• Korean Journal of Fisheries and Aquatic Sciences
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• v.42 no.5
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• pp.494-502
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• 2009

A quantitative model was developed in order to estimate fishery production damage due to anthropogenically induced environmental changes. The model is described in the following equation, $Y_D=\frac{{\phi}_D}{{\phi}_G}[Y_0{\cdot}(t_p-t_0)-\frac{Y_0}{{\phi}_G}(1-e^{-{\phi}_G(t_p-t_0)})]$, where, $Y_D$ is annual amount of fishery production by nuclear power plant. ${\varphi}$ D and ${\varphi}$ G are instantaneous decreasing coefficient of fishery production by nuclear power plant and instantaneous decreasing coefficient of gross fishery production, respectively. $Y_0$ is annual mean fishery production without damages. $t_p$ is the present time, and $t_0$ is the starting time of damages. The model was applied to fishing grounds near a nuclear power plant on the east coast of Korea. Since fishery production damages have become bigger with increasing emission of thermal effluents from generators activities in the power plant, this factor has also been considered as, $\delta_{D_i}=\delta_D\({\sum}\limits_{i=0}^{n}\;W_i/W_T\)$, where, $\delta_{Di}$ is the cumulative damage rate in fishery production from generators, $\delta_D$ is the total cumulative damage rate in fishery production, $W_i$ is the emission amount of thermal effluents by generator i, and n is the number of generators in the nuclear power plant. This model can be used to conduct initial estimates of fishery production damages, before more detailed assessments are undertaken.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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• pp.307-322
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• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1758-1764
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• 2019

As the safety level of nuclear power plants (NPPs) relates to the safety of individuals, society, and the environment, it is important to establish NPP safety goals. In Korea, two quantitative health objectives and one large release frequency (LRF) criterion were formally set as quantitative safety goals for NPPs by the Nuclear Safety and Security Commission in 2016. The risks of prompt and cancer fatalities from NPPs should be less than 0.1% of the overall risk, and the frequency of nuclear accidents releasing more than 100 TBq of Cs-137 should not exceed 1E-06 per reactor year. This paper reviews the hierarchical structure of safety goals in Korea, its relationship with those of other countries, and the relationships among safety goals and subsidiary criteria like core damage frequency and large early release frequency. By analyzing the effect of the release of 100 TBq of Cs-137 via consequence analysis codes in eight different accident scenarios, it was shown that meeting the LRF criterion results in negligible prompt fatalities in the surrounding area. Hence, the LRF criterion dominates the safety goals for Korean NPPs. Safety goals must be consistent with national policy, international standards, and the goals of other counties.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1935-1946
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• 2022

The investigations of the coated-particles of nuclear fuel samples are carried out in three stages: front-end, irradiation in the reactor core, and post-irradiation examination. The front-end stage is the initial analysis of the failures rates of produced samples before they are placed in the reactor core. The purpose of the verification is to prepare the particles for an experiment that will determine the degree of damage to the coated particles at each stage. Before starting experiments with the samples, they must be properly prepared. Polishing the samples in order to uncover the inner layers is an important, initial experimental step. The authors of this paper used a novel way to prepare samples for testing - by applying an ion polisher. Mechanical polishing used frequently for sample preparations generates additional mechanical damages in the studied fuel particle, thus directly affecting the experimental results. The polishing methods were compared for three different coated particles using diagnostic methods such as Raman spectroscopy, scanning electron microscopy, and confocal laser scanning microscopy. Based on the obtained results, it was concluded that the ion polishing method is better because the level of interference with the structures of the individual layers of the tested samples is much lower than with the mechanical method. The same technique is used for the fuel particles undergone ion implantation simulating radiation damage that can occur in the reactor core.

• Steel and Composite Structures
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• v.48 no.4
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• pp.385-403
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• 2023

To avoid premature damage to the connection joints of a conventional precast concrete shear wall, a new precast concrete shear wall system (NPSW) based on a plastic damage relocation design concept was proposed. Five specimens, including one monolithic cast-in-place concrete shear wall (MSW) as a reference and four NPSWs with different connection details (TNPSW, INPSW, HNPSW, and TNPSW-N), were designed and tested by lateral low-cyclic loading. To accurately assess the damage relocation effect and quantify the damage and deformation, digital image correlation (DIC) and conventional data acquisition methods were used in the experimental program. The concrete cracking development, crack area ratio, maximum residual crack width, curvature of the wall panel, lateral displacement, and deformed shapes of the specimens were investigated. The results showed that the plastic damage relocation design concept was effective; the initial cracking occurred at the bottom of the precast shear wall panel (middle section) of the proposed NPSWs. The test results indicated that the crack area ratio and the maximum residual crack width of the NPSWs were less than those of the MSW. The NPSWs were deformed continuously; significant distortions did not occur in their connection regions, demonstrating the merits of the proposed NPSWs. The curvatures of the middle sections of the NPSWs were lower than that of the MSW after a drift ratio of 0.5%. Among the NPSWs, HNPSW demonstrated the best performance, as its crack area ratio, concrete damage, and maximum residual crack width were the lowest.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1638-1653
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• 2024

Given the unpredictability of the occurrence of the earthquake and other potential disasters into consideration, the nuclear power plant may be confronted with beyond design-basis earthquake load in the future. The containment structure may be severely damaged under such severe earthquake loading, increasing the risk of containment concrete cracking and potential radioactive materials leaking. Moreover, initial damage caused by the earthquake may significantly alter the pressure performance of the containment under follow-up internal pressure. To compromise the dangers of beyond design-basis earthquake to the containment, an alternative of replacing the conventional concrete with fiber-reinforced concrete (FRC) to upgrade the seismic resistance capacity of the containment is attempted and thoroughly researched. In this study, the influence of various fiber types such as rigid fiber and mixed fiber is regarded to constitute fiber-reinforced PCCVs. The physical properties of traditional and fiber-reinforced PCCVs under earthquake ground motions are scientifically compared and identified by using traditional and proposed evaluation indices. The results indicate that both the traditional evaluation index (i.e. top displacement, stress, strain) and the proposed damage index are greatly reduced by the practice of fiber strengthening under earthquake ground motions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2009.10a
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• pp.441-445
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• 2009

The most of the goals of pump designers and users are efficient operation and productivity. But the safety-grade pumps in nuclear power plants are needed to operate continuously for an essential condition of system operation. Also, most of the rules and regulatory standards that have been prepared for nuclear pumps are dedicated to achieve public safety. The study examined pump vibration in a pump outlet flow and distinguished the regions of pump vibration frequency cause by cavitation and recirculation. The study made a counterproposal in determination of pump outlet flow so that the discharge flow will be able to minimize the long-term damage of the pump.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.837-843
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• 1998

This paper presents the thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the secondary piping of LMFR using Green's function method and standard FEM. The thermohydraulic loading conditions used in the present analysis are simplified sinusoidal thermal loads and the random type data thermal load. The thermomechainical fatigue damage was evaluated according to ASME code subsectionNH. The analysis results of fatigue for the sinusoidal and random load cases show that fatigue failure would occur at a geometrically discontinuous location during 90,000 hours of operation The fracture mechanics analysis showed that the crack would be initiated at an early stage of the operation. The fatigue crack was evaluated to propagate up to 5 ㎜ along the thickness direction during the first 944 and 1083 hours of operation for the sinusoidal and the random loading cases, respectively.