• Nuclear Engineering and Technology
• /
• v.25 no.2
• /
• pp.259-264
• /
• 1993

Ion chromatographic method has been applied for burnup measurement of irradiated nuclear fuel by dynamic system using 1-octanesulfonate as a cation exchanger and $\alpha$-hydroxyisobutyric acid as an eluant. A number of elution techniques were evaluated for the optimum separation of plutonium, uranium and neodymium. These elements were individually separated and collected by gradient elution between 0.05 M and 0.40 M of $\alpha$-hydroxyisobutyric acid in a single column, and finally determined by isotope dilution mass spectrometry. The burnup data from this method were compared with those from conventional anion exchange method. The results showed a good agreement within 3.5 % of difference between two methods.

• Journal of Radiation Protection and Research
• /
• v.26 no.3
• /
• pp.299-303
• /
• 2001

Using ultrasound techniques, the Korea Atomic Energy Research Institute has measured chest wall thicknesses of a group of male workers at the Korea Atomic Energy Research Institute. A site-specific biometric equation has been developed for these workers. Chest wall thickness is an important modifier on lung counting efficiency. These data have been put into the perspective of the ICRP recommended dose limits for occupationally exposed workers: 100 mSv in a 5-year period with a maximum of 50 mSv in anyone year. For measured chest wall thicknesses of 1.9 cm to 4.1 cm and a 30 min counting time, the achievable MDAs for natural uranium in the KAERI lung counter vary from 5.75 mg to 11.28 mg. These values are close to, or even exceed, the predicted amounts of natural uranium that will remain in the lung (absorption type M and S) after an intake equal to the Annual Limit on Intake corresponding to a committed dose of 20 mSv. This paper shows that the KAERI lung counter probably cannot detect an intake of Type S natural uranium in a worker with a chest wall thickness equal to the average value (2.7 cm) under routine counting conditions.

• Nuclear Engineering and Technology
• /
• v.33 no.1
• /
• pp.73-82
• /
• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.

• Nuclear Engineering and Technology
• /
• v.43 no.3
• /
• pp.309-316
• /
• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

• Nuclear Engineering and Technology
• /
• v.35 no.3
• /
• pp.206-213
• /
• 2003

This study focuses on an understanding of the environmental effect on delayed hydride cracking velocity (DHCV) of CANDU Zr-2.5Nb tubes. To simulate DHC susceptibility of the Zr-2.5Nb tubes in reactor operating conditions, DHC tests were successfully carried out in pressurized water at 180 and $250^{\circ}C$ using a self-designed autoclave for the first time. Using 17 mm compact tension specimens electorlytically charged to 34 and 60 ppm H, 3 to 7 DHCV data were determined in water at both temperatures and compared to those determined in air that were already confirmed to be valid through a round robin test on DHCV of Zr-2.5Nb tubes sponsored by a IAEA coordinated research program. The pressurized water environment has little effect on DHCV of Zr-2.5Nb tube in water at both temperatures even though DHCV is slightly lower in water than that in air. The lower DHCV of the Zr-2.5Nb tube during short-term tests is discussed in viewpoint of the cooling rate from the peak temperature to the test temperature.

• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.242-249
• /
• 2002

The relationship between the tritium release rate from the nuclear power plant and tritium concentration in the environment around the Kori site was modeled. The tritium concentration in the atmosphere was calculated by multiplying the release rates and $\chi$/Q values, and the d3V deposition rate at each sector according to the direction and the distance was obtained using a dry deposition velocity. The area around Kori site was divided into 6 zones according to the deposition rate. The six zones were divided into 14 compartments for the numerical simulation. Transfer coefficients between the compartments were derived using site characterization data. Source terms were calculated from the dry deposition rates. Tritium concentration in surface soil water and groundwater was calculated based upon a compartment model. The semi-analytical solution of the compartment model was obtained with a computer program, AMBER. The results showed that most of tritium deposited onto the land released into the atmosphere and the sea. Also, the estimated concentration in the top soil agreed well to that measured. Using the model, tritium concentration was predicted in the case that the tritium release rates were doubled.

• Journal of Radiation Industry
• /
• v.4 no.1
• /
• pp.47-51
• /
• 2010

Centipedegrass is a warm season turfgrass in the world. Chlorogenic acid (CA) is one of the important compounds present in the leaf of centipedegrass and already known as an antioxidant, CA has become a key resistance against insect pests and bacteria pathogens of agricultural and horticultural plants during seedling stage. Furthermore, CA is accumulated by abiotic stress such as an UV irradiation. In present study, we investigated enhancement of the level of CA upon gamma irradiation in centipedegrass. The high performance liquid chromatography (HPLC) data analysis showed an approximately increasing of the CA levels from among the irradiated samples. However, plants irradiated at 50 Gy showed a constant increase in the CA level (0.0066 to $0.114mg\;ml^{-1}$ and 0.0258 to $0.2211mg\;ml^{-1}$, respectively) from $3^{rd}$ to $15^{th}$ day among one and three month irradiated plants compared to control. The present study, indicates an increase in the CA level upon gamma irradiation, suggests strategy for conferment of strong resistance in seedling stage plants by gamma irradiation as simplicity and cheaply method.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.413-418
• /
• 1996

The main purpose of this paper is to develop the modified LTC code for accurate analysis of the boron concentration behavior of all components in the Nuclear Steam Supply System (NSSS). This is achieved by adapting a multi-cell mad to the existing Long Term Cooling (LTC) code. To verify the modified LTC, the simulated results were compared with the actual test results measured during YGN 4 initial criticality test. It was shown that the simulated results of this modified LTC were in good agreement with the actual test results. Also, the boron concentration behavior analysis were performed using the modified LTC code for both direct and indirect dilution/boration nude using YGN 3,4 design data. This modified LTC code can provide a valuable information in predicting boron concentration behavior during power maneuvering such as startup operation, shutdown operation and load follow operation. It is expected that the modified LTC can be applied to both on-line and off-line mode using Plant Computer System(PCS).

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.425-432
• /
• 1996

Integrated performance test of Chemical and Volume Control System (CVCS) was successfully performed in 1994. However, an extensive effort to correct hardware and software problems in the letdown line was required mainly due to the lack of adequate simulation code to predict the test accurately. Although the LTC computer code was used during the YGN 3'||'&'||'4 NSSS design process, the code can not satisfactorily predict the test due to its insufficient letdown line modeling. This study developed a numerical model to simulate the letdown test by modifying the current LTC code, and then verified the model by comparing with the test data. The comparison shows that the modified LTC computer code can predict the transient behavior of letdown system tests very well. Especially, the model was verified to be able to predict the "Stiction" phenomena which caused instantaneous fluctuations in the letdown backpressure and flowrate. Therefore, it is concluded that the modified LTC computer code with the ability of calculating the "Stiction" phenomena wi11 be very useful for future plant desist and test predictions.predictions.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05b
• /
• pp.129-136
• /
• 1997

The automated ball indentation(ABI) test was utilized to develop a semi-nondestructive method for estimating the fracture toughness( $K_{JC}$ ) in the transition temperature range. The key concept of the method is that the indentation deformation energy to the load at which the mean ball-specimen contact pressure reaches the fracture stress is related to the fracture energy of the material. ABI tests were performed for the reactor pressure vessel(RPV) base and weld metals at the temperatures of-15$0^{\circ}C$~$0^{\circ}C$ and the fracture toughness (estimated $K_{JC}$ ) was calculated from the indentation load-depth data. For all steels the temperature dependence of the estimated fracture toughness was almost the same as that ASTM $K_{JC}$ master curve The reference temperatures( $T_{o}$)of the steels were determined form the estimated $K_{JC}$ versus temperature curves. The reference temperature was well correlated with the index temperature of 41J Charpy impact energy( $T_{41J}$).).).