• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.128-141
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• 2000

The critical heat flux (CHF）experiments have been carried out in a wide range of pressures for an internally heated vertical annulus. The experimental conditions covered ranges of pressures from 0.57 to 15.01 MPa, mass fluxes of 0 kg/$m^2$s and from 200 to 650 kg/$m^2$s, and inlet subcoolings from 85 to 413 kJ/kg. The characteristics of the present data and the effect of pressure on CHF are discussed. Most of the CHFs were identified to dryout of the liquid film in the annular or annular-mist flow. For the mass flux of 200 kg/$m^2$s, there were the indications that the CHF occurred at the transition from annular to annular-mist How in the pressure range of 3~10 MPa. For the mass fluxes of 550 and 650 kg/$m^2$s, the CHFs had a maximum value at a pressure of 2~3 MPa, and the pressure at the maximum CHF values had a trend moving toward the pressure at the peak value of pool boiling CHF as the mass flux decreased. The CHF data under a zero mass flux condition indicate that both the effects of pressure and inlet subcooling on the CHF were smaller, compared with those on the CHF with net water upward flow.

• Journal of Radiation Protection and Research
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• v.29 no.3
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• pp.205-212
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• 2004

Three-dimensional wind field and atmospheric dispersion models have been developed for estimating the concentration distributions of radioactive materials released into atmosphere. The field tracer experiment near the Kori nuclear power plant located over complex terrain was carried out for validating the atmospheric dispersion model. The wind fields were one of the most important factors for calculating the concentration. Therefore several numerical simulations using the measured wind data were performed to get more accurate concentration distributions compared with the analyzed values of the tracer gas. The calculated concentration distributions agreed well in the case of the usage of the more measured wind data in wind field model.

• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.1-11
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• 1981

The KINS (KAERI-Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactors, has been developed and benchmarked against the first cycle of the Kori-1 reactor. The KINS program is based on the computational model used in FLARE code and has been modified to represent the PWR characteristics more explicitly. The critical boron concentration and three-dimensional power distribution at the beginning of life hot zero power have been calculated and compared with the operating data. A three-dimensional depletion calculation at the intervals of 1000 MWD/MTU turnup steps has been performed. As the result of comparison, our calculation is shown to be in excellent agreement with the operating data. It is displayed that, incorporated with the computing time, the KINS program is an effective and powerful tool for PWR core management.

• Corrosion Science and Technology
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• v.18 no.2
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• pp.61-71
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• 2019

Flow-Accelerated Corrosion (FAC) is a phenomenon in which a protective coating on a metal surface is dissolved by a flow of fluid in a metal pipe, leading to continuous wall-thinning. Recently, many countries have developed computer codes to manage FAC in power plants, and the FAC prediction model in these computer codes plays an important role in predictive performance. Herein, the FAC prediction model was developed by applying a machine learning method and the conventional nonlinear regression method. The random forest, a widely used machine learning technique in predictive modeling led to easy calculation of FAC tendency for five input variables: flow rate, temperature, pH, Cr content, and dissolved oxygen concentration. However, the model showed significant errors in some input conditions, and it was difficult to obtain proper regression results without using additional data points. In contrast, nonlinear regression analysis predicted robust estimation even with relatively insufficient data by assuming an empirical equation and the model showed better predictive power when the interaction between DO and pH was considered. The comparative analysis of this study is believed to provide important insights for developing a more sophisticated FAC prediction model.

• Journal of Radiation Protection and Research
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• v.30 no.3
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• pp.121-130
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• 2005

Geographical Information System(GIS) was established to display the calculation results which show the concentration change with time and regions in case of an accidental release of radionuclides from Kori Nuclear Power Plants. GIS included the commercial program, ArcView(ESRI), and a basic digital map of 1:5000 scale lot 20km by 20km around Kori area. The object for the presentation was $^{131}I$ concentration in rice which is one of staple foodstuffs. Provided by deposited $^{131}I$ concentrations, ECOREA-II code computed the $^{131}I$ concentration of the soil and the plant in the area divided by In unit cells in total, in which the concentrations also varied with time. The results were introduced into the attributed data of previously designed polygon cells in ArcView. In order to display the concentration change with time by monotonic color, the RGB value for ArcView color lamp was controlled. This display definitely helped the concentration change around Kori area be acceptable to public.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.108-113
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• 1996

Several analytical models for the departure from nucleate boiling (DNB) phenomenon have been developed during the last decade. Among these, Chang & Lee's model based on a bubble crowding mechanism is remarkable in the fundamental features characterized as the formulation of mass, energy, and momentum balance equation at thermal-hydraulic conditions leading to the DNB. However, Bricard and Souyri remarked that the assumption of stagnant bubbly layer at the DNB condition is questionable and the signs on the axial projections of the momentum fluxes at the core/bubbly layer interface in the momentum balance equations are erroneous. From this remark, Chang & Lee's model has been re-examined and modified by correcting the erroneous treatments in the momentum balance equations and removing the spurious assumptions. The revised model predicts well the extensive DNB data of water in uniformly heated tubes at low qualities and shows more accurate prediction compared with the original model.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.273-278
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• 1997

Critical two-phase flow rates of subcooled water through very short tube (L=20 mm) with small diameters (D=1.0 mm) has been measured for wide ranges of subcooling(0~186$^{\circ}C$) and pressure (0.5~2.0 MPa). Experimental results show that subcooled critical two-phase flow rates can be expressed in terms of two scaling parameters for geometries and initial conditions. They are discharge coefficient of cold water, ( $C_{d}$ )$_{ref}$, and dimensionless subcooling, $\Delta$ $T^{*}$$_{sub}$, respectively. A new empirical correlation expressed in terms of ( $C_{d}$ )$_{ref}$ and $\Delta$ $T^{*}$$_{sub}$ is obtained for subcooled two-phase flow rates through very short length tube. Comparisons between the mass fluxes calculated by Present correlation and a number of experimental data show that the agreement is very good.ood.ood.ood.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.389-400
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• 1994

The predictability of KAERI version of RELAP5/MOD3 on reflood peak cladding temperature during large break loss-of-coolant accident is assessed against 18 test runs in FLECHT SEASET test data. The associated uncertainty is statistically quantified. The selected test runs include a gravity feed test and several forced feed tests with wide range of the parameters such as flooding rate, system pressure, initial clad temperature, rod bundle power. The results show that the code under-predicts the peak cladding temperature by 7.56 K on average. The upper limit of the associated uncertainty at 95% confidence level is evaluated to be about 99 K, It including the bias due to the under-prediction.