• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1398-1406
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• 2024

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subcritical states. Moreover, the noise reduction technique has been adopted to smooth the large fluctuation existing in the original signals. In order to verify the engineering performance of the proposed measurement technique, the measured source-range detector count rates during the rod withdrawal process before reaching critical state in a CNP1000 reactor have been employed. It demonstrated that almost all estimated results of control rod worth satisfy the engineering acceptance criteria, except one control rod with the relative difference over 10 %, which indicates the capability of the proposed method in estimating control rod worth.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1135-1144
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• 2024

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MW_th sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MW_th for peak demand for the rest of the day.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Journal of the Korea Organic Resources Recycling Association
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• v.21 no.2
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• pp.63-70
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• 2013

Experiments in a pilot-scale hybrid multi-stage unit system (HMUS) combination of ATAD and EGSB followed by SBR process for pig slurry treatment were conducted to demonstrate the feasibility of using autothermal thermophilic aerobic digestion (ATAD) and expended granular sludge bed (EGSB) followed by sequencing batch reactor (SBR) system. Contaminants in pig slurry with high organic matter, nitrogen (N) and phosphorus (P) content were completely removed in the combined process. The highest removal rate for CODcr among contaminants in the feed pig slurry was attained by about 43.3% in ATAD unit process. Also TS removal rate of 96.5% was attained and the highest in the next coagulation unit process. The highest removal rate of CODcr under operating parameter conditions of OLR(organic loading rate), 3-6Kg $COD/m^3{\cdot}day$ and line velocity, 1.5-4m/h was earned at 3days of HRT. The disinfection of pathogens was effective at 50,000mg/L of TS in ATAD unit process. Biogas production per organic removal was $2.3{\sim}8.5m^3/kgTS{\cdot}d$ (average $5.2m^3/kgTS{\cdot}d$) in EGSB unit process. The average removal rates of CODcr 71.7%, TS 64.1%, TN 45.9%, and TP 50.4% were earned in the intermittent aeration SBR unit process.

• Korean Journal of Food Science and Technology
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• v.35 no.3
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• pp.417-422
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• 2003

The onions are considered to be a favorable functional source of beverage because they contain much sugar and various nutrients, and they are juicy vegetable. Recently, consumers have a new trend to take functional foods with health benefits. To meet this need, this study was the basic research to establish a manufacturing process of functional onion beverage by glucose oxidase. Glucose oxidase catalyzes reaction of glucose oxidation and makes generation of gluconic acid. Kinetics of the reaction was also investigated, and maximum glucose consumption rate $(V_{max})$ of $26.1{\times}10^{-2}\;g/L{\cdot}min$ and $K_m$ of 5.84 g/L were obtained. Optimum conditions were obtained when the glucose oxidase catalyzed reaction was carried out at temperature of $25^{\circ}C$, agitation rate of 450 rpm and aeration rate of 4 vvm in a 2.5 L jar fermentor. Finally, the enzyme reactor was 10-times scaled up and a similar glucose oxidation performance was achieved in the scaled-up reactor.

• Korean Chemical Engineering Research
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• v.52 no.1
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• pp.133-140
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• 2014

The experiment was designed to compare the char combustion kinetics of pulverized Indonesia coals commonly utilized in Korea power plants. The reaction rate of coal char has been formulated using the external and internal effectiveness factors to describe the diffusion effect quantitatively. The Random Pore Model (RPM) was used for applying internal specific surface area as a function of carbon conversion ratio. Reaction rate was obtained from reaction time using the Wire Heating Reactor (WHR) which can heat and measure the char particle temperature at the same time. BET and TGA were used to obtain physical properties such as internal specific surface area and structural parameter. Three kinds of Indonesia Sub-bituminous coals "BARAMULTI, ENERGYMAN, AGM" were used in order to derive the activation energy and pre-exponential factor. The results of this study showed that the effect of internal diffusion than that of external diffusion is the dominant as comparison of kinetics was reflected in external and internal effectiveness factors. For three kinds of coal char, finally, activation energy of intrinsic kinetics indicates 110~118 kJ/mol.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.842-848
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• 2022

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 ℃. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner subchannel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

• Journal of the Earthquake Engineering Society of Korea
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• v.2 no.4
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• pp.73-86
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• 1998

The purpose of this paper is to evaluate the shear stiffness, hysteretic behavior, and ultimate behavior of HDRB(High Damping Rubber Bearing), which will be included in the seismic isolation design guideline as requirements. To do this, two 1/8 scaled HDRB are designed, fabricated, and tested to show the mechanical characteristics. The shear stiffness obtained from the proposed equation of the shear stiffness shows a good agreement with those of the experiments. For analysis of the hysteretic behavior of HDRB using the modified rate model, the parameter equations are obtained from the experiments. Using the obtained parameter equations for the modified rate model, the seismic response analyses are carried out for 1-D system. The results of analysis well follow the hysteretic behavior of HDRB obtained from the experiments. To evaluate the ultimate behavior of HDRB used in this paper, the analyses are carried out using the modified macro model, which can consider the large shear deflection. The critical shear strain(CSS) is defined to express the maximum allowable shear strain and vertical load. From the analyses, the CSS, showing the instability, decreases significantly as increased the vertical loads. The CSS is not appeared for the design vertical load in the used HDRB. In analysis using about 5 times of design vertical load, the HDRB start to show the instability transient and for about 7 times, the CSS is about 350%.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.12
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• pp.1257-1263
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• 2015

A correlation between the transient regime and steady state regime on the creep crack growth (CCG) for Grade 91 steel, which is used as the structural material for the Gen-IV reactor systems, was investigated. A series of CCG tests were performed using 1/2" CT specimens under a constant applied load and at a constant temperature of $600^{\circ}C$. The CCG rates for the transient and steady state regimes were obtained in terms of $C^*$ parameter. The transient CCG rate had a close correlation with the steady-state CCG rate, as the slope of the transient CCG data was very similar to that of the steady state data. The transient rate was slower by 5.6 times as compared to the steady state rate. It can be inferred that the steady state CCG rate, which is required for long-time tests, can be predicted from the transient CCG rate obtained from short-time tests.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.