• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.10a
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• pp.404-405
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• 2013

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2880-2886
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• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.2
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• pp.137-148
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• 2012

In pressurized water nuclear reactors, dissimilar metal welds are susceptible to primary water stress corrosion cracking. To access this problem, accurate estimation of welding residual stresses is important. This paper provides general welding residual stress profiles in dissimilar metal nozzle butt welds using finite element analysis. By introducing a simplified shape for dissimilar metal nozzle butt welds, changes in the welding residual stress distribution can be seen using a geometry variable. Based on the results, a welding residual stress profile for dissimilar metal nozzle butt welds is proposed that modifies the existing welding residual stress profile for austenitic pipe butt welds.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.149-160
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• 2007

The IAEA launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and developed the INPRO Methodology to provide guidelines and to assess the characteristics of a future innovative nuclear energy system in areas such as safety, economics, waste management, and proliferation resistance. The proliferation resistance area of the INPRO Methodology is reviewed here, and modifications for further improvements are proposed. The evaluation metrics including the evaluation parameters, evaluation scales and acceptance limits are developed for a practical application of the methodology to assess the proliferation resistance. The proliferation resistant characteristics of the DUPIC fuel cycle are assessed by applying the modified INPRO Methodology based on the developed evaluation metrics and acceptance criteria. The evaluation procedure and the metrics can be utilized as a reference for an evaluation of the proliferation resistance of a future innovative nuclear energy system.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.291-298
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• 1997

In advanced nuclear reactors, the passive containment cooling has been suggested to enhance the safety. The passive cooling has two mechanisms, air natural convection and oater cooling with evaporation. To confirm the coolability of PCCS, many works have been performed experimentally and numerically. In this study, the water cooling test was performed to obtain the evaporative heat transfer coefficients in a scaled don segment type PCCS facility which have same configuration with AP600 prototype containment. Air-steam mixture temperature and velocity, relative humidity and well heat flux are measured. The local steam mass flow rates through the vertical plate part of the facility are calculated from the measured data to obtain evaporative heat transfer coefficients. The measured evaporative heat transfer coefficients are compared with an analytical model which use a mass transfer coefficients. From the comparison, the predicted coefficients show good agreement with experimental data however, some discrepancies exist when the effect of wave motion is not considered. Finally, a new correlation on evaporative heat transfer coefficients are developed using the experimental values.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.961-970
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• 2012

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.391-402
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• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.305-312
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• 2010

The protection sequence of an AC/DC converter for an ITER PF coil system is developed in this study. Possible faults in the coil system are simulated and the results reflected in the design of a sequence to protect the coil and converter. The inductances of the current sharing reactors and thyristor numbers in parallel with the bridge arms are optimized with the designed protection sequence.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1219-1225
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• 2017

In this work, a scalable algorithm for model calibration in nuclear engineering applications is presented and tested. The algorithm relies on the construction of surrogate models to replace the original model within the region of interest. These surrogate models can be constructed efficiently via reduced order modeling and subspace analysis. Once constructed, these surrogate models can be used to perform computationally expensive mathematical analyses. This work proposes a surrogate based model calibration algorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulics parameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used to simulate a three-dimensional core depletion problem. The proposed algorithm is then used to construct a reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronics and thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilation and calibration are highlighted in an uncertainty quantification study and requantification after the calibration process. Results showed that the proposed algorithm could help to reduce the uncertainty in key reactor attributes based on experimental and operational data.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.148-153
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• 1996

In this paper, an effective method for resolving difficulty resulting from the heterogeneity of the PWR baffle/reflector region is developed on the basis of the AFEN method. The essential difference of the new method from the conventional approach based on the equivalence theory is that the heterogeneous baffle/reflector is directly, without homogenization, considered as a node in nodal calculation Numerical results show that AFEN method with the new method can accurately predict both the multiplication factor and the power distribution of thermal reactors with baffle explicitly modeled.