• Resources Recycling
• /
• v.20 no.6
• /
• pp.56-62
• /
• 2011

Zirconium is used in nuclear reactors as a structural material due to its excellent corrosion resistance and to low neutron crosssection. Variation in the distribution and solubility of Zr(IV) with solution pH was obtained. Distribution of Zr(IV) containing species in HCl and $HNO_3$ solution was analyzed by considering the complex formation of Zr(IV) species with the anion of the inorganic acid. Bromley interaction parameter between $ZrO^{2+}$ and nitrate ion was estimated by using the reported data on the solvent extraction of Zr(IV) by Cyanex272 from $HNO_3$ solution. This Bromley parameter can be utilized in calculating extraction isotherm of Zr(IV) and in predicting the separation factor between Zr(IV) and Hf(IV).

• Journal of Advanced Marine Engineering and Technology
• /
• v.11 no.1
• /
• pp.72-79
• /
• 1987

When the wall temperature is very high, a stable vapor film covers the heat transfer surface. The vapor film creates a strong thermal resistance when heat is transferred to the liquid though it. This phenomenon, called "film boiling" is very important in the heat treatment of metals, the design of cryogenic heat exchangers, and the emergency cooling of nuclear reactors. In the practical engineering problems of the transient cooling process of a high temperature wall, the wall temperature history, the variation of the heat transfer coefficients, and the wall superheat at the rewetting points, are the main areas of concern. These three areas are influenced in a complex fashion such factors as the initial wall temperature, the physical properties of both the wall and the coolant, the fluid temperature, and the flow state. Therefore many kinds of specialized experiments are necessary in the creation of precise thermal design. The object of this study is to investigate the heat transfer characteristics in the transient cooling process of a high temperature wall. The slow transient cooling experiment was carried out with a copper block of high thermal capacity. The block was 240 mm high and 79 mm O.D.. The coolant flowed throuogh the center of a 10 mm diameter channel in the copper block. In the copper block, three sheathed thermocouples were placed in a line perpendicular to the flow. These thermocouples were used to take measurements of the temperature histories of the copper block.

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

• Nuclear Engineering and Technology
• /
• v.43 no.3
• /
• pp.257-270
• /
• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.8
• /
• pp.1105-1111
• /
• 2010

The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.