• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Proceedings of the Society of Korea Industrial and System Engineering Conference
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• 2002.05a
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• pp.25-30
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• 2002

The problem of qualitative reliability system is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, extensive environment destruction, and fatal damage of human. Therefore this study is to develop the reliability evaluation model through the normalized scoring model by the quantitative and qualitative factors considering the advanced safety factors In the Advanced Pressurized water Reactor 1400MWe(APR 1400) under uncertainty Especially, the qualitative factors considering the information and human factors for the systematic and rational justification have been closely analyzed. The reliability evaluation model can be simply applied in real fields in order to minimize the industrial accident and human error in the digitalized nuclear power plant.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.25 no.6
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• pp.23-35
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• 2002

The problem of system reliability is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, environment destruction, and fatal damage of human. Therefore the purpose of this study has developed the reliability evaluation model through the scoring model by the quantitative and qualitative factors in order to justify the evaluation considering the advanced safety factors in the Advanced Pressurized water Reactor 1400MWe(APR 1400MWe) under uncertainty. Especially, the qualitative factors considering the human, information control, and quality factors for the systematic and rational justification have been closely analyzed. The proposed model can be simply applied in real fields in order to minimize the industrial accidents in the digitalized nuclear power plant.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.91-96
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• 1996

The Dynamic Safety System (DSS) is a compute. based reactor protection system that has fail-safe nature and perform dynamic self-testing. In this paper, the implementation of DSS in PLC is presented for PWR. In order to choose adequate PLC implementation model of DSS, the reliability analysis is performed. The KO-RI unit 2 Nuclear power plant is selected as the reference plant, and the verification is carried out using the KO-RI unit 2 simulator FISA-2.

• Journal of the Earthquake Engineering Society of Korea
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• v.16 no.3
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• pp.35-42
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• 2012

We propose a probabilistic method to evaluate the uniform hazard spectra (UHS) of the soil of nuclear power plant(NPP) sites corresponding to that of a bedrock site. To do this, amplification factors on the surface of soil sites were estimated through site response analysis while considering the uncertainty in the earthquake ground motion and soil deposit characteristics. The amplification factors were calculated by regression analysis with spectral acceleration because these two factors are mostly correlated. The proposed method was applied to the evaluation of UHS for the KNGR (Korean Next Generation Reactor) and the APR1400 (Advanced Power Reactor 1400) nuclear power plant sites of B1, B4, C1 and C3. The most dominant frequency range with respect to the annual frequency of earthquakes was evaluated from the UHS analysis. It can be expected that the proposed method will improve the results of integrated risk assessments of NPPs rationally. We expect also that the proposed method will be applied to the evaluation of the UHS and of many other kinds of soil sites.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.330-338
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• 2016

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

• Journal of Hydrogen and New Energy
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• v.33 no.4
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• pp.451-461
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• 2022

CFD analysis including optimization process was conducted to design shell and tube CO₂ methanation reactor cooling system. The high-pressure saturated water flowed into the cooling system and was evaporated by heat flux from reacting tubes. The optimization process decided the gap between tubes and reactor diameter to satisfy objective functions related to temperature. The results showed that the gap and diameter reduced about 30% and 3.6% respectively. Averaged surface temperature satisfied the target value and the min-max deviation was minimized.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.109.2-109
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• 2001

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.