• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.9
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• pp.1384-1390
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• 2004

The full containment Liquefied Natural Gas(LNG) storage tank is based on a double liquid container concept : two separate containers, one within the other, are capable of containing the LNG. The outer concrete tank provides comer protection(secondary containment) to withstand and safely contain any spill from the inner tank. The comer protection is installed on inside corner surface of outer concrete tank. Because of high and complex stresses, corner protection is designed by ASME section ⅧI Div. 2, Appendix 4 on behalf of API 620 which is main design code for LNG tank. Design guidelines to determine design factors such as liner thickness and knuckle radius are not well understood because Appendix 4 is the design method not based on equation but FEM. Recently, the volume of LNG tank shows a tendency to increase. So it is necessary to set up the design guidelines to cope with change of LNG tank capacity and height/diameter ratio. In this paper, optimum design of corner protection was performed and the design guidelines were suggested by the results of FEM for LNG tanks which have different capacities and height/diameter ratio.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.15-19
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• 2014

A Failure or degradation of reactor upper head penetration is a recurring problem due to long term operation at nuclear power plants. And a flaw in the reactor upper head penetration has caused unplanned plant shutdown for repair as well as high economic impact on the plants. Consequently, a detection of flaws is of the utmost importance. Prior to the replacement of reactor upper head penetration, some utilities have repaired the flaws of reactor upper head penetration generated by overlay weld. Until now, only the base metal in reactor upper head penetration has been inspected according to 10 CFR 50.55a and ASME code case N-729-1. Accordingly, it is difficult to detect manufacturing defects and repair defects in overlay weld. This paper presents a case study on the application of Time of Flight Diffraction technique for reactor head penetration mockup with artificial flaws in overlay weld. This study offers a way to understand the flaws detected in reactor upper head penetration overlay weld.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.25-30
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• 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
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• v.14 no.1
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• pp.58-65
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• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• Journal of the Korean Society of Systems Engineering
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• v.13 no.1
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• pp.25-31
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• 2017

This paper aims to evaluate the conservatism in the design of APR1400 (Advanced Pressurized water Reactor 1400 designed by KHNP) reactor internals component, the LSS (Lower Support Structure). Re-engineering of the LSS is done based on the system design condition data and applicable ASME code that was used for the original APR1400 design. Systems engineering approach is applied to design the LSS of APR1400 without refering APR1400 LSS dimensional parameters and tries to verify important design parameters of APR1400 LSS as well as the validity of the re-engineering design process as independent verification method of reactor component design. Systems engineering approach applied in this study following V-model approach. The re-engineered LSS design showed more than enough conservatism for static loading case. The maximum deflection of LSS is under 1mm (calculated value is 0.25mm) from 4000 mm diameter of LSS. Hence the deflection can be ignored in other reactor internals for structural integrity assessment. Especially the effect of LSS deflection on fuel assembly can be minimized and which is one of the main requirements of LSS design. It also showed that the maximum stress intensity is 2.36MPa for the allowable stress intensity of 60.1 MPa. The stress resulted from the static load is also very small compared to the maximum allowable stress intensity, hence there is more than enough conservatism in the LSS design.

• Journal of the Korean Society of Safety
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• v.33 no.5
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• pp.164-169
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• 2018

A detailed fatigue evaluation procedure was developed to mitigate the excessive conservativeness of the conventional environmental fatigue evaluation method for the pressurizer spray line elbow of domestic new nuclear power plants. The pressurizer spray line is made of austenitic stainless steel, which is relatively sensitive to the environmentally assisted fatigue, and has a low degree of design margin in terms of environmentally assisted fatigue due to the thermal stratification phenomenon on the pipe cross section as a whole or locally. In this study, to meet the environmental fatigue design requirements of the pressurizer spray line elbow, the new environmental fatigue evaluation has been performed, which used the ASME Code NB-3200-based detailed fatigue analysis and the environmental fatigue correction factor instead of the existing NB-3600 evaluation method. As a result, the design requirements for environmentally assisted fatigue were met in all parts of the pressurizer spray line elbow including the fatigue weakened zones by thermal stratification.

• Nuclear Engineering and Technology
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• v.38 no.3
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• pp.277-284
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• 2006

In order to ensure the structural integrity of nuclear welded structures during design life, the fatigue life has to be evaluated by fatigue analysis procedures presented in technical codes such as ASME B&PV Code Section III. However, existing fatigue analysis procedures do not explicitly consider the presence of welded joints. A new fatigue analysis procedure based on a structural stress/fracture mechanics approach has been recently developed in order to reduce conservatism by erasing uncertainty in the analysis procedure. A recent review of fatigue crack growth data under various mean loading conditions using the structural stress/fracture mechanics approach, does not consider the mean loading effect, revealed some significant discrepancies in fatigue crack growth curves according to the mean loading conditions. In this paper, we propose the use of the stress intensity factor range ${\Delta}K$ characterized with loading ratio R effects in terms of the structural stress. We demonstrate the effectiveness in characterizing fatigue crack growth and S-N behavior using the well-known data. It was identified that the S-N data under high mean loading could be consolidated in a master S-N curve for welded joints.

• Journal of Welding and Joining
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• v.24 no.2
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• pp.48-56
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• 2006

The fatigue strength of the welds is affected by such factors as the weld geometry, microstructures, tensile properties and residual stresses caused by fabrication. It is very important to evaluate the structural integrity of the welds in nuclear power plant because the weldment undergoes the most of damage and failure mechanisms. In this study, the fatigue assessments for a reactor vessel outlet nozzle with the weldment to the piping system are performed considering the welding residual stresses as well as the effect of local brittle zone in the vicinity of the weld fusion line. The analytical approaches employed are the microstructure and mechanical properties prediction by semi-analytical method, the thermal and stress analysis including the welding residual stress analysis by finite element method, the fatigue life assessment by following the ASME Code rules. The calculated results of cumulative usage factors(CUF) are compared for cases of the elastic and elasto-plastic analysis, and with or without residual stress and local brittle zone effects, respectively. Finally, the fatigue life of reactor vessel outlet nozzle weld is slightly affected by the local brittle zone and welding residual stresses.

• Journal of Welding and Joining
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• v.27 no.2
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• pp.63-68
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• 2009

Microminiature heat exchanger has been applied to the gas turbine in order to increase energy efficiency. During the production of microminiature heat exchanger, however, it is very difficult to weld tube to tubesheet. In this study, therefore, welding process of resistance ring projection was used, and weld tensile tests were performed. Sound weld joint was obtained as a result of applying resistance ring projection welding to microminiature heat exchanger to tubesheet. Cold weld occurred at under 1600A. Even though tensile strength was increased with increasing current, splash occurred and tensile strength decreased at 2000A due to the excessive current. Therefore it was determine that the optimal current is 1900A. As result of tensile tests based on ASME code for tube to tubesheet weldment, rupture position was weldment due to Fs(Fractured section) of nugget, which was smaller than tube thickness (t), and it was proven as a partial strength welding because of the average joint efficiency fr = 0.90.