• Journal of The Korean Society of Grassland and Forage Science
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• v.10 no.2
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• pp.77-83
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• 1990

A three year's field experiment was carried out on newly reclaimed tidal saline soils to evaluate the salt tolerance and growht characteristics, and their relationship to dry matter production and nutrient quality of main selected pasture species. Nine temperate grasses (14 varieties) and two forage crops (sorghum and pearl millet) were grown under different mulching treatments with medium sand and red-yellow soils (fine loamy materials of Typic Hapludults) from 1986 to 1988. Tall wheatgrass, tall fescue, reed canarygrass and alfalfa showed a good tolerance to soil salinity, especially tall wheatgrass (cv. Alkar) produced 19.6 ton/ha dry matter yield annualy under mulching treatment with medium sand depth in lcm. Pearl millet (cv. Gahi-3) was also evaluated as a salt tolerable forage species. Under salt stress in newly reclaimed tidal lands, plant showed a decrease in the assimirable leaf area (LA) as well as specific leaf area (SP. LA) and a low leaf weight ratio(LWR), and it resulted in a low concentration of crude protein and low digestible dry matter contents. Absorption of macro and micro elements in the plant on tidal lands was increased markedly.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.608-619
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• 2010

A multi-scale analysis of water-cooled reactor thermalhydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermalhydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3264-3274
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• 2021

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.304-313
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• 2021

Nickel base Alloy X-750, which is used as fastener parts in light-water reactor (LWR), has experienced many failures by environmentally assisted cracking (EAC). In order to improve the reliability of passive components for nuclear power plants (NPP's), it is necessary to study the failure mechanism and to predict crack growth behavior by developing a probabilistic failure model. In this study, The Bayesian inference was employed to reduce the uncertainties contained in EAC modeling parameters that have been established from experiments with Alloy X-750. Corrosion fatigue crack growth rate model (FCGR) was developed by fitting into Paris' Law of measured data from the several fatigue tests conducted either in constant load or constant ΔK mode. These parameters characterizing the corrosion fatigue crack growth behavior of X-750 were successfully updated to reduce the uncertainty in the model by using the Bayesian inference method. It is demonstrated that probabilistic failure models for passive components can be developed by updating a laboratory model with field-inspection data, when crack growth rates (CGRs) are low and multiple inspections can be made prior to the component failure.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3301-3312
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• 2023

This study incorporates a high-fidelity transient analysis solver based on multigroup CMFD in the MOC code STREAM. Transport modeling with heterogeneous geometries of the reactor core increases computational cost in terms of memory and time, whereas the multigroup CMFD reduces the computational cost. The reactor condition does not change at every time step, which is a vital point for the utilization of CMFD. CMFD correction factors are updated from the transport solution whenever the reactor core condition changes, and the simulation continues until the end. The transport solution is adjusted once CMFD achieves the solution. The flux-weighted method is used for rod decusping to update the partially inserted control rod cell material, which maintains the solution's stability. A smaller time-step size is needed to obtain an accurate solution, which increases the computational cost. The adaptive step-size control algorithm is robust for controlling the time step size. This algorithm is based on local errors and has the potential capability to accept or reject the solution. Several numerical problems are selected to analyze the performance and numerical accuracy of parallel computing, rod decusping, and adaptive time step control. Lastly, a typical pressurized LWR was chosen to study the rod-ejection accident.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.819-823
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• 2024

The main criteria used in NEI 18-04 to define SSCs as risk-significant include (1) the SSC is required to keep all LBEs within the F-C target, and (2) the total frequency with the SSC failed exceeds 1% of the limit for at least one of the three cumulative risk metrics used for evaluating the integrated plant risk. The first one is a reasonable criterion in determining the risk significant SSCs. However, the second criterion may not be adequate to serve the purpose of determining the risk significance of SSCs. In the second criterion, the cumulative risk metric values representing the integrated plant risk (less the preventive and mitigative effects of the SSC being evaluated) are compared to a risk limit that represents a very small contribution to the overall integrated plant risk, which corresponds appropriately to the contributions from individual SSCs. The easiest approach to redefine the NEI 18-04 definition of risk-significant SSCs in relation to the integrated plant risk metrics is to compare the difference, between the risk metric value calculated with the SSC failed and the risk metric value calculated with the SSC credited, with 1% of the risk limit established for the integrated plant risk metrics.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Journal of The Korean Society of Grassland and Forage Science
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• v.15 no.1
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• pp.73-79
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• 1995

A simple conservation technique baled silage making of selected froage materials was discussed in Suwon and in Muan county during 1991 - 1992. Eleven species of forage crops and pasture grasses(maize. sorghum, pearl millet, barnyardgrass, rye, barley, spring oat, Italian ryegrass, orchardgrass, alfalfa and grass-legume pasture mixtures) were harvested at different growth stage from young plant to maturity and baled in a self constructed square baling chamber, and wrapped in a 0.05mm thick polyethylene plastic film. Each bales measured by 90cm long, 60cm wide and 50cm height and weighted between 15~20kg in dry matter basis. physio-molphologcal characteristics of the plants, leaf weight ratio(LWR), leaf area ratio(LAR), stalk ratio (SR), stalk hardness(SH) and other growth parameters, were analysed and were used as a parameter to evaluate the suitability of materials for baling. Italian ryegrass including orchardgrass, alfalfa and pasture mixtures produced high quality baled silage. Silage quality point(F1ieg-point) of Italian ryegrass was improved from 63 point in crushed custom silage to 75 piont in baled silage. Meterial of grass-legume pasture mixtures showed 55 point in crushed silage and 67 point in baled silage. Fodder rye, barley, spring oat and barnyardgrass were also evaluated as a good materials for baled silage making. On the other hand, meize including sorghum and peral millet were evaluated as a not suitable materals for baling due to its high value of SR and SH. Quality of maize was excellent with 88 point in clushed silage and medium with 47 point in baled silage making.