• Nuclear Engineering and Technology
• /
• v.52 no.11
• /
• pp.2638-2651
• /
• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.24 no.11
• /
• pp.2796-2803
• /
• 2000

For traditional Leak-Before-Break(LBB) analyses, symmetric conditions were assumed for a pipe-nozzle interface to simplify the analysis in calculating J-integral. However. this assumption could result in an overly conservative design criteria for a pipe-nozzle interface, Since the pipe-nozzle interface is asymmetric due to the difference of stiffness between pipe and nozzle, it is required to develop a new methodology considering the nozzle effect. The objective of this paper is to evaluate the effect of nozzle no the development of LBB design criteria for nuclear pipings. For this purpose, extensive finite element analysis were performed to evaluate the effect of nozzle on Crack Opening Area(COA), Detectable Leakage Crack(DLC) length and J-integral values. In conclusion, it was proven that the application of LBB concept could be extended for more nuclear piping system by considering the nozzle.

• Nuclear Engineering and Technology
• /
• v.54 no.5
• /
• pp.1726-1746
• /
• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Journal of Welding and Joining
• /
• v.29 no.2
• /
• pp.64-71
• /
• 2011

In this paper, crack propagation analyses in the inner diameter (ID) repair weld of the dissimilar metal weldment of a nozzle were performed using a finite element alternating method (FEAM). To calculate the theoretical solution for the crack tip stress intensity factor, a weak type singular integral equation consisted of crack surface traction and dislocation density function was constructed and solved in conjunction with the FEAM. A two-dimensional axisymmetric finite element nozzle model was prepared and ID repair welding was simulated. An initial crack, 10% depth of weld thickness, was assumed and crack propagation trajectory from the initial crack to the 75% depth of thickness was calculated using the FEAM. Crack growth versus time curve was also calculated and compared with the curves obtained from ASME code method. With the method constructed in this paper, crack propagation trajectory and crack growth time were calculated automatically and effectively.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.56-61
• /
• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.20 no.6
• /
• pp.1872-1881
• /
• 1996

LBB(Leak-Before-Break) analysis is performed for the highest stress location of each different type of mateerials in the nuclear piping line. In most cases, the highest stress occurs in the pipe and nozzle interface location. i.e. terminal end. The current finite element analysis approach utilizes the symmetry condition both for locations near the nozzle and for locationa away from the nozzle to minimize the size of the finite element model and to make analysis simple when calculating the J-integral values at the crack tip. In other words, the nozzle is not included in the finite element model. However, in reality, the symmetric condition is not applicable for the pipe-nozzle interface location. Because the pipe-nozzle interface location is asymmetric due to different stiffenss of the pipe and nozzle(both material and dimensions). The simplified analysis approach for pipe-nozzle interface locaiton is too conservative for a smaller diameter piping. In tlhis paper, various analyses are performed for the range of materials and crack sizes to evaluate the nozzle effect for a LBB anlaysis. This paper presents methodology for developing the piping evaluaiton diagram at the pipe-nozzle interface location.

• Journal of Welding and Joining
• /
• v.28 no.1
• /
• pp.100-106
• /
• 2010

In this paper crack propagation analyses in the dissimilar metal weldment of a nozzle were performed using a finite element alternating method (FEAM). A two-dimensional axisymmetric finite element nozzle model was prepared and welding simulation including the thermal heat transfer analysis and the thermal stress analysis was performed. Initial cracks were inserted at weld and heat affected zone in the finite element model which has welding residual stress distribution obtained from the welding simulation. To calculate crack propagation trajectories of these cracks, a new fatigue crack evaluation module was developed in addition to the previous FEAM program. With the new FEAM fatigue crack evaluation module, crack propagation trajectory and crack growth time were calculated automatically and effectively.

• Journal of Welding and Joining
• /
• v.29 no.2
• /
• pp.34-39
• /
• 2011

Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.24 no.9 s.180
• /
• pp.2252-2258
• /
• 2000

In order to investigate the effect of nozzle on stress concentration in pressure vessels, three dimensional finite element analyses were performed. The results were compared with those for corresponding two dimensional axisymmetric finite element analyses. A three dimensional finite element model with a surface crack was also designed to evaluate the effect of internal pressure and heat transfer on crack face, and the resulting stress intensity factors from the finite element analyses were compared with those for ASME Sec. XI and Raju-Newman's stress intensity factor solution. As a result, the validity of currently available stress intensity factor solutions for a surface crack was reviewed in the presence of geometrical complexity, heat transfer and internal pressure.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2188-2197
• /
• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.