• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.637-642
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• 1996

This paper presents the analysis results for the core degradation processes and the fission product release of the PHEBUS FPT0 experiment using MELCOR1.8.3. The objective of this study is to assess models associated with the core damage and fission product behavior in MELCOR. The calculation results were much improved through sensitivity studies. Thermal/hydraulic behavior in the core and the circuit was well predicted under the intact core geometry. In non-eutectic model case. the UO$_2$ dissolution model in the MELCOR always showed such a tendency that the resulting dissolved UO$_2$ mass was small at the highly oxidized condition due to the model logic. Total H$_2$ generation mass was underpredicted because the stiffner was not modeled and the liner in the shroud was not allowed to be oxidized in MELCOR. Some difficulties were found in modeling the activation product were solved by manipulating the RN input associated with the initial fission product inventory. These problem were occurred because there are no control rod model in MELCOR. Generally the fission product release ratio showed a similar trend compared with the measured data except the activation product. which have no model to simulate in MELCOR.

• Journal of the Korean Ceramic Society
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• v.46 no.2
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• pp.123-130
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• 2009

To understand the formation of core-shell structure in $BaTiO_3$ (BT) grains in multilayer ceramic capacitors, specimens were prepared with BT powders mixed with Y and Mg, and their microstructures were investigated with scanning electron microscopy, x-ray diffractometry, and transmission electron microscopy. Microstructural investigation showed that Y dissolved easily in BT lattice to a certain depth inside of the grain, whereas Mg tended to stay at grain boundaries rather than become incorporated into BT. It was considered that in case of Y and Mg addition in a proper ratio, Y could play a dominant role in the formation of shell leading to a slight dissolution of Mg in the shell. Next, the effects of ball-milling conditions on the core-shell formation were studied. As the ball-milling time increased, the milled powders did not show a significant change in size distribution but rather an increase of residual strain, which was attributed to the milling damage. The increase in milling damage facilitated the shell formation, leading to the increased shell portion in the core-shell grain.

• Steel and Composite Structures
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• v.44 no.1
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• pp.65-79
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• 2022

The main goal of this paper is to study the vibration of damaged core laminated annular plates with FG face sheets based on a three-dimensional theory of elasticity. The structures are made of a damaged isotropic core and two external face sheets. These skins are strengthened at the nanoscale level by randomly oriented Carbon nanotubes (CNTs) and are reinforced at the microscale stage by oriented straight fibers. These reinforcing phases are included in a polymer matrix and a three-phase approach based on the Eshelby-Mori-Tanaka scheme and on the Halpin-Tsai approach, which is developed to compute the overall mechanical properties of the composite material. In this study the effect of microcracks on the vibrational characteristic of the sandwich plate is considered. In particular, the structures are made by an isotropic core that undergoes a progressive uniform damage, which is modeled as a decay of the mechanical properties expressed in terms of engineering constants. These defects are uniformly distributed and affect the central layer of the plates independently from the direction, this phenomenon is known as "isotropic damage" and it is fully described by a scalar parameter. Three complicated equations of motion for the sectorial plates under consideration are semi-analytically solved by using 2-D differential quadrature method. Using the 2-D differential quadrature method in the r- and z-directions, allows one to deal with sandwich annular plate with arbitrary thickness distribution of material properties and also to implement the effects of different boundary conditions of the structure efficiently and in an exact manner. The fast rate of convergence and accuracy of the method are investigated through the different solved examples. The sandwich annular plate is assumed to have any arbitrary boundary conditions at the circular edges including simply supported, clamped and, free. Several parametric analyses are carried out to investigate the mechanical behavior of these multi-layered structures depending on the damage features, through-the-thickness distribution, and boundary conditions.

• Journal of the Korean Society of Safety
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• v.34 no.3
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1217-1233
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• 2018

Following a surge of interest in multi-unit risk in the last few years, many recent studies have suggested methods for multi-unit probabilistic safety assessment (MUPSA) and addressed several related aspects. Most of the existing studies though focused on two-unit nuclear power plant (NPP) sites or used rather simplified probabilistic safety assessment (PSA) models to demonstrate the proposed approaches. When considering an NPP site with three or more units, some approaches are inapplicable or yield very conservative results. Since the number of such sites is increasing, there is a strong need to develop and validate practical approaches to the related MUPSA. This article provides several detailed approaches that are applicable to multi-unit Level 1 PSA for sites with up to six or more reactor units. To validate the approaches, a multi-unit Level 1 PSA model is developed and the site core damage frequency is estimated for each of four representative multi-unit initiators, as well as for the case of a simultaneous occurrence of independent single-unit initiators in multiple units. For this purpose, an NPP site with six identical OPR-1000 units is considered, with full-scale Level 1 PSA models for a specific OPR-1000 plant used as the base single-unit models.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.66-73
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• 2021

This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.

• Journal of the Korean Society of Industry Convergence
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• v.10 no.3
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• pp.149-155
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• 2007

Damage assessment and Damage index for RC members failed in flexure was investigated by using the nonlinear finite element analysis, included with nonlocal constitutive law, which is analyzed for the localization of the failure on the post-peak region. In the nonlcal constitutive law, The local strains obtained at gauss points were averaged over a particular length, i.e. characteristic length and it was used to evaluate the damage of RC column member. As the analysis results, The value of nonlocal strain shows less mesh sensibility. In the damage assessment, It was confirmed that evaluations of damage of RC member were able to use nonlocal compressive strain on a cover concrete and a core concrete of the member. Moreover it was confirmed that damage process for the statically indeterminate structure was able to evaluate the damage context of the component members of the structure.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.915-921
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• 2019

Radiation damage due to neutrons has been calculated in Ni-based alloys in Wolsong CANDU reactor environments. Two damage parameters are considered: displacement damage, and transmutation gas production. We used the SPECTER and SRIM computer codes in quantifying radiation damage. In addition, damage caused by Ni two-step reactions was considered. Estimations were made for the annulus spacers in a CANDU reactor that are located axially along a fuel channel and made of Inconel X-750. The calculation results indicate that the transmutation gas production from the Ni two-step reactions is predominant as the effective full power year increases. The displacement damage due to recoil atoms produced from Ni two-step reactions accounts for over 30% out of the total displacement damage.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.592-597
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• 1996

The KALIMER core, of which nuclear design is largely governed by inherent safety and reactivity control issues, is fueled with metallic fuel, and the initial core will be loaded with 20% enriched Uranium metal fuel. KALIMER safety design objectives include the accommodation of unprotected, ATWS events without operator action, and without the support of active shutdown, shutdown heat removal, or any automatic system without damage to the plant and without jeopardizing public safety. The transient analysis of the core designs has been focused on severe events to assess the margins in the design, and ATWS events are the most severe events that must be accommodated by the KALIMER design. The ATWS performance has been evaluated for the preliminary initial core design of KALIMER with a particular emphasis on the inherent negative reactivity feedback effects, including the Doppler, sodium density, fuel axial expansion, core radial expansion, and control rod driveline expansion. Results show that the Uranium metal core design meets the temperature limits with margin.