• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.126-136
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• 2021

During operation of nuclear power plants, the fatigue assessment should be conducted repeatedly, considering changes of operating environments. For the case that cumulative usage factors (CUFs) may exceed the acceptance limit, flaw tolerance evaluation can be an alternative method to meet the regulatory requirements. In this respect, this paper analyzes the effects of the input variables for flaw tolerance evaluation based on ASME BPVC Section XI Appendix L. The reference analysis is performed for the example problem in NUREG/CR-6934. Then effects of the crack orientation, stress intensity factor solutions, thermal stress profiles, fatigue stress decomposition and fatigue crack growth curves are considered for the sensitivity analysis. The results show that the stress analysis considering the actual environment plays a crucial role in flaw tolerance evaluation.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1320-1329
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• 2024

When wall-thinning occurs in nuclear Class 2 and 3 pipes, reinforcement is typically applied rather than replacement. To analyze the structural integrity of reinforced wall-thinned pipe, stress analysis results using full 3-D FE analysis are not compatible to the design code equation, ASME BPVC Sec. III NC/ND-3650. Therefore, the efficient stress evaluation method for the reinforced wall-thinned pipe, compatible to the design code equation, needs to be developed. In this paper, stress evaluation methods for the reinforced wall-thinned pipe are proposed using the equivalent straight pipe concept. Furthermore, for fatigue analysis of the reinforced wall-thinned pipe, the stress intensification factor of reinforced wall-thinned pipe is presented using the structural stress method given in ASME BPVC Sec. VIII Div.2.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1616-1629
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• 2023

To quantify the conservatism of existing ASME strain-based evaluation methods for seismic loading, this paper presents very low cycle fatigue test data of elbows under various cyclic loading conditions and comparison of evaluation results with experimental failure cycles. For strain-based evaluation methods, the method presented in ASME BPVC CC N-900 and Sec. VIII are used. Predicted failure cycles are compared with experimental failure cycle to quantify the conservatism of evaluation methods. All methods give very conservative failure cycles. The CC N-900 method is the most conservative and prediction results are only ~0.5% of experimental data. For Sec. VIII method, the use of the option using code tensile properties gives ~3% of experimental data, and the use of the material-specific reduction of area can reduce conservatism but still gives ~15% of experimental data.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.54-60
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• 2022

ASME BPVC provides stress evaluation rules for Class 2 and 3 nuclear piping. However, such rules are difficult to be applied to reinforced wall-thinned pipes during service. To resolve this issue, a new method for stress evaluation of reinforced wall-thinned pipes is proposed in this work, based on the equivalent stiffness concept. By converting a reinforced wall-thinned pipe to an equivalent straight pipe having the same stiffness, stress evaluation can be proceeded using the current ASME BPVC rules. The proposed method is applied to pipes with 4 different normal pipe size and the effects of reinforcement and wall-thinning dimensions on evaluated stresses are discussed.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.206-210
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• 2020

Super fine-grained graphite is a type of advanced nuclear graphite which was developed for Molten Salt Reactor (MSR). It is necessary to establish a failure assessment method used for nuclear graphite components in MSR. A modified assessment approach based on ASME BPVC-III-5_2017 is presented. The new approach takes a new parameter, KIC, into account and abandons the parameter, grain size, which is unrealistic for super fine-grained graphite as the computation is enormous if we use conventional methods. Three methodologies (KTA 3232, ASME, New approach) were also evaluated by theoretical prediction and experimental verification. The results indicated the new developed code can be used for design and failure assessment of super fine-graphite components and has more extensive applicability.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.401-413
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• 2022

This paper provides a simple method by which to estimate the cross-section stress profiles for nozzles designed according to ASME Code Section III. Further, this method validates the effectiveness of earlier work performed by the authors on standard nozzles. The method requires only the geometric information of the pressure vessel and the attached nozzle. A PWR direct vessel injection nozzle, a PWR outlet nozzle, a PWR inlet nozzle and a BWR recirculation outlet nozzle are selected based on their corresponding specific designs, e.g., a varying nozzle radius, a varying nozzle thickness and an outlet nozzle boss. A cross-section stress profile comparison shows that the estimates are in good agreement with the finite element analysis results. Differences in stress intensity factors calculated in accordance with ASME BPVC Section XI Appendix G are discussed. In addition, a change in the dimensions of an alternate nozzle design relative to the standard values is discussed, focusing on the stress concentration factors of the nozzle inside corner.

• 대한조선학회논문집
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• 제53권2호
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• pp.144-153
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• 2016

This paper aims to demonstrate the design, structure analysis, and hydrostatic pressure test of the cylinder used in 2000m water depth. The cylinder was designed in accordance with ASME pressure vessel design rule. The 1.5 times safety factor required by the general rule was applied to the design of the cylinder, because ASME rule is so excessive that it is not proper to apply to the hydrostatic pressure test. The finite element analysis was conducted for the cylinder. The cylinder was produced according to the design. The hydrostatic pressure test was conducted at the hyperbaric chamber in KRISO. The results of finite element analysis(FEM) and those of the hydrostatic pressure test were almost the same, which showed that the design was exact and reliable.

• 한국전산구조공학회논문집
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• 제30권3호
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• pp.207-214
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• 2017

국내의 72.5kV 이상, 주파수 60Hz의 송배전설비인 옥내 및 옥외용 가스절연개폐장치(GIS)는 내진 안전성에 대해 국가에서 정한 한전표준규격(ES-6110-0002)을 만족해야 한다. 이 규격에서 명시되지 않은 사항은 IEC 62271-203, 62271-207 등의 관련 기기 규격에 준한다. 한전표준규격에서 기기는 정상사용상태와 특수사용상태에서 건전성이 유지되어야 한다. 안전성 판단을 위해 ASME BPVC SEC.VIII 내압용기 설계 기준에 의해 A6061-T6 재질의 GIS에 대한 정상사용상태 기준과 국내 한전표준규격과 국외 IEC 62271-207에 의한 특수사용상태 기준(지진)에 대한 총체적 응력상태를 판단하였다. 한전표준규격 기준(0.22g) 적용시, 최종응력이 알루미늄인 Part A는 78.2MPa, Part D2의 경우 102.3MPa로, ASME 허용응력 값 181.5MPa를 만족하고 있다. IEC 62271-207 High 0.5g의 경우에도 최종응력은 Part A는 90.5MPa, Part D2는 103.8MPa이다. 본 연구 결과, 72.5kV GIS는 한전표준규격의 구조안전성과 내진성능을 충분히 만족함을 보이고 있다. 내진해석으로 내진시험을 수행할 수 없는 대형 전력기기의 내진성능 실증에 활용될 수 있을 것으로 기대된다.

• 한국해양공학회지
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• 제30권3호
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• pp.201-207
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• 2016

In this paper, an aluminum pressure vessel (cylinder) for a 200 m water depth is designed and analyzed. Because of their lack of usage in the deep sea, only a few papers about pressure vessels subjected to external pressures have previously been published. Moreover, the high level of imported external-pressure-vessel products limits the academic pursuit. Yet, research on internal pressure vessels is widely available because of their broad usage at onshore. This paper presents the process of basic designing and modelling of pressure vessels using the design rules of American Standard of Mechanical Engineering (ASME) Section VIII Division 1. To promote understanding, finite element analysis (FEA) result of an existing sample cylinder which was not designed by ASME code is compared with the design obtained in this paper. Several methodologies are used for the finite element analysis, including rectangular, cylindrical, and axisymmetric coordinate, to attain an accurate stress result. Same dimensions except the thickness of the cylinder and loading condition of 0.200 MPa was given for the current study. Finally, a rigorous design procedure is added for the bolt and boundary conditions of the cylindrical body and its ends. The obtained stress level satisfies the allowable design stress value specified in the ASME code.

• 시스템엔지니어링학술지
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• 제13권1호
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• pp.15-23
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• 2017

This paper describes verification of APR1400 reactor vessel by applying the system engineering approach, in which the design re-engineering method is used to check the design parameters of APR1400 RV (reactor vessel). The RV is classified as safety class 1 and therefore must adhere strictly to the rules of ASME BPVC section III, subsection NB and seismic category I. This study explores designing the RV by following the ASME guidelines and making a comparative study with the current design. To meet this objective we apply system engineering methodologies to structure the process and allow for verification and validation of the major RV design parameters such as thickness of RV. The structural thicknesses of various part of RV are determined as well as reinforcements on the RV major nozzles. A 3D virtual reality model was created based on the design parameters using CATIA V5 and animation using Dassault Composer V2016. A comparison of re-engineered ARP1400 RV and standard APR1400 RV was done to show which design parameters were taken more conservative approach.