• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.2
• /
• pp.18-24
• /
• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.688-706
• /
• 2024

Austenitic stainless steels (ASS) are extensively employed in various sectors such as nuclear, power, petrochemical, oil and gas because of their excellent structural strength and resistance to corrosion. SS304 and SS316 are the predominant choices for piping, pressure vessels, heat exchangers, nuclear reactor core components and support structures, but they are susceptible to stress corrosion cracking (SCC) in chloride-rich environments. Over the course of several decades, extensive research efforts have been directed towards evaluating SCC using diverse methodologies and models, albeit some uncertainties persist regarding the precise progression of cracks. This review paper focuses on the application of Acoustic Emission Technique (AET) for assessing SCC damage mechanism by monitoring the dynamic acoustic emissions or inelastic stress waves generated during the initiation and propagation of cracks. AET serves as a valuable non-destructive technique (NDT) for in-service evaluation of the structural integrity within operational conditions and early detection of critical flaws. By leveraging the time domain and time-frequency domain techniques, various Acoustic Emission (AE) parameters can be characterized and correlated with the multi-stage crack damage phenomena. Further theories of the SCC mechanisms are elucidated, with a focus on both the dissolution-based and cleavage-based damage models. Through the comprehensive insights provided here, this review stands to contribute to an enhanced understanding of SCC damage in stainless steels and the potential AET application in nuclear industry.

• Journal of Mechanical Science and Technology
• /
• v.19 no.10
• /
• pp.1891-1901
• /
• 2005

Recently the authors tried to find damage position only using measured frequency response functions. According to their work, it seems that the algorithm is very practical since it needs only measured frequency responses while other methods require exact analytic model. But when applying the method to a real structure, it requires lots of experiment. The authors, in this time, propose a method to reduce its experimental load by detecting damage within a substructure. This method searches damages not within an entire structure but within substructures. In addition, damage severity was treated in this paper since it is worthy to know damage severity. Optimization technique is used to estimate damage level using measured responses and damage model. Two test examples, a plate and a jointed structure, are chosen to verify the suggesting method.

• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.553-564
• /
• 2013

This study evaluates the local damage of a turbine in an auxiliary building of a nuclear power plant due to an external impact by using the LS-DYNA finite element program. The wall of the auxiliary building is SC structure and the material of the SC wall plate is high manganese steel, which has superior ductility and energy absorbance compared to the ordinary steel used for other SC wall plates. The effects of the material of the wall, collision speed, and angle on the magnitude of the local damage were evaluated by local collision analysis. The analysis revealed that the SC wall made of manganese steel had significantly less damage than the SC wall made of ordinary steel. In conclusion, an SC wall made of manganese steel can have higher effective resistance than an SC wall made of ordinary steel against the local collision of an airplane engine or against a turbine impact.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.429.1-429
• /
• 2002

• Nuclear Engineering and Technology
• /
• v.52 no.2
• /
• pp.397-416
• /
• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part II, based on the verified finite element (FE) models of aircrafts Airbus A320 and A380, as well as the NPP containment and auxiliary buildings in Part I of this paper, the whole collision process is reproduced numerically by adopting the coupled missile-target interaction approach with the finite element code LS-DYNA. The impact induced damage of NPP plant under four impact locations of containment (cylinder, air intake, conical roof and PCS water tank) and two impact locations of auxiliary buildings (exterior wall and roof of spent fuel pool room) are evaluated. Furthermore, by considering the inner structures in the containment and raft foundation of NPP, the structural vibration analyses are conducted under two impact locations (middle height of cylinder, main control room in the auxiliary buildings). It indicates that, within the discussed scenarios, NPP structures can withstand the impact of both two aircrafts, while the functionality of internal equipment on higher floors will be affected to some extent under impact induced vibrations, and A380 aircraft will cause more serious structural damage and vibrations than A320 aircraft. The present work can provide helpful references to assess the safety of the structures and inner equipment of NPP plant under commercial aircraft impact.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1280-1290
• /
• 2016

This paper presents an advanced estimation method for obtaining the probability density functions of a damage parameter for valve leakage detection in a reciprocating pump. The estimation method is based on a comparison of model data which are simulated by using a mathematical model, and experimental data which are measured on the inside and outside of the reciprocating pump in operation. The mathematical model, which is simplified and extended on the basis of previous models, describes not only the normal state of the pump, but also its abnormal state caused by valve leakage. The pressure in the cylinder is expressed as a function of the crankshaft angle, and an additional volume flow rate due to the valve leakage is quantified by a damage parameter in the mathematical model. The change in the cylinder pressure profiles due to the suction valve leakage is noticeable in the compression and expansion modes of the pump. The damage parameter value over 300 cycles is calculated in two ways, considering advance or delay in the opening and closing angles of the discharge valves. The probability density functions of the damage parameter are compared for diagnosis and prognosis on the basis of the probabilistic features of valve leakage.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.238.2-238
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1394-1405
• /
• 2022

The importance of ensuring the inherent safety and security has been more emphasized in recent years to demonstrate the integrity of nuclear facilities under external human-induced events (e.g. aircraft crashes). This work suggests a simulation methodology to effectively evaluate the impact of a commercial aircraft engine onto a dry storage facility. A full-scale engine model was developed and verified by Riera force-time history analysis. A reinforced concrete (RC) structure of a dry storage facility was also developed and material behavior of concrete was incorporated using three constitutive models namely: Continuous Surface Cap, Winfrith, and Karagozian & Case for comparison. Strain-based erosion limits for concrete were suitably defined and the local responses were then compared and analyzed with empirical formulas according to variations in impact velocity. The proposed methodology reasonably predicted such local damage modes of RC structure from the engine missile, and the analysis results agreed well with the calculations of empirical formulas. This research is expected to be helpful in reviewing the dry storage facility design and in the probabilistic risk assessment considering diverse impact scenarios.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2053-2068
• /
• 2023

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.