• Tribology and Lubricants
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• v.30 no.2
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• pp.108-115
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• 2014

The friction behaviors of bulk amorphous thermal spray coating (BAC) and second phase-reinforced composite coatings using a high velocity oxy-fuel spraying process were investigated using a ball-on-disk test rig that slides against a ceramic ball in an atmospheric environment. The surface temperatures were measured using an infrared thermometer installed 50 mm from the contact surface. The crystallinities of the coating layers were determined using X-ray diffraction. The morphologies of the coating layers and worn surfaces were observed using a scanning electron microscope and energy-dispersive spectroscopy. The results show that the friction behavior of the monolithic amorphous coating was sensitive to the testing conditions. Under lower than normal loads, a low and stable friction coefficient of about 0.1 was observed, whereas under a higher relative load, a high and unstable friction coefficient of greater than 0.3 was obtained with an instant temperature increase. For the composite coatings, a sudden increase in friction coefficient did not occur, i.e., the transition region did not exist and during the friction test, a gradual increase occurred only after a significant delay. The BAC morphology observations indicate that viscous plastic flow was generated with low loads, but severe surface damage (i.e., tearing) occurred at high loads. For composite coatings, a relatively smooth surface was observed on the worn surface for all applied loads.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.573-580
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• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• Journal of Energy Engineering
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• v.28 no.3
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• pp.1-9
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• 2019

A major stress determining the remaining life of the tube in feedwater heater of fossil fuel power plant is hoop stress by the internal pressure. However, thermal stress due to temperature difference across the wall thickness also contributed to reduce the remaining life of the tube. Therefore, thermal loading must be considered even though the contribution of internal pressure loading to the stresses of the tube was known to be much higher than that of the thermal loading. In this study, thermal stress of the tubes in the de-superheating zone was estimated, which was generated due to the temperature difference across the tube thickness. Analytic equations were shown for determining the hoop stress and the radial stress of the tube with uniform thinning and for the temperature across the tube thickness. Accuracy and effectiveness of the analytic equations for the stresses were verified by comparing the results obtained by the analytic equations with those obtained from finite element analysis. Using finite element analysis, the stresses for eccentric thinning were also determined. The effect of heat transfer coefficient on thermal stress was investigated using series of finite element analyses with various values of heat transfer coefficient for both inner and outer surface of the tube. It was shown that the effect of heat transfer coefficient at outer surface was larger than that of heat transfer coefficient at inner surface on the thermal stress of the tube. Also, the hoop stress was larger than the radial stress for both cases of uniformly and eccentrically thinned tubes when the thermal loading was only considered without internal pressure loading.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.639-647
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• 2018

In this study, the effects of selecting water/silver nanofluid as both a coolant and a reactivity controller during the first operating cycle of a light water nuclear reactor are investigated. To achieve this, coupled neutronic-thermo-hydraulic analysis is employed to simulate the reactor core. A detailed VVER1000/446 reactor core is modeled in monte carlo code (MCNP), and the model is verified using the porous media approach. Results show that the maximum required level of silver nanoparticles is 1.3 Vol.% at the beginning of the cycle; this value drops to zero at the end of cycle. Due to substitution of water/boric acid with water/Ag nanofluid, reactor operation time at maximum power extends to 357.3 days, and the energy generation increases by about 27.3%. The higher negative coolant temperature coefficient of reactivity in the presence of nanofluid in comparison with the water/boric acid indicates that the reactor is inherently safer. Considering the safety margins in the presence of the nanofluid, minimum departure from nucleate boiling ratio is calculated to be 2.16 (recommendation is 1.75).

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• Journal of the Korean Ceramic Society
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• v.45 no.12
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• pp.777-781
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• 2008

Experiments were conducted using SUS430 and Crofer 22 APU steels coated by LSM using plasma spray and slurry spray methods, respectively. High-temperature conductivity and oxidation resistance were investigated. For comparison, SUS430 and Crofer 22 APU without LSM coating were also investigated and coefficient of thermal expansion (CTE) was measured. The results show that the materials without LSM coating exhibit almost the same CTE as YSZ electrolyte in a range of temperatures of $550{\sim}850^{\circ}C$. When coated with LSM, the oxidation rate of the steels decreases by $30{\sim}40%$ using slurry spray and by $10{\sim}30%$ using plasma spray whereas the steels using plasma spray have a better high-temperature conductivity than the steels using slurry spray. It is thus concluded that the LSM coating has a limited effect on increasing high-temperature conductivity while it can effectively reduce the oxidation of the steels.

• Journal of the Korean Society of Propulsion Engineers
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• v.10 no.3
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• pp.90-98
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• 2006

The Heated Tube Facility(HIF) was fabricated to identify the forced convective heat transfer and the cooling characteristic for the hydrocarbon fuel(Jet A-1), which is used for the coolant of the regenerative cooling system. The forced convective heat transfer coefficient was calculated from the measured coolant and tube surface temperature. In case of using the Jet A-1, the maximum heat flux which the coolant can absorb was identified by determining the critical wall temperature generating the burnout on the fixed flow condition. The inlet bulk-temperature of the coolant has a direct influence on the forced convective heat transfer characteristic.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.1
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• pp.9-16
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• 2024

The bentonite buffer material is a crucial component in an engineered barrier system used for the disposal of high-level radioactive waste. Because a large amount of heat from the disposal canister is released into the bentonite buffer material, the thermal conductivity of the bentonite buffer is a crucial parameter that determines the design temperature. At the Korea Atomic Energy Research Institute (KAERI), a new standard bentonite (Bentonil-WRK) has been used since 2022 because Gyeongju (KJ) bentonite is no longer produced. However, the currently available data are insufficient, making it essential to investigate both the basic and complex properties of Bentonil-WRK. Thus, this study evaluated its geotechnical and thermal properties and developed a thermal conductivity empirical model that considers its dry density, water content, and temperature variations from room temperature to 90℃. The coefficient of determination (R²) for the model was found to be 0.986. The thermal conductivity values of Bentonil-WRK were 1-10% lower than those of KJ bentonite and 10-40% higher than those of MX-80 bentonites, which were attributable to mineral-composition differences. The thermal conductivity of Bentonil-WRK ranged between 0.504 and 1.149 W·(m^-1·K^-1), while the specific heat capacity varied from 0.826 to 1.138 (kJ·(kg^-1·K^-1)).

• Journal of the Korean Society of Propulsion Engineers
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• v.23 no.5
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• pp.101-106
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• 2019

Carbon deposition on the turbine nozzle throat of a LOx/kerosene gas generator cycle(open cycle) engine causes performance reduction of the engine. Estimation methods for a turbine nozzle throat area are proposed. The discharge coefficient of the turbine nozzle was estimated with the turbine gas properties such as gas constant, specific heat ratio, and temperatures. The pressure ratio and temperature ratio of the turbine nozzle throat, was utilized to estimate the discharge coefficient also. Estimated discharge coefficient of turbine nozzle throat of KSLV-II 1st stage engine shows the carbon deposition effects on the turbine nozzle throat of a LOx/kerosene open cycle engine.