• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.165-174
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• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.331-338
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• 1996

In the thermal-hydraulic design, the effect of fuel rod bow is quantified tv the rod bow DNBR penalty which is a key design parameter to assure the coolability of fuel assembly in the pressurized water reactor. In this work, a computer program for the evaluation of the rod bow DNBR penalty based on Westinghouse methodology is developed and its application procedure is proposed. The computer simulation is based on the Monte-Carlo method. The qualification of developed computer program is performed by a comparison of calculational result with that given by Westinghouse's document. A new application procedure is built using batch mean and batch standard deviation. The normality of sample population generated by the batch calculation is confirmed by means of a chi-square test for goodness of fit. On the view point of statistics it is effected that the more reliable design value may be produced by the new application procedure.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Journal of the Korean Society of Safety
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• v.29 no.6
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• pp.166-171
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• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.4
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• pp.292-303
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• 1998

Recent study has demonstrated that some magnetic properties are sensitive to the microstructural state of material. The ASTM A 508 Gr. 3 reactor pressure vessel steel has various microstructural changes including martensitic and bainitic phases, and various sizes of grain and precipitates in the weld heat-affected zone (HAZ). To correlate the microstructural state with Barkhausen noise (BN), specimens were prepared through simulating various weld thermal cycles using a thermal simulator. The conventional magnetic properties, i.e. coercive force, remanence and maximum induction, did not change significantly, whereas the BN amplitude and energy during a magnetization cycle changed markedly with microstructural state. The BN increased with increasing grain and carbide sizes, and the tempered bainite structure showed higher BN parameter than tempered martensite.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1429-1435
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• 2020

Applying an accurate parametric prediction model to identify abnormal or false pressurizer water levels (PWLs) is critical to the safe operation of marine pressurized water reactors (PWRs). Recently, deep-learning-based models have proved to be a powerful feature extractor to perform high-accuracy prediction. However, the effectiveness of models still suffers from two issues in PWL prediction: the correlations shifting over time between PWL and other feature parameters, and the example imbalance between fluctuation examples (minority) and stable examples (majority). To address these problems, we propose a cost-sensitive mechanism to facilitate the model to learn the feature representation of later examples and fluctuation examples. By weighting the standard mean square error loss with a cost-sensitive factor, we develop a Cost-Sensitive Long Short-Term Memory (CSLSTM) model to predict the PWL of PWRs. The overall performance of the CSLSTM is assessed by a variety of evaluation metrics with the experimental data collected from a marine PWR simulator. The comparisons with the Long Short-Term Memory (LSTM) model and the Support Vector Regression (SVR) model demonstrate the effectiveness of the CSLSTM.

• KIEE International Transaction on Electrical Machinery and Energy Conversion Systems
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• v.3B no.1
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• pp.10-15
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• 2003

This paper presents the digital computer performance evaluations of the three-phase self-excited induction generator (SEIG) driven by the variable speed prime mover such as the wind turbine using the nodal admittance approach steady-state frequency domain analysis with the experimental results. The three-phase SEIG setup is implemented for small-scale rural renewable energy utilizations. The experimental performance results give a good agreement with those ones obtained from the digital computer simulation. Furthermore, a feedback closed-loop voltage regulation of the three-phase SEIG as a power conditioner which is driven by a variable speed prime mover employing the static VAR compensator (SVC) circuit composed of the thyristor phase controlled reactor (TCR) and the thyristor switched capacitor(TSC) is designed and considered herein for the wind-turbine driven the power conditioner. To validate the effectiveness of the SVC-based voltage regulator of the terminal voltage of the three-phase SEIG, an inductive load parameter disturbances in stand-alone are applied and characterized in this paper. In the stand-alone power utilization system, the terminal voltage response and thyristor triggering angle response of the TCR are plotted graphically. The simulation and the experimental results prove the effectiveness and validity of the proposed SVC which is controlled by the Pl controller in terms of fast response and high performances of the three-phase SEIG driven directly by the rural renewable energy utilization like a variable-speed prime mover.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.791-798
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• 2012

Since standardized fracture test specimens cannot be easily extracted from in-service components, several alternative fracture toughness test methods have been proposed to characterize the deformation and fracture resistance of materials. One of the more promising alternatives is the local approach employing the SP(Small Punch) testing technique. However, this process has several limitations such as a lack of anisotropic yield potential and tediousness in the damage parameter calibration process. The present paper investigates estimation of ductile fracture resistance(J-R) curve by FE(Finite Element) analyses using an anisotropic damage model and enhanced calibration procedure. In this context, specific tensile tests to quantify plastic strain ratios were carried out and SP test data were obtained from the previous research. Also, damage parameters constituting the Gurson-Tvergaard-Needleman model in conjunction with Hill's 48 yield criterion were calibrated for a typical nuclear reactor material through a genetic algorithm. Finally, the J-R curve of a standard compact tension specimen was predicted by further detailed FE analyses employing the calibrated damage parameters. It showed a lower fracture resistance of the specimen material than that based on the isotropic yield criterion. Therefore, a more realistic J-R curve of a reactor material can be obtained effectively from the proposed methodology by taking into account a reduced load-carrying capacity due to anisotropy.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.60-66
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• 2017

The standard master curve (MC) approach has the major limitation because it is only applicable to homogeneous datasets. In nature, materials are macroscopically inhomogeneous and involve scatter of fracture toughness data due to various deterministic material inhomogeneity and random inhomogeneity. RPV(reactor pressure vessel) steel has different fracture toughness with varying distance from the inner surface of the wall due to cooling rate in manufacturing process; deterministic inhomogeneity. On the other hand, reference temperature, $T_0$, used in the evaluation of fracture toughness is acting as a random parameter in the evaluation of welding region; random inhomogeneity. In the present paper, four regions, the surface, 1/8T, 1/4T and 1/2T, were considered for fracture toughness specimens of KSNP (Korean Standard Nuclear Plant) SA508 Gr. 3 steel to investigate deterministic material inhomogeneity and random inhomogeneity. Fracture toughness tests were carried out for four regions and three test temperatures in the transition region. Fracture toughness evaluation was performed using the bimodal master curve (BMC) approach which is applicable to the inhomogeneous material. The results of the bimodal master curve analyses were compared with that of conventional master curve analyses. As a result, the bimodal master approach considering inhomogeneous materials provides better description of scatter in fracture toughness data than conventional master curve analysis. However, the difference in the $T_0$ determined by two master curve approaches was insignificant.