unit 1 reactor. The utility and applicability of two codes for the fuel management analyses are described.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.1-10
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• 2016

The number of nuclear power plants operating in Korea was 24 as of year 2015. Nine units out of 24 units have been operated for a period over 20 years. Kori unit 1 has been in operation for 40 years, and an extended operation for Wolsong unit 1 was decided in 2015. There has been reported some crackings in reactor internals in PWR have been reported in Europe, USA, Japan and Korea, and some of them were replaced with new one. Repair and replacement technologies for the reactor internals have been developing in order to meet the regulatory requirements for long term operation in Korea. The technologies will also be used for the exported nuclear units. It is required to review degradation history of the reactor internals worldwide as a part of the degradation management program development. Schematics of reactor internals designed and supplied by Westinghouse, Framatome and Combustion Engineering are described herein. Materials degradation history of reactor internals of PWR plants in USA, Japan and Europe is surveyed and summarized. Some events from Korean plants are also described. Aging management strategy for the internals is suggested.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• The Science & Technology
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• v.11 no.2 s.105
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• pp.43-44
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• 1978

한국원자력 연구소방사성오염처리실장 이상훈 박사팀(김관식, 송희열, 박상훈)은 최근 원자력 발전소(PWR)에서 생성되는 방사성폐액을 국내실정에 맞게 고화처리 할수 있는 새로운 공정과 함께 이를 위한 시험공장(Pilot plant)설계도 완료하여 앞으로 본격화된 원자력 발전소 가동에 크게 대비하게 되었다.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.147-154
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• 2006

Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modelling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code_Saturne, developped at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code_Aster). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mock-up, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.253-262
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• 2015

Radionuclide deposited on the surface of several internal and external systems in a nuclear power plant is created by the activation of corrosion products from nuclear reactor structural materials and fission products. Especially, the constant contact between water and the surface corrodes the inside where primary system makes coolants and corrosion products mixed. Also, these are circulated along the systems. For comparing models, CRUDTRAN, DISER, MIGA-RT and CPAIR codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor that are used at the stage of designing. The corrosion products behavior of domestic PWR primary system was predicted by using CRUDTRAN. This study aims to increase the reliability of corrosion product evaluation model by comparing the actual values and calculated values with the data of a Westing House-type Nuclear Power Plant.