• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.18 no.5
• /
• pp.53-59
• /
• 2019

Continuous ship unloaders (CSUs) are used for the uninterrupted transport of material in processing industries, power plants, and harbors in accordance with the stream rate of the material. This study analyzed the structural integrity and fatigue life of a CSU structure using finite element structural analysis in ANSYS APDL software. The stress varied greatly depending on the luffing angle and the slew angle of the boom conveyor. The structural integrity of the CSU girder was evaluated by applying ASME BPVC Section VIII Division 2. The fatigue cycle at the angle with the greatest stress difference was calculated. The fatigue cycle was calculated by applying the JIS B 8821:2013 fatigue curve. It was confirmed that the fatigue cycle of the CSU satisfies the allowable fatigue of 200,000 cycles.

• The KSFM Journal of Fluid Machinery
• /
• v.17 no.1
• /
• pp.5-11
• /
• 2014

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

• Plant Journal
• /
• v.14 no.2
• /
• pp.45-50
• /
• 2018

Nuclear safety law's amendment caused many problems to use radiography testing(RT). Phased array ultrasonic testing(PAUT) was adapted instead of RT for NDE of welding points in boiler tubes these days. Unfortunately, PAUT doesn't give us the discrimination characteristics about flaws distinction and flaws size clearly. In this thesis, the distinction characteristics of flaw types and the detection characteristics of flaw size using PAUT of welding points in boiler tubes were analyzed. It was concluded that PAUT can distinguish between planar flaws and rounded flaws, but it is hard to tell apart the types of flaw respectively. We paid attention to the discrimination of flaws size because PAUT tends to underestimate the flaw size of porosity and underestimate or overestimate the flaw size of porosity.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.39 no.4
• /
• pp.405-413
• /
• 2015

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.