• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• Journal of Energy Engineering
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• v.26 no.3
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• pp.64-70
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• 2017

The concept of Hybrid Safety Injection Tank (Hybrid SIT) proposed by the Korea Atomic Energy Research Institute (KAERI) has been introduced for the purpose of application to the Advanced Power Reactor Plus (APR+). In this study, the SBO situation of the APR+ was analyzed by using the MARS-KS code in order to evaluate whether the operation of the Hybrid SIT has an effect on the cooling performance of the Reactor Coolant System (RCS). According to the analysis, when the actuation valve on the pressure balancing line (PBL) is opened, the Hybrid SIT's pressure rises rapidly, forming equilibrium with the RCS pressure; subsequently, a flow is injected from the Hybrid SIT into the reactor vessel through the direct vessel injection (DVI) line. The analysis showed that it is possible to keep the core temperature below melting temperature during the operation of a Hybrid SIT.

• Journal of Ocean Engineering and Technology
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• v.36 no.5
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• pp.340-352
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• 2022

This article presents a modeling method for the uncorrelated measurement error of the ultra-short baseline (USBL) acoustic positioning system for aiding navigation of underwater vehicles. The Mahalanobis distance (MD) and principal component analysis are applied to decorrelate the errors of USBL measurements, which are correlated in the x- and y-directions and vary according to the relative direction and distance between a reference station and the underwater vehicles. The proposed method can decouple the radial-direction error and angular direction error from each USBL measurement, where the former and latter are independent and dependent, respectively, of the distance between the reference station and the vehicle. With the decorrelation of the USBL errors along the trajectory of the vehicles in every time step, the proposed method can reduce the threshold of the outlier decision level. To demonstrate the effectiveness of the proposed method, simulation studies were performed with motion data obtained from a field experiment involving an autonomous underwater vehicle and USBL signals generated numerically by matching the specifications of a specific USBL with the data of a global positioning system. The simulations indicated that the navigation system is more robust in rejecting outliers of the USBL measurements than conventional ones. In addition, it was shown that the erroneous estimation of the navigation system after a long USBL blackout can converge to the true states using the MD of the USBL measurements. The navigation systems using the uncorrelated error model of the USBL, therefore, can effectively eliminate USBL outliers without loss of uncontaminated signals.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1414-1422
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• 2017

Remote response technology has advanced to the extent that a robot system, if properly designed and deployed, may greatly help respond to beyond-design-basis accidents at nuclear power plants. Particularly in the aftermath of the Fukushima accident, there is increasing interest in developing disaster robots that can be deployed in lieu of a human operator to the field to perform mitigating actions in the harsh environment caused by extreme natural hazards. The nuclear robotics team of the Korea Atomic Energy Research Institute (KAERI) is also endeavoring to construct disaster robots and, first of all, is interested in finding out to what extent safety benefits can be achieved by such a disaster robotic system. This paper discusses a new approach based on the probabilistic risk assessment (PRA) technique, which can be used to quantify safety benefits associated with disaster robots, along with a case study for seismic-induced station blackout condition. The results indicate that to avoid core damage in this special case a robot system with reliability > 0.65 is needed because otherwise core damage is inevitable. Therefore, considerable efforts are needed to improve the reliability of disaster robots, because without assurance of high reliability, remote response techniques will not be practically used.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.587-591
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• 1997

The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shut down the nuclear reactor safely in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defects, its failure modes were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless maintenance, the preventive maintenance, the database of maintenance information, and the predictive maintenance.

• Journal of the Korean Society of Safety
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• v.17 no.3
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• pp.123-130
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• 2002

This paper presents a new dynamic human reliability analysis method and its application for quantifying the human error probabilities in implementing accident management actions. The action associated with implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concept of the quantified correlation between the performance requirement and performance achievement. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which m most frequency used method in PSAs, are discussed. The MAAP code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic obtained, human error probabilities are calculated with respect to the various means and variances of the things. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.2
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.3
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• pp.10-21
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• 2014

In Fukushima accident, when the severe accident such as a natural disaster happens, it is impossible to monitor the plant status due to a extreme environment and station blackout and most I&C systems break downs. Finally, these cause the loss of emergency cooling function and thus results in a hydrogen explosion and radiation leak. In this paper, the emergency response system is introduced that monitors and controls properly when the sever accidents like Fukushima accident happen, And the performance requirements of a wireless communication system used in the emergency respons system is described and the performance of emergency communication system is analyzed using the markov model.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.360-372
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• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.66-73
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• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.