• Journal of the Society of Naval Architects of Korea
• /
• v.51 no.3
• /
• pp.220-230
• /
• 2014

Typical technical issues associated with Floating LNG (FLNG: FSRU and LNG FPSO) design are categorized in terms of global performance evaluation. Although many proven technologies developed through LNG carrier and oil FPSO projects are available for FLNG design, we are still faced with several technical challenges to clear for successful FLNG projects. In this paper, some of the challenges encountered during development of the floating LNG facility (i.e. LNG FPSO and FSRU) will be reviewed together with their investigated solution. At the same time, research of design improvement including new LNG-related technologies such as combined containment system will be presented to overcome the unrevealed challenges for the FLNG development.

• Journal of the Korean Society of Safety
• /
• v.15 no.2
• /
• pp.126-135
• /
• 2000

Probabilistic Safety Assessment (PSA) is an engineering analysis method to identify possible contributors to the risk from a nuclear power plant and now it has become a standard tool in safety evaluation of nuclear power plants. PSA consists of three phases named as Level 1, 2 and 3. Level 2 PSA, mainly focused in this paper, uses a step-wise approach. At first, plant damage states (PDSs) are defined from the Level 1 PSA results and they are quantified. Containment event tree (CET) is then constructed considering the physico-chemical phenomena in the containment. The quantification of CET can be assisted by a decomposition event tree (DET). Finally, source terms are quantitatively characterized by the containment failure mode. As the main benefit of PSA is to provide insights into plant design, performance and environmental impacts, including the identification of the dominant risk contributors and the comparison of options for reducing risk, this technique is expected to be applied to the industrial safety area.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1174-1183
• /
• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• Nuclear Engineering and Technology
• /
• v.47 no.5
• /
• pp.567-578
• /
• 2015

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in accident conditions. In the present study, CUPID code, which has been developed by KAERI for the analysis of transient two-phase flows in nuclear reactor components, is improved for simulating film condensation in the presence of noncondensable gases. In order to evaluate the condensate heat transfer accurately in a large system using the two-fluid model, a mass diffusion model, a liquid film model, and a wall film condensation model were implemented into CUPID. For the condensation simulation, a wall function approach with a heat/mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model, and then introduces the simulation result using the improved CUPID for a conceptual condensation problem in a large system.

• Nuclear Engineering and Technology
• /
• v.55 no.3
• /
• pp.1125-1139
• /
• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• Journal of Korea Multimedia Society
• /
• v.8 no.2
• /
• pp.154-163
• /
• 2005

Because most researches that were studied so far on XML documents used an absolute coordinate system in most of the index techniques, the update operation makes a large burden. To express the structural relations between elements, attributes and text, we need to reconstruct the structure of the coordinates. As the reconstruction process proceeds through out the entire XML document in a cascade manner, which is not limited to the current changing node, a serious performance problem may be caused by the frequent update operations. In this paper, we propose an index technique based on extensible index that does not cause serious performance degradations. It can limit the number of node to participate in reconstruction process and improve lots of performance capacities on the whole. And extensible index performs the containment relationship query by the simple expression using SQL statement.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.24 no.2
• /
• pp.193-200
• /
• 2011

A BOR(Boil-Off Rate) prediction model for the NO96 membrane-type LNG insulation containment filled with superlite powders during laden voyage is presented in this paper. Finite element model for the unsteady-state heat transfer analysis is constructed by considering the air and water conditions and by employing the homogenization method to simplify the complex insulation material composition. BOR is evaluated in terms of the total amount of heat invaded into LNGCC and its variation to the major variables is investigated by the parametric heat transfer analysis. Based upon the parametric results, a BOR prediction model which is in function of the LNG tank size, the insulation layer thickness and the powder thermal conductivity is derived. Through the verification experiment, the accuracy of the derived prediction model is justified such that the maximum relative difference is less than 1% when compared with the direct numerical estimation using the FEM analysis.

• Nuclear Engineering and Technology
• /
• v.54 no.9
• /
• pp.3310-3316
• /
• 2022

Owing to the rising concerns about the safety of nuclear power plants (NPP), it is essential to study the venturi scrubber in detail, which is a key component of the filtered containment venting system (FCVS). FCVS alleviates the pressurein containment byfiltering and venting out the contaminated air. Themain objective of this research was to perform a CFD investigation of different configurations of a circular, non-submerged, self-priming venturi scrubber to estimate and improve the performance in the removal of elemental iodine from the air. For benchmarking, a mass transfer model which is based on two-film theory was selected and validated by experimental data where an alkaline solution was considered as the scrubbing solution. This mass transfer model was modified and implemented on a unique formation of two self-priming venturi scrubbers in series. Euler-Euler method was used for two-phase modeling and the realizable K-ε model was used for capturing the turbulence. The obtained results showed a remarkable improvement in the removal of radioactive iodine from the air using a series combination of venturi scrubbers. The removal efficiency was improved at every single data point.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.19 no.1 s.71
• /
• pp.93-103
• /
• 2006

It is absolutely essential that safety assessment of the containment buildings during service life because containment buildings are last barrier to protect radioactive substance due to the accidents. Therefore, this study describes an enhanced degenerated shell finite element(FE) which has been developed for nonlinear FE analysis of reinforced concrete(RC) containment buildings with elasto-plastic material model. For the purpose of the material nonlinear analysis, Drucker-Prager failure criteria is adapted in compression region and material parameters which determine the shape of the failure envelop are derived from biaxial stress tests. Reissner-Mindlin(RM) assumptions are adopted to develop the degenerated shell FE so that transverse shear deformation effects is considered. However, it is found that there are serious defects such as locking phenomena in RM degenerated shell FE since the stiffness matrix has been overestimated in some situations. Therefore, shell formulation is provided in this paper with emphasis on the terms related to the stiffness matrix based on assumed strain method. Finally, the performance of the present shell element to analysis RC containment buildings is tested and demonstrated with several numerical examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.

• Journal of the Korea Institute of Building Construction
• /
• v.15 no.4
• /
• pp.425-431
• /
• 2015

Since the construction of Kori 1 was completed in 1978, the construction projects for nuclear power plant are increasingly expanded into domestic and foreign sites. However, some of construction sites of nuclear power plant have the problems of process delay and cost loss due to lack of ability of risk management. The construction of reactor containment building in nuclear power plant is especially dotted with many risk factors because it needs professional skills and large-scale resources due to long duration compared with different construction phase. Therefore, it needs the study that analyzes risk factors expected in construction of reactor containment building and suggests way of stable performance of projects. So, this study assesses risk factors of construction of reactor containment building. For the objectives, this study uses survey for group of minority specialists of 36 experts. The risks of 24 factors is classified by criterions of process, cost, safety, and quality and the results of assessment is analyzed by analytic hierarchy process. As the results, the importance and priority of risk factors classified by each criterion were calculated and the applicability of analytic hierarchy process was identified to analyze risk factors of nuclear power plant construction. These will be baseline data for risk management in construction phase of reactor containment building.