• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.951-956
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• 1995

원자력 발전소 2차측 배관에서 가장 심각한 문제로 대두되고 있는 침식부식현상을 예측/감시하는 시스템을 개발하는 연구를 수행하였다. 본 시스템은 침식부식 이론 및 관련 변수들로부터 단상 및 2상유체 배관의 침식부식률과 ASME Code의 허용기준에 따른 잔여수명을 예측할 수 있는 Prototype의 평가시스템 형태로 개발되었으며, 특정호기의 2차측 배관들 중에서 침식부식에 특히 민감한 T-형 및 L-형 배관을 선택한 후 시스템화시켜 검토하였다.

• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.444-450
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• 1997

The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops design load factors for the serviceability limit state of reinforced concrete containment structures. The target limit state probability is determined and the load factors are calculated by the numerical analysis. Design load factors are proposed and carried out the reliability assessments.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.188-194
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• 1996

현재 국내에서 관심이 고조되고 있는 원전 연장운전 방안의 일환으로, 주요 기기의 기술적 타당성 검토 대상기기 중 하나인 원자로 냉각재 배관 노즐의 피로수명 평가를 수행하였다. 본 노즐의 피로수명 평가는 원자로 냉각재 계통 노즐의 최종 설계문서에 제시된 응력해석 결과를 참조하여 ASME Code, Sec III의 절차에 따라 수행하였으며, 평가결과 이들 노즐의 설계수명 30년을 향후 40년 또는 그 이상 연장 운전할 경우에도 무리가 없는 것으로 판단된다.

• International Journal of Ocean Engineering and Technology Speciallssue:Selected Papers
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• v.5 no.1
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• pp.58-63
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• 2002

This paper presents a mechanical design of the deepwater pipeline wall thickness using the recent design rules. Characteristics and limitations of the new codes were identified through a case study design in the Gulf of Mexico. In addition to the ASME, API, and DVD codes, the code of federal regulations (CFR) was also utilized in the design. It was found that conservatism still exists within the collapse prediction for water depth greater than 1500 m. Comparision of the results from DNV and API codes were presented.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.689-696
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• 2006

For a perforated plate, it is challenging to develop a finite element model due to the necessity of the fine meshing of the plate, especially if it is submerged in fluid. This necessitates the use of a solid plate with equivalent material properties. Unfortunately, the effective elastic constants suggested by the ASME code are deemed not valid for a modal analysis. Therefore, in this study the equivalent material properties of a perforated plate are suggested by performing several finite element analyses with respect to the ligament efficiencies.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.4
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• pp.417-424
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• 2011

A fuel assembly consists of five major components, i.e., a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT); in addition, it also includes fuel rods (FRs). The TEP/BEP should satisfy stress intensity limits according to the conditions A and B of ASME, Section III, Division 1-Subsection NB. In a dual-cooled fuel assembly, the array and position of fuel rods are different from those in a conventional PWR fuel assembly; these changes are necessary for achieving power uprating. The flow plates of the TEP and BEP have to be modified accordingly. The pattern and shape of the flow holes were newly designed. To verify the strength compatibility, the Tresca stress limit according to the ASME code was investigated in the case of an axial load of 22.241 kN. In this paper, the stress linearization procedure for strength evaluation of a newly designed TEP is presented.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.809-814
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• 2016

A standard procedure for determining water-steam properties has been established through an international collaboration in addition to a domestic effort. The current accepted international standard for industrial application is based on the IAPWS-IF97 (International Association for the Properties of Water and Steam-Industrial Formation 97). Based on this standard, the ASME (American Society of Mechanical Engineers)/NIST (National Institute of Standard and Technology) developed the REPROP program in the USA, and the JSME (Japan Society of Mechanical Engineers) developed the steam table and calculation code. Upon applying this standard procedure, modified procedures were proposed for computational convenience, particularly in the supercritical pressure region where non-smooth variations of water-steam properties were distinctively observed. In this paper, the internationally adopted procedures and the progress of related activities are briefly summarized. Some practical considerations are presented for the efficient execution of computational code.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.16 no.12
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• pp.8753-8759
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• 2015

Establishment of analysis procedures and validated performance measurements for generator output is required to maintain stable management of generator output in turbine power generation cycle. We developed turbine expansion model and measurement validation model for the performance calculation of generator using turbine output based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). We also developed verification model for uncertain measurement data related to the turbine and generator output. Although the model in previous researches was developed using artificial neural network and kernel regression, the verification model in this paper was based on algorithms through Support Vector Machine (SVM) model to overcome the problems of unmeasured data. The selection procedures of related variables and data window for verification learning was also developed. The model reveals suitability in the estimation procss as the learning error was in the range of about 1%. The learning model can provide validated estimations for corrective performance analysis of turbine cycle output using the predictions of measurement data loss.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.5
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• pp.453-460
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• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.91-99
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• 2017

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radio-active materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B&PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.