• Nuclear Engineering and Technology
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• 제12권4호
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• pp.255-266
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• 1980

핵연료봉의 동특성을 구명하였다. 지지그림과 연료봉의 모델을 작성하였고, 이들 모델을 실험을 통해 입증하였다. 유한요소모델과 SAPV전자계산 푸로그램을 사용하여 자연진동수와 모드를 구하였다. 핵연료봉의 n=1모드동특성을 쉽게 계제할 수 있는 간단한 보모델을 제시하였다. 보모델의 결과는 유한요소법에 의한것과 거의 비슷함을 보였다.

• Nuclear Engineering and Technology
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• 제29권2호
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• pp.148-157
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• 1997

A new capsule that has a unique structure in which the test environments including neutron flux and fluence, and irradiation temperature can be controlled precisely during irradiation, was conceptually designed. The capsule structure and instrumentation were successfully designed according to the JMTR's standard procedures of capsule design. Based on the target irradiation, the details of the irradiation such as neutron fluence and irradiation temperature ore calculated and the related capsule safety was evaluated. In addition, the effects of design parameters including the changes in inner-capsule configuration, heater capacity, and Helium gas pressure on the specimen temperature were analyzed with a computer program. Through these thermal and strength evaluations, this capsule was proved to be safe during the irradiation in the JMTR.

• 전기학회논문지
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• 제56권7호
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• pp.1224-1230
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• 2007

In this paper, we study the classification of defects at steam generator tube in nuclear power plant using eddy current testing (ECT). We consider 4 defect patterns of SG tube: I-In type, I-Out type, V-In type, and V-Out type. Through numerical analysis program based on finite element modeling, 400 ECT signals are generated by varying width and depth of each defect type. In order to improve the classification performance, we propose new feature extraction technique. After extracting new features from the generated ECT signals, multi-layer perceptron is used to classify the defect patterns. Through the computer simulation study, it is shown that the proposed method achieves 100% classification success rate while the previous method yields 91% success rate.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.56-64
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• 2020

The safety related heat exchangers have been evaluated for their performance during the operation of the nuclear power plant. The evaluation program for the safety related heat exchanger was developed in 2010 and used by KHNP based on EPRI TR-10739 algorithms. The spend fuel pool heat exchanger is one of the safety related heat exchanger in the nuclear power plant and also evaluated for their performance. Recently the performance evaluation for the spend fuel pool heat exchanger was not available because of the decreased heat in the spend fuel pool due to the long term overhaul. This paper analyzes the main cause of evaluation failure in the evaluation process and suggests the criteria for the heat exchanger performance evaluation during the long term overhaul.

• Nuclear Engineering and Technology
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• 제35권6호
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• pp.581-595
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• 2003

Transport lattice codes require the resonance integral tables for the resonant nuclides where the resonance integral is a function of the background cross section and can be prepared through a special program solving the slowing down equation. In case the cross section libraries do not include the resonance integral table for the resonant nuclides, the computational prediction produces a large error. We devised a new method using a Monte Carlo calculation for the effective resonance cross sections to solve this problem provisionally. We extended this method to obtain the resonance integral table for general purpose. The MCNP code is used for the effective resonance integrals and the LIBERTE code for the effective background cross sections. We modified the HELIOS library with the effective cross sections and the resonance integral tables obtained by the newly developed Monte Carlo method, and performed sample calculations using HELIOS and LIBERTE. The results showed that this method is very effective for the resonance treatment.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.66-82
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• 1999

The national long-term R&D program, updated in 1997, requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor(KALIMER), along with supporting R&D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R&D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of HAMMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 가을 학술발표회 논문집
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• pp.312-319
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• 2004

This paper is mai y focused to develop new concrete material model such as ultimate failure surface in compression-compression region, hardening rule and cracking criteria which are basically used in the nonlinear finite element analysis of nuclear prestressed concrete containment building. From the Kepri's experimental results, failure surface of the concrete based on the elasto-plastic material model is modified and new cracking criteria is proposed. Nonlinear FE analysis program using a new material model is implemented to analysis plane concrete. Finally, numerical simulation to compare the performance of the new material model with experimental results is employed. The numerical results by the proposed model in this study agree very well with the experimental data.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 춘계학술대회 논문집 전력기술부문
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• pp.10-13
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• 2001

이 연구에서는 2010년도 154kV, 34skV 및 765kV 송전 계통을 대상으로 전력 조류 계산 및 과도 안정도 해석 프로그램을 개발하여 새로 건설되는 대용량 원전 전력계통의 안전성을 평가하였다. 개발된 프로그램은 실계통 전체 송전망 및 전국 발전기 모선을 대상으로 전력 조류계산, 3상 단락 고장계산, 과도안정도 계산을 위하여 통합된 형태의 MS-Windows 기반 프로그램 패키지로 개발되었다. 프로그램의 효용성 검증을 위한 사례연구로 2010년의 송전계통에 대한 전력 조류를 계산하여 과도안정도를 해석한 결과 대용량 원전 전력계통은 극단적인 경우를 제외하고는 과도적으로 안정함을 확인하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 추계학술대회 논문집 학회본부 A
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• pp.26-28
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• 2000

국내 원전 소외 전력 계통과 실계통 대상과도 안정도 해석은 대형 해석프로그램인 PSS/E 등에 의존하여 해석해왔으나, 규제 관점에서 보수적인 과도 안정도를 해석할 수 있는 간이 프로그램을 개발하여, 규제 요건 적합성 여부를 확인할 수 있게 하였다. 본 프로그램은 원전 전력계통에 적용되는 품질요건을 적용하지 않는 제어기에 대하여 신뢰를 두지 않았으며, 송전망 보호 계통-보호계전기 및 차단기-에 의존하여 3상 평형 단락 고장 조건 과도 안정도를 해석한 결과, 과도적인 안정도 판별에 적합함을 확인하였다. 가혹한 송전망 조건 아래 원전 발전기의 고장제거 지연시간에 대한 민감도 분석을 통하여 취약한 원전 부지 및 원전 발전기를 확인할 수 있다.

• Journal of Mechanical Science and Technology
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• 제17권9호
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• pp.1297-1315
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• 2003

This study deals with the free vibration of two identical circular plates coupled with a bounded or unbounded fluid. An analytical method based on the finite Fourier-Bessel series expansion and Rayleigh-Ritz method is suggested. The proposed method is verified by the finite element analysis using commercial program with a good accuracy The case of bounded or unbounded fluid is studied for the effect on the vibration characteristics of two circular plates. Also, the effect of gap between the plates on the fluid-coupled natural frequencies is investigated.