• Nuclear Engineering and Technology
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• 제30권4호
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• pp.364-376
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• 1998

The statistical analysis method was applied to the evaluation of fracture toughness in the ductile-brittle transition temperature region. Because cleavage fracture in steel is of a statistical nature, fracture toughness data or values show a similar statistical trend. Using the three-parameter Weibull distribution, a fracture toughness vs. temperature curve (K-curve) was directly generated from a set of fracture toughness data at a selected temperature. Charpy V-notch impact energy was also used to obtain the K-curve by a $K_{IC}$ -CVN (Charpy V-notch energy) correlation. Furthermore, this method was applied to evaluate the neutron irradiation embrittlement of reactor pressure vessel (RPV) steel. Most of the fracture toughness data were within the 95% confidence limits. The prediction of a transition temperature shift by statistical analysis was compared with that from the experimental data.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.571-581
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• 2004

A methodology is proposed for the evaluation of a steam explosion load on a reactor scale by evaluating the steam explosion model against the experimental data. Being part of the OECD/SERENA program,, appropriate data was selected by international experts and the analytical model of TEXAS-V was chosen. The procedure consists of two steps. the pre-mixing model was verified against the FARO L-14 and FARO L-28 data. The explosion model was verified against the experimental data of KROTOS-44, FARO L-33, TROI-13, and TROI-34. The capabilities and deficiencies of the fundamental models of the TEXAS-V are reviewed in terms of their adequacy in a simulation of steam explosion on a reactor scale.

• Nuclear Engineering and Technology
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• 제34권5호
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• pp.482-493
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• 2002

The MOX fuel for LWR is fabricated either by direct mechanical blending of UO$_2$ and PuO$_2$ or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX fuel, the thermal conductivity of LWR MOX fuel scatters from 80 to 100 % of UO$_2$ fuel. This paper describes a mechanistic thermal conductivity model for MOX fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX fuel. The database used for verification consists of homogeneous MOX fuel at beginning-of-life and inhomogeneous MOX fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.

• 방사선산업학회지
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• 제6권1호
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• pp.41-47
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• 2012

A field tracer experiment using a radioisotope was conducted to analyze the transport and dispersion characteristics of pollutants in the coastal area near a Wolsung. A rod float including GPS was released to track the paths of radioisotope. NaI detector was installed to measure the released radioisotope from the boat, and measurements were performed with the real time. The measured tracer data by a field experiment can be used as the basic data for understanding the transport characteristics of pollutants and verifying numerical models near the offshore.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.837-842
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• 2019

A large amount of radioactive material was released from the Fukushima Daiichi Nuclear Power Plant (FDNPP) in 2011 and dispersed into the environment. Though seven years have passed since the Fukushima Daiichi Nuclear Power Plant accident, some parts of Japan are still under the influence of radionuclide contamination, especially Fukushima Prefecture and prefectures neighboring Fukushima Prefecture. The long-term effective doses and the contributions of each exposure pathway (5 exposure pathways) and radionuclide ($^{131}I$, $^{134}Cs$, and $^{137}Cs$) were evaluated for people living in the regions of Fukushima and neighboring prefectures in Japan using a developed dose assessment code system with Japanese specific input data. The results estimated in this study were compared with data from previously published reports. Groundshine and ingestion were predicted to contribute most significantly to the total long-term dose for all regions. The contributions of each exposure pathway and radionuclide show different patterns for certain regions of Japan.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.619-632
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• 2014

The Best Estimate Plus Uncertainty (BEPU) method has been widely used to evaluate the uncertainty of a best-estimate thermal hydraulic system code against a figure of merit. This uncertainty is typically evaluated based on the physical model's uncertainties determined by expert judgment. This paper introduces the application of data assimilation methodology to determine the uncertainty bands of the physical models, e.g., the mean value and standard deviation of the parameters, based upon the statistical approach rather than expert judgment. Data assimilation suggests a mathematical methodology for the best estimate bias and the uncertainties of the physical models which optimize the system response following the calibration of model parameters and responses. The mathematical approaches include deterministic and probabilistic methods of data assimilation to solve both linear and nonlinear problems with the a posteriori distribution of parameters derived based on Bayes' theorem. The inverse problem was solved analytically to obtain the mean value and standard deviation of the parameters assuming Gaussian distributions for the parameters and responses, and a sampling method was utilized to illustrate the non-Gaussian a posteriori distributions of parameters. SPACE is used to demonstrate the data assimilation method by determining the bias and the uncertainty bands of the physical models employing Bennett's heated tube test data and Becker's post critical heat flux experimental data. Based on the results of the data assimilation process, the major sources of the modeling uncertainties were identified for further model development.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.274-279
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• 1997

Breakthrough curves for separation of cesium and strontium binary components from aqueous solutions in fixed bed by chabazite and 13X were experimentally obtained and simulated the breakthrough curves for binary component adsorption. It is shown that agreements between model predictions and experimental data are relatively good although some deviations are observed. From engineering point of view, the simple model used here can be applied to simulate adsorption breakthrough curves satisfactorily.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2229-2236
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• 2021

In nuclear-related facilities, such as nuclear power plants, research reactors, accelerators, and nuclear waste storage sites, radiation detection, and mapping are required to prevent radiation overexposure. Sensor network systems consisting of radiation sensor interfaces and wxireless communication units have become promising tools that can be used for data collection of radiation detection that can in turn be used to draw a radiation map. During data collection, malfunctions in some of the sensors can occasionally occur due to radiation effects, physical damage, network defects, sensor loss, or other reasons. This paper proposes a reproduction strategy for radiation maps using a U-net model to compensate for the loss of radiation detection data. To perform machine learning and verification, 1,561 simulations and 417 measured data of a sensor network were performed. The reproduction results show an accuracy of over 90%. The proposed strategy can offer an effective method that can be used to resolve the data loss problem for conventional sensor network systems and will specifically contribute to making initial responses with preserved data and without the high cost of radiation leak accidents at nuclear facilities.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2007년도 학술논문요약집
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• pp.327-328
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• 2007

The deposition conditions for plating polonium have been optimized with deposition parameters such as pH, volume and temperature of the deposition and deposition time. In the tap water, the chemical yields of polonium forthe deposition solution adjusted to pH 0 were higher than those for the deposition solution adjusted to pH 2. This modified auto-deposition method made it possible to obtain reliable data of activity concentration of Po-210.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.