• Nuclear Engineering and Technology
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• 제36권5호
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• pp.420-429
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• 2004

The local power density should be estimated accurately to prevent fuel rod melting. The local power density at the hottest part of a hot fuel rod, which is described by the power peaking factor, is more important information than the local power density at any other position in a reactor core. Therefore, in this work, the power peaking factor, which is defined as the highest local power density to the average power density in a reactor core, is estimated by fuzzy neural networks using numerous measured signals of the reactor coolant system. The fuzzy neural networks are trained using a training data set and are verified with another test data set. They are then applied to the first fuel cycle of Yonggwang nuclear power plant unit 3. The estimation accuracy of the power peaking factor is 0.45% based on the relative $2_{\sigma}$ error by using the fuzzy neural networks without the in-core neutron flux sensors signals input. A value of 0.23% is obtained with the in-core neutron flux sensors signals, which is sufficiently accurate for use in local power density monitoring.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.573-580
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• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.212-221
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• 1998

In the present study, a method for the dynamic analysis of a reactor core is developed. Peak responses for the motions induced from earthquake are obtained for a core model. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are investigated. Prediction of fuel assembly stress during an earthquake requires development of a fuel assembly stress analysis model capable of interfacing with the models and results discussed in the dynamic analysis of a reactor core. This analysis uses beam characteristics which describe the overall fuel assembly response. The stress analysis method and its application for the case of an increased seismic level are also presented.

• 한국분말재료학회지
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• 제25권2호
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• pp.132-136
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• 2018

Core/shell CdSe/ZnS quantum dots (QDs) are synthesized by a microfluidic reactor-assisted continuous reactor system. Photoluminescence and absorbance of synthesized CdSe/ZnS core/shell QDs are investigated by fluorescence spectrophotometry and online UV-Vis spectrometry. Three reaction conditions, namely; the shell coating reaction temperature, the shell coating reaction time, and the ZnS/CdSe precursor volume ratio, are combined in the synthesis process. The quantum yield of the synthesized CdSe QDs is determined for each condition. CdSe/ZnS QDs with a higher quantum yield are obtained compared to the discontinuous microfluidic reactor synthesis system. The maximum quantum efficiency is 98.3% when the reaction temperature, reaction time, and ZnS/CdSe ratio are $270^{\circ}C$, 10 s, and 0.05, respectively. Obtained results indicate that a continuous synthesis of the Core/shell CdSe/ZnS QDs with a high quantum efficiency could be achieved by isolating the reaction from the external environment.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.150-164
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• 2004

According to the Korean long-term plan for nuclear technology development, KAERI is conducting a few R&D projects related to the proliferation-resistant back-end fuel cycle. The R&D activities for the back-end fuel cycle are reviewed in this work, especially focusing on the study of the partitioning and transmutation(P&T) of long-lived radionuclides. The P&T study is currently being carried out in order to develop key technologies in the areas of partitioning and transmutation. The partitioning study is based on the development of pyroprocessing such as electrorefining and electrowinning because they can be adopted as proliferation-resistant technologies in the fuel cycle. In this study, various behaviors of the electrodeposition of uranium and rare earth elements in the LiCl-KCl electrorefining system have been examined through fundamental experimental work. As for the transmutation system, KAERI is studying the HYPER (HYbrid Power Extraction Reactor), a kind of subcritical reactor which will be connected with a proton accelerator. Up to now, a conceptual study has been carried out for the major elemental systems of the subcritical reactor such as core, transuranic fuel, long-lived fission product target, and the Pb-Bi cooling system, etc. In order to enhance the transmutation efficiency of the transuranic elements as well as to strengthen the reactor safety, the reactor core was optimized by determining its most suitable subcriticality, the ratio of height/diameter, and by introducing the concepts of optimum core configuration with a transuranic enrichment as well as a scattered reloading of the fuel assemblies.

• 한국안전학회지
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• 제28권3호
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1181-1190
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• 2009

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.377-383
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• 2019

Targeting at simulating the application of thorium-uranium (TU) fuel in the CANDU-6 reactor, this paper analyzes the process using the code DRAGON/DONJON where the discrete TU fuel pins are applied in the CANDU-6 reactor under the time-average equilibrium refueling. The results show that the coolant void reactivity of the assembly analyzed in this paper is lower than that of 37-element bundle cell with natural uranium and 37-element bundle cell with mixed TU fuel pins; that the max time-average channel/bundle power of the core meets the limits - less than 6700kW/860 kW; that the fuel conversion ratio is higher than that of the CANDU-6 reactor with natural uranium; and that the exit burnup increases to 13400 MWd/tU. Thus, the simulation in this paper with the fuel in the 37-element bundle cell using discrete TU fuel pins can be considered to be applied in CANDU-6 reactor with adequate modifications of the core structure and operating modes.