• Journal of Radiation Protection and Research
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• v.16 no.1
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• pp.53-72
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• 1991

An integrated model is developed in this paper for the performance assessment of high level radioactive waste repository. This integrated model consists of two simple mathematical models. One is a multiple-barrier failure model of the repository system based on constant failure rates which provides source terms to biosphere. The other is a biosphere model which has multiple pathways for radionuclides to reach to human. For the parametric uncertainty and sensitivity analysis for the risk assessment of high level radioactive waste repository, Latin hypercube sampling and rank correlation techniques are applied to this model. The former is cost-effective for large computer programs because it gives smaller error in estimating output distribution even with smaller number of runs compared to crude Monte Carlo technique. The latter is good for generating dependence structure among samples of input parameters. It is also used to find out the most sensitive, or important, parameter groups among given input parameters. The methodology of the mathematical modelling with statistical analysis will provide useful insights to the decision-making of radioactive waste repository selection and future researches related to uncertain and sensitive input parameters.

• International Journal of Reliability and Applications
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• v.18 no.2
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• pp.45-63
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• 2017

In this paper, we have formulated optimum Accelerated Life Test (ALT) plan for a parallel system with two independent components using masked data with ramp-stress loading scheme and Type-I censoring. Consider a system of two independent and non-identical components connected in parallel. Such a system fails whenever all of its components has failed. The exact component that causes the system to fail is often unknown due to cost and time constraint. For each parallel system at test, we observe its system's failure time and a set of component that includes the component actually causing the system to fail. The stress-life relationship is modelled using inverse power law, and cumulative exposure model is assumed to model the effect of changing stress. The optimal plan consists in finding out the optimum stress rate using D-optimality criterion. The method developed has been explained using a numerical example and sensitivity analysis carried out.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3472-3482
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• 2024

This study proposes a modified extended cognitive reliability and error analysis method (CREAM) for achieving a more accurate human error probability (HEP) in advanced control rooms. The traditional approach lacks failure data and does not consider the common performance condition (CPC) weights in different cognitive functions. The modified extended CREAM decomposes tasks using a method that combines structured information analysis (SIA) and the extended CREAM. The modified extended CREAM performs the weight analysis of CPCs in different cognitive functions, and the weights include cognitive, correlative, and important weights. We used the extended CREAM to obtain the cognitive weight. We determined the correlative weights of the CPCs for different cognitive functions using the triangular fuzzy decision-making trial and evaluation laboratory (TF-DEMATEL), and evaluated the importance weight of CPCs based on the interval 2-tuple linguistic approach and ensured the value of the importance weight using the entropy method in the different cognitive functions. Finally, we obtained the comprehensive weights of the different cognitive functions and calculated the HEPs. The accuracy and sensitivity of the modified extended CREAM were compared with those of the basic CREAM. The results demonstrate that the modified extended CREAM calculates the HEP more effectively in advanced control rooms.

• Journal of Power System Engineering
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• v.9 no.4
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• pp.137-142
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• 2005

The effects of temperature and initial crack length on impact fracture behavior of side plate material of 35 ton class FRP ship, which are composed by glass fiber and unsaturated polyester resin, were investigated. Impact fracture toughness of GF/PE composites displayed maximum value when the temperature of specimen is room temperature and $50^{\circ}C$, and with decrease in temperature of specimen, impact fracture toughness decreased. Impact fracture energy of GF/EP composites decreased with increase in initial crack length of specimen, and this value decreased rapidly when the temperature of specimen is lowest, $-25^{\circ}C$,. It is believed that sensitivity of notch on impact fracture energy were increased with decrease in temperature of specimen. As the GF/EP composites exposed in low temperature, impact fracture toughness of composites decreased gradually owing to the decrease of interface bonding strength caused by difference of thermal expansion coefficient between the glass fiber/polyester resin. Further, decrease of interface bonding strength of composites with decrease in specimen temperature was ascertained by SEM photograph of impact fracture surface.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.210-229
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• 2021

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.50 no.12
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• pp.819-828
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• 2022

Sensitivity of aircraft sizing depending on battery performance was studied for a generic quad tilt rotor type electric vertical takeoff and landing vehicle. The mission requirements proposed by Uber Elevate and NASA were used for initial sizing, and the calculated gross weight is ranged between 5,000lb and 11,000lb for battery specific energy range of 200-400Wh/kg in pack level and continuous discharge rate range of 4-5C. For the assumed gross weight of 7,000lb, the required battery performance was calculated with two different criteria: available power and energy, and the effects of battery specific energy and discharge rate are analyzed. The maximum discharge rate is also recommended considering failure cases such as one battery pack inoperative and one prop rotor inoperative.

• Journal of the Korean Society of Safety
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• v.35 no.5
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• pp.121-127
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• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.505-513
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• 2002

During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2002.05a
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• pp.134-139
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• 2002

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. However, effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. Acoustic emission(AE) has been widely used in various fields because of its extreme sensitivity, dynamic detection ability and location of growing defects. In this study, we investigated failure modes of locally wall thinned pipes and AE signals by bending test. From test results, we could be divided four types of failure modes of ovalization, crack initiation after ovalization, local buckling and crack initiation after local buckling. And fracture behaviors such as elastic region, yielding range, plastic deformation range and crack progress could be evaluated by AE counts, accumulative counts and time-frequency analysis during bending test. It is expected to be basic data that can protect a risk according to local wall thinning of pipes, as a real time test of AE.