• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.795-801
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• 1998

Metallic fatigue of Pressurized Water Reactor(PWR) materials is a generic safety issue for commercial nuclear power plants. It is very important to obtain the fatigue usage factor for component integrity and life extension. In this paper, fatigue usage was obtained at the inside surface of Kori unit 2, 3 and 4 RCP casing weld, based on the design transient. And it was intended to establish the procedure and the detailed method of fatigue evaluation in accordance with ASME Section III Code. According to this code rule, two methods to determine the stress cycle and the number of cycles could be applied. One method is the superposition of cycles of various design transients and the other is based on the assumption that a stress cycle correspond to only one design transient. Both method showed almost same fatigue usage in the RCP casing weld.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.153-158
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• 2000

Nuclear reactor pressure vessel should be safety even in the case that hypothetical defects with allowable size are in vessel. Therefore, the materials should have excellent fracture resistance characteristics. The purpose of this study is to analyze the results of preirradiation baseline test of nuclear pressure vessel for Yonggwang Unit 5/6. In experiments, drop weight tests and impact tests are carried out to obtain nil-ductility transition reference temperature, $RT_{NDT}$ and static and dynamic fracture toughness tests are performed to compare with $K_{IR}$ curve in accordance with ASME Sec.III. The test results show that the materials had sufficiently fracture resistance characteristics for 40 years of design life.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.4
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• pp.483-491
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• 2016

Precision of theoretical group velocity of waves in shell structures was discussed for the purpose of source localization of loose parts impact in pressure vessels of nuclear power plants. Estimating exact location of loose parts impact inside a reactor or a steam generator is very important in safety management of a NPP. Evaluation of correct propagation velocity of impact signals in pressure vessels, most of which are shell structures, is essential in impact source localization. Theoretical group velocities of impact signals in a plate and a shell were calculated by wave equations and compared to the velocities measured experimentally in a plate specimen and a scale model of a nuclear reactor. The wave equation applicable to source localization algorithm in shell structures was chosen by the study.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.212-218
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• 2011

APR1400 RVI CVAP using the non-prototype category is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during preoperational and initial startup testing and to detemine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the IBA and the UGS were established to 0.03 in and 0.01 in, respectively.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.11
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• pp.1016-1021
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• 2008

The development of the first wall whose major function is to withstand high neutron and heat fluxes is a critical path to fusion power. The materials database and the fabrication technology are being developed for design, construction and safety operation of the fusion reactor. The first wall was designed to consist of the plasma facing armor, the heat sink layer and the supporting plates. and Porous materials are of significant interest due to their wide applications in catalysis, separation, lightweight structural materials. In this study, the characteristics of the sintering process of SiC ceramic, $SiC_f$/SiC composite and porous $C_f$/SiC composite have been introduced order to study of the fusion blanket materials and heat-exchange pannel.

• Proceedings of the KWS Conference
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• 2003.05a
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• pp.216-218
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• 2003

A fracture mechanics analysis methodology for stress corrosion cracks (SCCs) existing in the Alloy 600 nozzle weld joint for control rod drive mechanisms (CRDMs) of pressurized water reactor is studied. Effects of weld residual stresses on the sub-critical crack behavior during the reactor operation are investigated by a fracture mechanics analysis, which is combined with the finite element alternating method. It is found that effects f the residual stresses on the stress intensity factor (SIF) and crack growth rate (CGR) are dominant and values of SIF and CGR of cracks in the region of weld joint are increased by a factor of three or more on an average.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.58-67
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• 1999

The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsong 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using Statechart-based Formalism and Stalemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language though we manually made Software Requirement Specification(SRS) for safety-critical software using statechart-based formalism. Most of the phases of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering (CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.551-562
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• 1996

This paper introduces the five-step methodology for identifying information needs and assessing instrument availability during the course of severe accidents in nuclear power plants. The methodology is applied to the Yonggwang (YGN) 3&4 to shed light on accident management. It constructs three safety objective trees to prevent the reactor vessel failure, to prevent the containment failure, and to mitigate the fission product release from the containment. The study assesses information needs and instrument availability under severe conditions for preventing the reactor vessel failure of YGN 3&4, and recommends additional instrument that m8y prove to be of vital importance in managing the accident.

• International Journal of Air-Conditioning and Refrigeration
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• v.9 no.2
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• pp.77-84
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• 2001

Experimental and computational studies were carried out to investigate the turbulent heat transfer enhancement of the cooling system in nuclear reactor by large scale vortex generation. The large scale vortex motion was generated by rearranging the inclination angels of mixing vanes to the coordinate direction. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity concept based on $\kappa{-}\varepsilon$ model was employed to calculate the turbulent heat and momentum transfers in the subchannel. The turbulences generated by split mixing vanes has small length scales so that they maintain only about $10D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex motions continue longer and remain up to $25D_H$ after the spacer grid.