• Corrosion Science and Technology
• /
• v.3 no.5
• /
• pp.216-221
• /
• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

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

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.19 no.1
• /
• pp.27-35
• /
• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.6 no.1
• /
• pp.65-71
• /
• 2010

The plugging rate of reheater tubes of Wolsung unit 1 nuclear power plant has been increased by corrosion and erosion since 1990. As the dimensions of the new first stage reheater bundle tubes which were supplied by Hanjung company to replace were different from old one, numerical calculations are carried out for flow and heat transfer in the reheater bundle tubes of the N.P.P. Numerical calculations consists of thermal performance, drain line pressure drop, flow change by pressure drop of line, stress analysis of finned tubes and analysis of flow induced vibration. Computational analysis using heat transfer research institute program is adopted to verify the results of the numerical calculations. It contains the evalution of performance in the system with view to location of the new reheater bundle and it shows the differences between the numerical calculation results and heat transfer research institute program output.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.141-148
• /
• 2016

This paper presents an overview of worldwide electric power development and National $700^{\circ}C$ Hyper Supercritical coal-fired power generation(HSC) focus on materials and manufacturing process. To Increase the efficiency of electric power generation, It is necessary to increase steam temperature and pressure. In that case, New material and manufacturing process shall be developed for boiler and turbine component in high temperature and pressure operating condition. Therefore, Much Efforts in worldwide are progressing to develop materials and manufacturing technology and to build and operate an HSC.

• Nuclear Engineering and Technology
• /
• v.36 no.4
• /
• pp.285-294
• /
• 2004

Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.25-30
• /
• 2013

The periodic inspection of piping and pressure vessels welds in nuclear power plant has to provide reliable result related to weld flaws, such as location, maximum amplitude response, ultrasonic length, height and finally the nature or flaw pattern. The founded flaw in ultrasonic inspection is accepted or rejected based on these data. Specially, the amplitude of flaw response is used as basic parameter for flaw sizing and it may cause some deviation in length sizing result. Currently the ultrasonic inspections in nuclear power plant components are performed by specific inspection procedure which describing inspection technique include inspection system, calibration methodology and flaw characterizing. To perform ultrasonic inspection during in-service inspection, reference gain should be established before starting ultrasonic inspection by the requirement of ASME code. This reference gain used as basic criteria to evaluate flaw sizing. Sometimes, a little difference in establishing reference gain between calibration and field condition can lead to deviation in flaw sizing. Due to this difference, the inspection result may cause flaw sizing error. Therefore, the objective of this study is to compare and evaluate the ultrasonic amplitude difference between air filled and water filled pipe in nuclear power plant. Additionally, the accuracy of flaw sizing is estimated by comparing both conditions.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.11 no.1
• /
• pp.74-78
• /
• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• Journal of Power System Engineering
• /
• v.11 no.4
• /
• pp.38-43
• /
• 2007

Pulsation is an inherent phenomenon in reciprocating compressors. It interacts with piping to cause vibrations and performance problems. Indiscriminately connecting to a compressor can be dangerous and cost money in the form of broken equipment and piping, poor performance, inaccurate metering, unwanted vibration, and sometimes noise. Piping connected to a compressor can materially affect the performance and response. To minimize these detrimental effects, reciprocating compressor system should be equipped by pulsation suppression system. The system usually comprises bottle volume, called snubber. Snubber is one of the most important parts in hydrogen compressing system. It has installed reciprocating hydrogen compressor. One of these components is snubber which has function to reduce pulsation waveform and to remove the impurities in the hydrogen gas. A snubber has an inclined plate as a buffer, which is installed inside snubber. When the pressure loss and the pulsation of pressure within a snubber is minimized, the snubber could get more applicability. Therefore, a study to find an optimum geometric size on a several snubbers which have different buffer width, has been conducted using a numerical analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.13 no.1
• /
• pp.75-83
• /
• 2017

This study investigated deformation behavior of major nuclear structural materials under cyclic loading conditions via cyclic stress-strain test. The cyclic stress-strain tests were conducted on SA312 TP316 stainless steel and SA508 Gr.3 Cl.1 low-alloy steel, which are used as materials for primary piping and reactor pressure vessel nozzle respectively, under cyclic load with constant strain amplitude and constant load amplitude at room temperature (RT) and $316^{\circ}C$. From the results of tests, the cyclic hardening and softening behavior, stabilized cyclic stress-strain behavior, and ratcheting behavior of both materials were investigated at both RT and $316^{\circ}C$. In addition, appropriate considerations for cyclic deformation behavior in the structural integrity evaluation of major nuclear components under excessive seismic condition were discussed.