• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.224-233
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• 2017

In Korea, attempts have been made to develop high strength concrete for the safety and design life improvement of nuclear power plants. In this study, the cracking potentials of nuclear power plant-high strength concretes (NPP-HSCs) containing industrial by-products with W/B 0.34 and W/B 0.28, which are being reviewed for their application in the construction of containment structures, were evaluated through autogenous shrinkage, unrestrained drying shrinkage, and restrained drying shrinkage experiments. The cracking potentials of the NPP-HSCs with W/B 0.34 and W/B 0.28 were in the order of 0.34FA25 > 0.34FA25BFS25 > 0.34BFS50 > 0.34BFS65SF5 and 0.28FA25SF5 >> 0.28BFS65SF5 > 0.28BFS45SF5 > 0.28 FA20BFS25SF5, respectively. The cracking potentials of the seven mix proportions excluding 0.28FA25SF5 were lower than that of the existing nuclear power plant concrete; thus, the durability of a nuclear power plant against shrinkage cracking could be improved by applying the seven mix proportions with low cracking potentials.

• Korean Journal of Construction Engineering and Management
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• v.5 no.2 s.18
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• pp.144-152
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• 2004

Defining value as a measure of how well the project value objectives are met, Value Management Process (VMP) is considered to be any management effort or process that can proactively pursue one or more project objectives (i.e., security/safety, cost effectiveness, schedule optimization, and risk containment). The collection of 44 VMPs has been established based on a rigorous effort conducted by Construction Industry Institute (CII). Because varying circumstances on each project determine the level of suitability, it is crucial to identify which VMP should be implemented on a particular project. The current VMP selection process is primarily based on human intuition. The main objective of this paper is to provide a systematic method to facilitate the usage of VMPs on a particular project. This paper identified and quantified the selection principles (i.e., targeted value objectives, timing of initiation, project characteristics, and relative impact). The data collected from industry practitioners and VMP experts characterized each VMP in terms of the magnitude of benefit. An automated selection tool by Visual Basic Application (VBA) on MS Excel TM, was developed and proved its validity. As a pioneering study, this paper provides a comprehensive and structured knowledge on the subject of VMPs. From the industry's perspective, the automated selection tool, the premier of this study, contributes the facilitation of the VMP implementations in the construction industry thereby maximizing the potential benefits to a particular project.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3298-3313
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• 2021

The objective of this study is to estimate the occurrence probabilities of severe accident phenomena based on a fuzzy elicitation technique. Normally, it is difficult to determine these probabilities due to the lack of information on severe accident progression and the highly uncertain values currently in use. In this case, fuzzy set theory (FST) can be best exploited. First, questions were devised for expert elicitation on technical issues of severe accident phenomena. To deal with ambiguities and the imprecision of previously developed (reference) probabilities, fuzzy aggregation methods based on FST were employed to derive the occurrence probabilities of severe accidents via four phases: 1) choosing experts, 2) quantifying weighting factors for the experts, 3) aggregating the experts' opinions, and 4) defuzzifying the fuzzy numbers. In this way, this study obtained expert elicitation results in the form of updated occurrence probabilities of severe accident phenomena in the OPR-1000 plant, after which the differences between the reference probabilities and the newly acquired probabilities using fuzzy aggregation were compared, with the advantages of the fuzzy technique over other approaches explained. Lastly, the impact of applying the updated severe accident probabilities on containment integrity was quantitatively investigated in a Level 2 PSA model.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1052-1065
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• 2024

Advanced Pressurized Water Reactors (APWRs) and Boiling Water Reactors (BWRs) employ a suppression pool as a heat sink to prevent containment overpressure. Steam can be discharged into the pool through multi-hole spargers or blowdown pipes in both normal and accident conditions. Direct Contact Condensation (DCC) creates sources of momentum and heat. The competition between these two sources determines the development of thermal stratification or mixing of the pool. Thermal stratification is of safety concern as it reduces the cooling capability compared to a completely mixed pool condition. In this work we develop a scaling approach to prediction of the thermal stratification in a water pool induced by steam injection through spargers. Experimental data obtained from large-scale pool tests conducted in the PPOOLEX and PANDA facilities, as well as simulation results obtained using validated codes are used to develop the scaling. Two injection orientations, namely radial injection through multi-hole Sparger Head (SH) and vertical injection through Load Reduction Ring (LRR), are considered. We show that the erosion rate of the cold layer can be estimated using the Richardson number. In this work, scaling laws are proposed to estimate both the (i) transient erosion velocity and (ii) the stable position of the thermocline. These scaling laws are then implemented into a 1D model to simulate the thermal behavior of the pool during steam injection through the sparger.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2079-2091
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• 2024

In this paper, a model is developed for calculating pH in the presence of organic impurities due to dissolution of paint and/or continuous injection of organic impurities in the sump. The model is implemented in the AnCheBi code for the analysis of chemical behaviors of the iodine in the containment when the pH changes during a severe accident. Validation of the model is performed with P10T2 and P11T1 experiments carried out by AECL in Canada under the BIP project. Importance analyses of the pH calculation model in the AnCheBi code are then performed with the aforementioned experimental data via Latin hypercube sampling on the reaction coefficients, sensitivity analyses of AnCheBi, and calculation of the correlation coefficients between the reaction coefficients and figure of merits (the pH and the concentrations of the various iodine species). From the importance analyses, we provide the sensitivity of the pH calculation model to the change of pH and the concentrations of the various iodine species and the reaction coefficients related with the dominant phenomena underlying the change of pH and the concentrations of the species.

• Journal of the Earthquake Engineering Society of Korea
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• v.8 no.5 s.39
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• pp.79-89
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• 2004

This paper presents the results of experimental studies on the equipment isolation effect in the nuclear containment. For this purpose, shaking table tests were performed. The purpose of this study is enhancement of seismic safety of equipment in the Nuclear Power Plant. The isolation system, known as Friction Pendulum System (FPS), combines the concepts of sliding bearings and pendulum motion was selected. Peak ground acceleration, bidirectional motion, effect of vertical motion and frequency contents of selected earthquake motions were considered. As a result, these are founded that the vertical motion of seismic wave affect to the base isolation and the isolation effect decreased in case of near fault earthquake motion.