• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3100-3111
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• 2021

Numerical modeling for the safety-related equipment used in a nuclear power plant (i.e., cabinet facilities) plays an essential role in seismic risk assessment. A full finite element model is often time-consuming for nonlinear time history analysis due to its computational modeling complexity. Thus, this study aims to generate a simplified model that can capture the nonlinear behavior of the electrical cabinet. Accordingly, the distributed plasticity approach was utilized to examine the stiffness-degradation effect caused by the local buckling of the structure. The inherent dynamic characteristics of the numerical model were validated against the experimental test. The outcomes indicate that the proposed model can adequately represent the significant behavior of the structure, and it is preferred in practice to perform the nonlinear analysis of the cabinet. Further investigations were carried out to evaluate the seismic behavior of the cabinet under the influence of the constitutive law of material models. Three available models in OpenSees (i.e., linear, bilinear, and Giuffre-Menegotto-Pinto (GMP) model) were considered to provide an enhanced understating of the seismic responses of the cabinet. It was found that the material nonlinearity, which is the function of its smoothness, is the most effective parameter for the structural analysis of the cabinet. Also, it showed that implementing nonlinear models reduces the seismic response of the cabinet considerably in comparison with the linear model.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Coupled systems mechanics
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• v.7 no.6
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• pp.731-753
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• 2018

The recent seismic events have led to concerns on safety and vulnerability of Reinforced Concrete Moment Resisting Frame "RC-MRF" buildings. The seismic design demands are greatly dependent on the computational tools, the inherent assumptions and approximations introduced in the modeling process. Thus, it is essential to assess the relative importance of implementing different modeling approaches and investigate the computed response sensitivity to the corresponding modeling assumptions. Many parameters and assumptions are to be justified for generation effective and accurate structural models of RC-MRF buildings to simulate the lateral response and evaluate seismic design demands. So, the present study aims to develop reliable finite element model through many refinements in modeling the various structural components. The effect of finite element modeling assumptions, analysis methods and code provisions on seismic response demands for the structural design of RC-MRF buildings are investigated. where, a series of three-dimensional finite element models were created to study various approaches to quantitatively improve the accuracy of FE models of symmetric buildings located in active seismic zones. It is shown from results of the comparative analyses that the use of a calibrated frame model which was made up of line elements featuring rigid offsets manages to provide estimates that match best with estimates obtained from a much more rigorous modeling approach involving the use of shell elements.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1394-1405
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• 2022

The importance of ensuring the inherent safety and security has been more emphasized in recent years to demonstrate the integrity of nuclear facilities under external human-induced events (e.g. aircraft crashes). This work suggests a simulation methodology to effectively evaluate the impact of a commercial aircraft engine onto a dry storage facility. A full-scale engine model was developed and verified by Riera force-time history analysis. A reinforced concrete (RC) structure of a dry storage facility was also developed and material behavior of concrete was incorporated using three constitutive models namely: Continuous Surface Cap, Winfrith, and Karagozian & Case for comparison. Strain-based erosion limits for concrete were suitably defined and the local responses were then compared and analyzed with empirical formulas according to variations in impact velocity. The proposed methodology reasonably predicted such local damage modes of RC structure from the engine missile, and the analysis results agreed well with the calculations of empirical formulas. This research is expected to be helpful in reviewing the dry storage facility design and in the probabilistic risk assessment considering diverse impact scenarios.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.386-392
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• 2021

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file. The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units. A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

• The Korean Journal of Emergency Medical Services
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• v.25 no.3
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• pp.163-178
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• 2021

Purpose: This study explored a plan for improving the overall preventive activity of Korean Citizen-Corps-Active-in-Disaster (CAIND). Methods: The study used the documentary survey and the technical approach methods. This way, detailed information regarding civil anti-disaster organizations was found in scholarly monographs, specialty publications, and master's/doctoral dissertations. It further utilized the statistical yearbooks between 2015 and 2018 from the Korean Ministry of Public Administration, Security, and Fire Service, and the National Statistical Office, to discover practical problems through a statistical analysis. Volunteer activities being inherent, related issues were reviewed at the same time to for purposes of clarifying the characteristics of disaster prevention activity by CAIND. Results: The study provided four major suggestions for improvement. First, the quota management system of Korean CAIND considering the characteristics of rural areas should be supplemented. Second, the education and training systems of Korean CAIND should be established to satisfy regional conditions. Third, management members' readership competency in operating organizations, including that of the Korean CAIND captain in charge, should be strengthened. Fourth, the reward system for Korean CAIND activities should be improved. Conclusion: In the future, the results of this study are expected to be utilized as a basic data to develop Korean CAIND.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1560-1569
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• 2022

Liquid fueled molten salt reactors (MSRs) have seen renewed interest because of their inherent safety features, higher thermal efficiency and potential for efficient thorium utilisation for power generation. Thorium fluoride is one of the salts used in liquid fueled MSRs employing Th-U cycle. In the present study, ThF₄ was prepared by hydro-fluorination of ThO₂ using anhydrous HF gas. Process parameters viz. bed depth, hydrofluorination time and hydrofluorination temperature, were optimized for the preparation of ThF₄ in a static bed reactor setup. The products were characterized with X-Ray diffraction and experimental conditions for complete conversion to ThF₄ were established which also corroborated with the yield values. Hydrofluorination of ThO₂ at 450 ℃ for half an hour at a bed depth of 6 mm gave the best result, with a yield of about 99.36% ThF₄. No unconverted oxide or any other impurity was observed. Rietveld refinement was performed on the XRD data of this ThF₄, and Chi2 value of 3.54 indicated good agreement between observed and calculated profiles.

• 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.

• Journal of the Korean Institute of Gas
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• v.13 no.1
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• pp.14-20
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• 2009

As the consumption of LNG has increased, the capacity and number of LNG facilities are getting bigger and bigger. Such circumstances supports the need for a dedicated risk analysis model to help review and check major issues of the safer construction and operation of LNG storage facilities systematically. Therefore this study suggests an appropriate risk analysis model that enables us to evaluate hazards of LNG storage facilities more easily and systematically, and then to use its result in siting, design and construction stages of the facilities. ill order to develop the model, lots of existing studies and domestic and foreign codes and standards were fully reviewed and a series of case studies also were carried out. The suggested model consists of 4-stage evaluations: in selecting a site, in determining a layout, in designing and constructing the facilities, and in operating them. This model also suggests the weather condition necessary for estimating the consequence of accident-scenarios, and the easy, systematic approach to the analysis of their probability. We expect that the model may help secure LNG storage facilities' inherent safety in determining their site and layout.