• Journal of the Korean Society of Manufacturing Process Engineers
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• v.17 no.4
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• pp.150-159
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• 2018

The present research deals with a finite element analysis and fatigue evaluation of a steam separator of a high-pressure evaporator for the Heat Recovery Steam Generator (HRSG). The fatigue during the expected life of the HRSG was evaluated according to the ASME Boiler and Pressure Vessel Code Section VIII Division 2 (ASME Code). First, based on the eight transient operating conditions prescribed for the HRSG, temperature distribution of the steam separator was analyzed by a transient thermal analysis. Results of the thermal analysis were used as a thermal load for the structural analysis and used to determine the mean cycle temperature. Next, a structural analysis for the transient conditions was carried out with the thermal load, steam pressure, and nozzle load. The maximum stress location was found to be the riser nozzle bore, and hence fatigue was evaluated at that location, as per ASME Code. As a result, the cumulative usage factor was calculated as 0.00072 (much less than 1). In conclusion, the steam separator was found to be safe from fatigue failure during the expected life.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.654-659
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• 2003

PVAP (Pressure Vessel Analysis Program V1.0) was developed by adopting the finite element analysis program ANSYS V6.0, and Microsoft Visual Basic V6.0 was also utilized for the interfacing and handling of input and output data during the analysis. PVAP offers the end user the ability to design and analyze vessels in strict accordance with ASME Section VIII, Division 2. More importantly, the user is not required to make any design decisions during the input of the vessel. PVAP consists of three analysis modules for the finite element analysis of the primary components of pressure vessel such as head, shell, nozzle, and skirt. In each module, finite element analysis can be performed automatically only if the end user gives the dimension of the vessel. Furthermore, the calculated results are compared and evaluated in accordance with the criteria given in ASME Boiler and Pressure Vessel Code, Section VIII, Division 2. In particular, heat transfer analysis and consecutive thermal stress analysis for the junction between skirt and head can be carried out automatically in the skirt-tohead module. Finally, report including the above results is created automatically in Microsoft Word format.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• Journal of the KSME
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• v.20 no.6
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• pp.428-433
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• 1980

우리는 수차에 걸친 경제개발5개년 계획을 성공적으로 추진하여 오며 지금에는 공업구조의 고 도화 즉 중화학공업의 급속한 성장이 착실히 진전되어 나가고 있다. Plant산업은 기술축적이 기 반이 되어야 하는 것은 물론 양질의 기술축적 위에 완전한 품질보증체계를 갖추기 위하여는 전 사적 품질 보증체계를 갖추어야 한다. 이런 전사적 품질보증체계를 갖추기 위하여 미국기계기 술자협회(ASME)발생하는 ASME Boiler and Pressure Vessel Code에 의한 중공업부문및 원자력 발전소 부문의 설계 제조 시험에 대한 풍질보증장증(ASME Stamp)을 소개하고자 한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.101-106
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• 2016

NSSS (Nuclear Steam Supply System) and BOP (Balance of Plant) design works for PGSFR (Prototype Gen-IV Sodium Fast Reactor) have been conducted in Korea. NSSS major components, e.g. reactor vessel, steam generator and secondary sodium main pipes, are designed according to the rule of ASME boiler and pressure vessel code division 5, in which DBA (Design by Analysis) methods are used in the stress assessments. However, there is little discussions about detail rules for BOP piping design. In this paper, the detail methodologies of BOP piping stress assessment are discussed including safety systems and non-safety system pipings. It is confirmed that KEPIC MGE(ASME B31.1) and ASME BPV code division 5 HCB-3600 can be used in stress assessments of non-safety pipes and class B pipes, respectively. However, class A pipe design according to ASME BPV code division 5 HBB-3200 has many difficulties applying to PGSFR BOP design. Finally, future development plan for class A pipe stress assessment method is proposed in this paper.

• Journal of the Korean Society for Nondestructive Testing
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• v.4 no.2
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• pp.20-29
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• 1985

There are two basic concepts in industrial radiographic examination; one is a radio-graphic sensitivity, and the other is a acceptance criteria. The comparison of these main points are studied for KS Standard and ASME Boiler and Pressure Vessel Code. From the results of the experiment, higher radiographic sensitivity is required in KS Standard when the thickness of material to be examined is less than 20mm in single wall technique. The acceptance criteria for linear type indications are described on same concept in two standards, whereas the acceptance criteria for rounded indications of KS Standard which mainly depends upon the object thickness are more severe than those of ASME BPV Code.

• Journal of Advanced Marine Engineering and Technology
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• v.35 no.4
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• pp.443-450
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• 2011

Fuel gas compressor system is applied to medium FPSO. In order to avoid surge, this system used the anti-surge valves. When surge occurs it may lead to system's fracture. So anti-surge valves are evaluated structural strength and structural safety. Especially, in emergency mode, valves are must be guaranteed structural safety. In this study, structural strength and structural safety of anti-surge valve was evaluated using the numerical simulation. Unigraphics NX 4.0 was used as Geometrical models, structural strength and structural safety calculation were carried out by ANSYS Workbench 12.1. The ASME Boiler & Pressure Vessel Code is refer to allowable strength and safety factor of the valves.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.310-319
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• 1997

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary(B$_1$), Secondary(C$_1$) and Peak(K$_1$) stress indices for pressure, the Primary (B$_2$), Secondary(C$_2$) and Peak(K$_2$) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.17-26
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• 2011

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

• Journal of the KSME
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• v.33 no.8
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• pp.717-727
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• 1993

주요국의 보일러 및 압력용기의 기술기준 개발방향이 금속재료기술의 발달, 비파괴검사 기술의 개발, 용접기술의 급진전, 품질요건의 국제적인 규정의 제정 및 준수 등의 현상으로 볼 때 이를 대체적으로 수용하는 ASME Sec. III의 방향으로 통합되어 가는 느낌을 받고 있다. 따라서 우 리나라도 전담기구의 설립 또는 지정을 서둘러 장기적인 안목에서 체계적으로 대처하여야 급격히 변화하는 세계적인 기술흐름에 맞추어 우리의 관련산업이 지속적으로 발전 될 수 있으며 이와 관련된 기술개발방향이 바르게 갈 수 있을 것이다.