• Journal of the Korean Professional Engineers Association
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• v.34 no.2
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• pp.14-18
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• 2001

The primary cooling piping of HANARO is classified as safety class 3, seismic class 1 and quality class Q. This piping as safety related feature has been designed, manufactured and tested in accordance with ASME SEC. Ⅲ, DIV 1, Class 3. In October of 2000, the first step of the in-service inspection for this piping was carried out in accordance with ASME SEC. XI. This describes the results of the Inspection including the preparation of inservice inspection plan and inspection method. It is verified through the results that the safety related piping is maintained the mechanical and structural Integrities.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.316-321
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• 2001

A previous feasibility study for the Korean lead plant, PLiM Phase I, showed a strong possibility of continued operation beyond the original licensed period. In 1998, PLiM Phase II study was initiated aimed at performing additional detailed evaluations on a wider range of components. The objective of this paper is to present the Korean PLiM efforts for Class 1 piping which is identified as one of the critical components with regard to long-term operation. The key findings such as typical design features, degradation mechanisms, technical issues, draft results from the lifetime evaluation for Class 1 piping of the lead plant are briefly described.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.1-10
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• 2022

Currently, there is no technical standard and regulation for seismic analysis of non-nuclear safety piping. Accordingly, ASME Sec.III ND, a standards applied to safety class 3 piping, is applied. However, the technical standard applied for other than seismic analysis is ASME B31, which leads to controversy. In this study, the feasibility of applying ASME B31E was confirmed by reviewing rulescomparing technical standards, and evaluating piping allowable stress margins. The evaluation revealed that applying ASME B31.1 as a technical standard is too conservative compared to ASME Sec.III ND. On the other hand, ASME B31E (issued at the request of the industry) clearly presents the technical standards for seismic analysis of ASME B31 piping, and shows a similar level of conservatism compared to ASME Sec.III ND. It is expected to reduce the controversy over technical standards for seismic analysis of non-nuclear safety piping by applying ASME B31E.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.11-17
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• 2015

High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.69-76
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• 2018

The paper presents preliminary investigation results for the effect of the baseline correction in the acceleration excitation method on finite element seismic analysis results (such as accumulated equivalent plastic strain, equivalent plastic strain considering cyclic plasticity, von Mises effective stress, etc) of nuclear safety Class I components. For investigation, finite element elastic-plastic time-history seismic analysis is performed for a surge line including a pressurizer lower head, a pressurizer surge nozzle, a surge piping, and a hot leg surge nozzle using the Chaboche hardening model. Analysis is performed for various seismic loading methods such as acceleration excitation methods with and without the baseline correction, and a displacement excitation method. Comparing finite element analysis results, the effect of the baseline correction is investigated. As a result of the investigation, it is identified that finite element analysis results using the three methods do not show significant difference.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• Journal of Welding and Joining
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• v.31 no.6
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• pp.1-7
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• 2013

In nuclear power plants, lined carbon steel pipes or PCCPs (pre-stressed concrete cylinder pipes) have been widely used for sea water transport systems. However, de-bonding of linings and oxidation of PCCP could make problems in aged NPPs (nuclear power plants). Recently at several NPPs in the United States, the PCCPs or lined carbon steel pipes of the sea water or raw water system have been replaced with HDPE (high density polyethylene) pipes, which have outstanding resistance to oxidation and seismic loading. ASME B&PV Code committee developed Code Case N-755, which describes rules for the construction of buried Safety Class 3 polyethylene pressure piping systems. Although US NRC permitted HDPE materials for Class 3 buried piping, their permission was limited to only 10-year operation because of several concerns including the quality of fusion zone of HDPE. In this study, various requirements for fusion qualification test of HDPE and some regulatory issues raised during HDPE application review in foreign NPPs are introduced.

• Progress in Superconductivity and Cryogenics
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• v.24 no.4
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• pp.65-70
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• 2022

A vacuum system is designed for thermal insulation of a 10 ton/day class air liquefaction cold box for liquid air energy storage. The vacuum system is composed of a turbomolecular pump, a backing pump and vacuum piping for the vacuum pumps. The turbomolecular pump is in combination with the backing pump for pumping capacity. The vacuum piping is designed with system installation conditions, such as distance from the cold box, connections to vacuum pumps and installation space. The capacity of the vacuum pump combination, namely pumping speed, is determined by analysis of the vacuum system, and pump-down time to 1×10^-5 mbar is estimated. Vacuum piping conductance, system pumping speed and outgassing rate are calculated for the pump-down time with the ultimate pumping speed range of the vacuum pump combination of 1400 - 2300 l/s. Although the pump-down time gets shorter by larger capacity vacuum pumps, it mainly depends on target vacuum degree and outgassing rate in the cold box. The pump-down time is estimated as 3 - 6 hours appropriate for cold box operation for the pumping speed range. Considering the outgassing rate has uncertainty, the vacuum pump combination with pumping speed of 1900 l/s is chosen for the vacuum system, which is middle value of the pumping speed range.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.