Ученые из Южной Кореи разработали метод оценки рисков разрыва трубопроводов на атомных станциях

In nuclear power plants, the primary coolant piping systems are designed to maintain the structural integrity of the nuclear reactor under normal operating conditions. A pipe rupture that can result in a loss-of-coolant accident is a key consideration in this design and also an important safety concern. Traditionally, the double-ended guillotine break of the largest primary pipes, which assumes a complete break of the pipe into two sections, has been considered a key design-basis accident. However, such large breaks are extremely unlikely. Assessing the frequency of pipe rupture can therefore help engineers better understand the likelihood of different failure scenarios and focus resources on events that are more relevant to actual risks. Deterministic leak-before-break (LBB) and statistical approaches based on operating experience have been used previously to evaluate rupture frequency, but these methods do not account for degradation mechanisms or the influence of individual parameters. To address these limitations, a research team led by Professor Nam-Su Huh from the School of Mechanical Systems Engineering at Seoul National University of Science and Technology in South Korea utilized probabilistic fracture mechanics-based sensitivity analysis to investigate rupture behavior of the Korean nuclear power plant piping systems. "Probabilistic fracture mechanics makes it possible to estimate rupture frequency while accounting for the stochastic nature of material behavior, degradation over time, loading conditions, and even effectiveness of inspections," explains Prof. Huh. Their study was made available online on July 01, 2026, and will be published in Volume 197, Part B of Engineering Failure Analysis on November 01, 2026. The researchers utilized the eXtremely Low Probability of Rupture (xLPR) code to conduct their assessment. To this end, the researchers selected two LBB-approved piping systems from a reference Korean nuclear power plant—SC piping and a surge nozzle. The SC piping consisted of two pipes made of stainless steel welded together, also using the same stainless steel as the weld material. The surge nozzle was made up of a dissimilar metal weld composed of a low-alloy steel and stainless steel, and the weld metal for this case was a nickel alloy. The team first established a base case as a reference condition for sensitivity analysis. It included a fixed set of parameters, including piping geometry, loading conditions, and material properties. The researchers simulated 80 years of plant operation, considering stress corrosion cracking (SCC) as the only degradation mechanism. Sensitivity analysis was then conducted to evaluate the effect of weld residual stress (WRS), crack growth rate (CGR), weld overlay (WOL) repair, and inspection performance. WRS was the most influential parameter governing rupture predictions. Since residual stress affects the local stress state relevant to crack initiation and the crack-driving force for subsequent crack growth, variations in WRS can lead to large differences in predicted crack growth behavior. For the 95th-percentile, geometry-specific WRS profile, the predicted rupture frequency for the SC piping decreased considerably compared with the base case. In contrast, the mean, 5th-percentile, and probabilistic WRS cases showed no rupture during the 80-year simulation period. CGR was also significant. The 95th percentile CGR profile increased rupture frequency compared to the base case, while the probabilistic case reduced it. The 5th percentile case showed no rupture. The surge nozzle showed no rupture in all cases. In addition, WOL analysis also showed no rupture for either piping system. Interestingly, for the SC piping, periodic inspections significantly reduced the rupture frequency by several orders of magnitude, highlighting the effectiveness of inspections in reducing rupture frequency. Overall, the results show how uncertainties and modeling assumptions can lead to variations in predicted rupture frequency. "A probabilistic framework can help engineers identify which factors govern the predicted failure behavior. In addition, rupture-frequency estimates can help distinguish extremely unlikely large breaks from more credible break sizes and provide a technical basis for treating them differently in plant design and safety evaluations," remarks Prof. Huh. "In the long term, this type of research could contribute to risk-informed approaches to nuclear safety. Specifically, it can help maintain the safety of aging plants, and for new plants it could help incorporate structural safety into an earlier design stage." By highlighting the importance of probabilistic rupture frequency assessments, this study contributes to the development of better design and maintenance strategies for making nuclear power plants safer and more economical. Reference: Title of original paper: Impact of input uncertainties on the failure frequency of Korean nuclear piping systems based on probabilistic fracture mechanics Journal: Engineering Failure Analysis DOI: https://doi.org/10.1016/j.engfailanal.2026.111197 About the institute Seoul National University of Science and Technology (SEOULTECH) Seoul National University of Science and Technology, commonly known as 'SEOULTECH,' is a national university located in Nowon-gu, Seoul, South Korea. Founded in April 1910, SEOULTECH has grown into a large and comprehensive university with a campus size of 504,922 m2. It comprises 10 undergraduate schools, 35 departments, 6 graduate schools, and has an enrollment of approximately 15,108 students. Website: https://en.seoultech.ac.kr/ About Professor Nam-Su Huh Dr. Nam-Su Huh is a Professor in the School of Mechanical Systems Engineering at Seoul National University of Science and Technology and also the Principal Investigator of the Reliability-Based Mechanical System Design Laboratory (REMSYS Lab). His research focuses on the structural integrity and reliability of mechanical components, particularly in nuclear and energy systems, including fracture mechanics, structural assessment, and life evaluation of piping and pressure-boundary components. He has also worked at the Korea Atomic Energy Research Institute (KAERI) and Materialprüfungsanstalt Universität Stuttgart (MPA Stuttgart).