• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.419-433
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• 2016

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.621-631
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• 2017

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.9-18
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• 2024

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.457-469
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• 2016

During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue) test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Membrane Journal
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• v.4 no.3
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• pp.152-162
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• 1994

The photodegradation efficiency of formic acid on $TiO_2$ photocatalytic membranes was investigated. A new titania membrane reactors for purification of water combining microfiltration with photocatalytic degradation of organic compounds were developed. Titania membrane tubes(average pore size of $0.2\mu m$) were prepared by the slip casting, and porous thin films of $TiO_2$ were formed on the tube surface by the sol-gel process to increase the surface area, and consequently to increase photodegradation efficiency of organic compounds. The UV light with the wavelength of 365 nm was used as a light source for photocatalytic reactions. The photodegradation efficiency of the organic compounds was strongly dependent on the flux of the solution, the microstructure of the membrane (sol pH), and the amount of $O_2$ supplied. The effects of the primary oxidant such as $H_2O_2$ and dopants such as $Nb_2O_5$ on the photodegradation efficiency were also investigated. The results showed that more than 80% of formic acid could be degraded using membrane coated with a $TiO_2$ sol of pH 1.45. The photodegradation efficiency could be improved by about 20% when adding $H_2O_2$ in feed solution or doping $TiO_2$ membranes with $Fe_2O_3$.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.453-458
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• 2014

SiC and its composites have been considered for use as nuclear fuel cladding materials of pressurized light water reactors. In this study, a $SiC_f$/SiC composite as a constituent layer of SiC triplex fuel cladding was fabricated using a chemical vapor infiltration (CVI) process in which tubular SiC fiber preforms were prepared using a filament winding method. To enhance the matrix density of the composite layer, winding patterns, deposition temperature, and gas input ratio were controlled. Fiber arrangement and porosity were the main parameters influencing densification behaviors. Final density of the composites decreased as the SiC fiber volume fraction increased. The CVI process was optimized to densify the tubular preforms with high fiber volume fraction at a high $H_2$/MTS ratio of 20 at $1000^{\circ}C$; in this process, surface canning of the composites was effectively retarded.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2054-2063
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• 2020

The high-temperature oxidation behavior of ZrSi₂ used as a coating material for nuclear fuel cladding was investigated for developing accident-tolerant fuel cladding of light water reactors. Bulk ZrSi₂ samples were prepared by spark plasma sintering. In situ X-ray diffraction was conducted in air at 900, 1000, and 1100 ℃ for 20 h. The microstructures of the samples before and after oxidation were examined by scanning electron microscopy and transmission electron microscopy. The results showed that the oxide layer of zirconium silicide exhibited a layer-by-layer structure of crystalline ZrO₂ and amorphous SiO₂, and the high-temperature oxidation resistance was superior to that of Zircaloy-4 owing to the SiO₂ layer formed. ZrSi₂ was coated on the Zircaloy-4 tube surface using laser 3D printing, and the coated tube was oxidized for 2000 s at 1200 ℃ under a vapor/argon mixture atmosphere. The outer surface of the coated tube was hardly oxidized (10-30 ㎛), while the inner surface of the uncoated tube was significantly oxidized to approximately 300 ㎛.

• Journal of Advanced Marine Engineering and Technology
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• v.8 no.1
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• pp.85-99
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• 1984

The stability of the two-phase flow in a heated channel is of great importance in the design and operation of the boilers and light water nuclear reactors, because it can cause flow oscillations and lead to a violation of thermal limits with resultant overheating of the channels and cladding. This paper presents a systematic evaluation to the variation effects of the basic four (4) dimensionless parameters in a homogeneous equilibrium model. The flow stability is examined on the ground of static characteristic curves. The complicated transfer function of flow dynamics which gives consideration to the transport lag of density wave is derived, and the transient flow stability is analysed by applying the Nyquist stability criterion in control engineering. The analysis results summed up as follows 1. The coolant flow becomes stable in large friction number and specific flow, while it is unstabale in small friction number and flow. 2. Large phase-change number and Froude number destabilize the two-phase flow, but small numbers stabilize it. The effect to variation of phase-change number is more dominant compared with Froude number. 3. The dynamic analysis is required to hold the sufficient safety of heated channels since only static results does not keep it. The special attention could be payed in the design and operation of heat engines, because the unstaable region exists within the stable boundary at small and middle phase-change number and Froude number.