• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.285-296
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• 1992

The conceptual design of the Plant Operations Regulator (POR) is presented for the pressurized water reactor plants. The POR is a digital supervisory limitation control system The POR assures that the plant does not exceed the operating limits by regulating the plant operations through monitoring the operating margins of the critical parameters. The POR is aimed at increasing the operating flexibility which allows the nuclear plant to meet the grid demand in very efficient manner. It responds to the grid demand without penalizing plant availability by limiting the load demand or by modifying the plant control schemes when the operating limits are approached or violated. The POR design concepts were tested using simulation responses of the 1000 MWe pressurized water reactors, Yonggwang Units 3 & 4. The simulation results illustrate that the POR can be used to improve operating flexibility.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.223-236
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• 1977

An integral computer code has been developed for a mechanical and thermal design and performance analysis of an oxide fuel rod in a pressurized water reactor. The code designated as FROD 1.0 takes into account the phenomena of radial power depression within the pellet, cracking, densification and swelling of the pellet, fission gas release, clad creep, pellet-clad contact, heat transfer to coolant and buildup of corrosion layers on the clad surface. The FROD 1.0 code yields two-dimensional temperature distributions, dimensional changes, stresses, and internal pressure of a fuel rod as a function of irradiation time within a reasonable computation time. The code may also be used for the analyses of oxide fuel rods in other thermal reactors. As an application of FROD 1.0 the behavior of fuel rod loaded in the first core of Go-ri Nuclear Power Plant Unit 1 is predicted for the two power histories corresponding to steady state operation and Codition II of the ANS Classification. The results are compared with the design criteria described in the Final Safety Analysis Report and a discrepancy between these two values is discussed herein.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.296-307
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• 2020

Permanent magnet sodium flowmeter (PMSF) have been used to measure the sodium flow in fast breeder reactors. Due to the effects of irradiation, thermal cycling, time lapse, etc., the magnetic flux density of the PMSF will decrease after being used in the reactor for a period of time. Therefore, it must be calibrated regularly. But some flowmeters that immersed in sodium cannot be removed for an off-line calibration, so the on-line calibration is required. However, the best online calibration accuracy of PMSF using cross-correlation analysis method was 2.0-level without considering the repeatability. In order to further improve this work, the operational principle of the transducer in PMSF is analyzed and the design principle of the transducer is proposed. The transducers were tested on the sodium flow loop to collect the experimental data. The signal characteristics are analyzed from the time and frequency domains, respectively. The cross-correlation analysis method based on biased estimation is adopted to obtain the flow rate. The verification experimental results showed that the measurement accuracy is 1.0-level when the flow velocity is above 0.5 m/s, and the measurement accuracy is 3.0-level when the flow velocity is in the range of 0.2 m/s to 0.5 m/s.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.337-343
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• 2020

Oxide dispersed strengthened (ODS) steel is an important candidate for Gen-IV reactors. Oxide embedded in Fe can help to trap irradiation defects and enhances the strength of steel. It was observed in this study that the size of oxide has a profound impact on the depinning mechanism. For smaller sizes, the oxide acts as a void; thus, letting the dislocation bypass without any shear. On the other hand, oxides larger than 2 nm generate new dislocation segments around themselves. The depinning is similar to that of Orowan mechanism and the strengthening effect is likely to be greater for larger oxides. It was found that higher shear deformation rates produce more fine-tuned stress-strain curve. Both molecular dynamics (MD) simulations and BKS (Bacon-Knocks-Scattergood) model display similar characteristics whereby establishing an inverse relation between the depinning stress and the obstacle distance. It was found that (110)_oxide || (111)_Fe (oriented oxide) also had similar characteristics as that of (100)_oxide || (111)_Fe but resulted in an increased depinning stress thereby providing greater resistance to dislocation bypass. Our simulation results concluded that critical depinning stress depends significantly on the size and orientation of the oxide.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.344-351
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• 2020

AgInCd alloy is widely used as neutron absorber in nuclear reactors. However, the AgInCd control rods may fail during service due to the irradiation swelling. In the present study, a calculational method is proposed to calculate the composition change of the AgInCd absorber. Calculated results show that neutron fluence has significant impact on the chemical compositions. Ag and In contents gradually decrease while Cd and Sn conversely increases from the center to the rim of AgInCd absorber due to the depression of neutron flux. The composition change at the surface is higher almost two times than that at the center. Based on the calculated compositions, six simulated AgInCdSn alloys were prepared and examined. With the increase of Cd and Sn, the simulated AgInCdSn alloys transform from a single fcc phase into the mixed fcc and hcp phases, and finally into the single hcp phase. The atomic volume of the hcp phase is obviously larger than the fcc phase. The fcc-hcp transformation results in considerable volume swelling of the AgInCd absorber. Moreover, the lattice parameters of the fcc and hcp phases gradually increase with Cd and Sn contents, which also can induce small volume swelling.

• Journal of Energy Engineering
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• v.7 no.1
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• pp.73-80
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• 1998

Zinc titanate sorbents were prepared and regeneration of used sorbents for high temperature desulfurization of fuel gases was studied. Zn/Ti molar ratio of prepared sorbents was 1.5 and quartz fixed-bed reactors with 1 cm and 3 cm diameters were used. Regeneration of zinc titanate sorbents at high temperature is exothermic reaction; that brings about deterioration of sorbents. Therefore, we experimented regeneration reaction of zinc titanate sorbents for the purpose of obtaining the most suitable regeneration conditions by changing experimental parameters such as reaction temperature, oxygen concentration, flow rate and steam content. $H_2S$ and $SO_2$ breakthrough curves were obtained during desulfurization-regeneration. Also, properties of the sorbents before and after regeneration were analyzed using SEM, XRD, Hg-porosimetry and BET method. From such results, we obtained the most suitable regeneration conditions including regeneration temperature of 650$^{\circ}$C, $O_2$ content of 5% and steam content of 10% in the gas stream.

• Ecology and Resilient Infrastructure
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• v.5 no.3
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• pp.174-179
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• 2018

$TiO_2$-embedded expanded polystyrene (TiEPS) balls with powdered $TiO_2$ particles embedded on the surface of EPS were developed, and the growth inhibition of Chlorella ellipsoidea, a green algae, was evaluated. The experiment was conducted using four reactors with various conditions of (A) natural sunlight, (B) natural sunlight + TiEPS balls, (C) dark, and (D) dark + TiEPS balls on the roof of the building during five days. Based on the analysis of cell number, cell morphology, concentrations of chlorophyll-a and phaeopigments, both surface reactions in heterogeneous photocatalysis and light shielding could inhibit the growth of C. ellipsoidea. The highly reactive hydroxyl radicals ($OH{\cdot}$) from TiEPS balls degraded the lipid cell membrane through the peroxidation reaction with the light shielding, eventually resulting in cell inactivation. Although dominant inhibitory effects on the growth of C. ellipsoidea were ambiguous, TiEPS balls were feasible to prevent and inhibit the excessive growth of algae in eutrophic water body.

• Journal of the korean society of oceanography
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• v.33 no.1-2
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• pp.35-40
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• 1998

In February and April 1997, three temporary interruptions of electric power production at the Uljin Nuclear Power Plant in the east coast of Korea were caused by the malfunction of the cool-ing-water supply unit. The clogging of the drum screens inside the unit caused by the surface swarm of the euphausiid Euphausia pacifica Hansen might be responsible for the malfunction. These incidents were of particular interest since such interruption of reactors' operation by krill swarms had not previously been reported. Using samples caught by the drum screens inside the cooling water-supply unit, we investigated the population structure of surface swarms. One occasion of nighttime and three occasions of daytime surface swarms were found in February and April 1997, respectively. The foreguts of more than 60% of E. pacifica in nighttime surface swarm were in full condition. This evidence suggests that E. pacifica aggregates to the surface water at night for feeding. In daytime surface swarms consisting of mature E. pacifica, however, foreguts in full condition were only found in less than 10eio of krill examined, suggesting that daytime surface swarms are closely related to breeding activity. During the study period, the growth rate of mature females was more than twice higher than that of mature males. Analyses of the sex-ratio and length-frequency data show a decrease in the portion of males with increasing size. There was a decline in the number of males of 19 mm in length. Energy loss during spermatophore transfer may result in the death of male E. pacifica, as found in male E. superba.