[en] The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis

[en] The objective of the irradiation test of DUPIC fuel at HANARO is to obtain the data of in-core behavior and evaluate the nuclear, thermal and mechanical performance of DUPIC fuel. The irradiation of DUPIC fuel will start at April 25, 2000 for about 2 months, and the burnup of 2,000 MWD/MTU will be attained for this period. The pre-irradiation examinations for DUPIC fuel, such as visual inspection, dimension measurement, He leak test and microstructure observation, was carried out. The post-irradiation examination items for the irradiated DUPIC fuel are planned to be the NDA test, visual inspection and dimension measurement, as well as the analyses for the fission gas release, the microstructure of pellets and the distribution and shape of imbedded nuclides. The DUPIC mini-elements were fabricated in the DFDF (IMEF M6 cell) using the G23-G2 rod. For the HANARO core calculation, the initial composition of DUPIC fuel was estimated using ORIGEN-2 code based on the burnup history of the G23-G2 rod. The design features of DUPIC pellets, the mini-element and the irradiation capsule, were supplemented considering the characteristics of DUPIC fuel and the results from the irradiation test of the simulated DUPIC fuel performed in 1999. The nuclear, thermohydraulic and mechanical characteristics of DUPIC fuel under the normal operation condition were evaluated for the safety analysis on the HANARO. Using these results, potential accidents initiated by DUPIC fuel were estimated, and Safety analyses on the locked rotor and RIA accidents were carried out in order to assess the integrity of DUPIC fuel under the accident condition initiated by the HANARO. Based on the results of these safety analyses, the supplemental countermeasures for securing the sufficient thermal margins were set up, as well. At the last, similar overseas and domestic cases were introduced

[en] A safety analysis on the irradiation of simulated DUPIC fuel at HANARO was performed. Thermal, stress, reactivity, thermohydraulic and mechanical analysis was carried out in normal operation condition. RIA possibility and reliability of fuel was assessed during reactor induced accidents. Temperature change and MDNBR was analysed during the critical Hanaro accidents. All kinds of accidents of HANARO was considered to evaluate DNB. Based on the results, the irradiation of simulated DUPIC fuel at HANARO was concluded to be safe. (author). 3 refs., 3 tabs., 6 figs

[en] This report is contained research status of foam separation technique and thus theoretical backgrounds, the kinds of surfactants using for the foam separation technique, their characteristics, general structure and role, surfactant absorption mechanism at liquid/vapor/solid interfaces, effectiveness and efficiency were interpreted with well known models. Ion flotation and precipitate flotation which are applicable for the treatment of very low radioactive liquid wastes were analyzed on the effect of pH, foreign ions, initial concentration of metal ion through the recent presented papers. As the result of technical analysis of foam separation technique, foam separation technique seems to be applicable for the treatment of very low radioactive liquid wastes such as laundry and shower waste. (author). 42 refs., 2 tabs., 37 figs

[en] Martensitic Cr steels were selected for primary candidate alloys for first wall of fusion reactor bacause martensitic steels exhibited excellent high temperature strength and creep behavior in addition to low swelling rate. Irradiation properties of this alloy were of great concern and many studies are now in progress about it. Irradiation caused hardening below about 400 deg C and softening above 450 deg C. But ductility was always decreased at all temperature with dose. Irradiation did not change the microstructure. Dislocations were aligned within certain planes. But cell structures were more prevail at higher temperatures. Martensitic steels also showed He bubbles and He bubbles located at boundaries. They did not caused embrittlement. (Author)

[en] Knowing the true thickness of thin foil is very important in quantitative analysis in TEM study. Especially, in calculating the distribution of particle, bubble or pore size of irradiated material, some amount of error will be included if the exact thickness is not known. Normally the thickness of thin foil is assumed to be 1 micron in TEM work. It causes serious error in some analysis. A simple and accurate method of measuring the thickness of thin foil is recommended in this paper. It uses the relationship between extinction distance, X-ray counts and thickness. During the process extinction distance of activated plane can be also calculated. (Author)

No abstract available

[en] Microstructures of rapidly quenched Al-Co alloys made by a melt spinner were investigated and in-situ decomposition behavior of the solidified microstructures were also studied. Two regions of microstructures were observed in the as-quenched strips; one was a chilled structure formed on the side facing the copper substrate and the other was a free surface structure on the opposite side. The chill structure consisted of micrograins and fine precipitates, i.e. B phase, which are metastable and have (001)_Al//[001]_B and (130)_Al//)[10]_B orientation relationship with Al matrix. The free surface structure was composed of large and stable Al₉Co₂ precipitates surrounded by microdendritic, eutectic and cell structures. Upon annealing it was found that the finer the micro-structure, the faster the decomposition accompanied by nucleation of the slowly growing precipitates. The microhardness of the as-quenched structure increased with increasing Co contents and with addition of the third elements, and the value decreased monotonically with aging time and temperature. Among the various third elements tried, Mn showed the most effective results in hardness increase. (Author)

[en] Type 316L stainless steel has been used for first wall material of NET (Next European TOKAMAK). The 316L stainless steel was implanted with helium and hydrogen to investigate the irradiation behavior in the temperature range 120-420 deg C. The strength of implanted material increased at 120, 220 deg C while elongation decreased. At 420 deg C, abrupt decrease in strength and elongation occured due to helium bubbles. Slip bands were well developed during tensile test like channel deformation. Dislocations were along the (111) planes and cell structure was also generated at higher temperature. With 500 appm hydrogen implantation, microstructure did not change much but contained small amount of dislocations and stacking faults. (Author)

[en] Research on the etching reaction of UO₂ in CF₄/O₂ r. f. plasma is carried out at temperatures of up to 370 .deg. C under the total pressure of 0.3 Torr. The reaction rates are investigated as functions of CF₄/O₂ ratio, plasma power, and substrate temperature. It is found that the highest etching rate is obtained at 20%O₂ mole fraction regardless of r. f. power and substrate temperature. The highest etching reaction rate at 370 .deg. C under 150W exceeds 1000monolayers/min., which is equivalent to 0.4μm/min. The mass spectrometry analysis result reveals that the major reaction product is uranium hexa-fluoride UF₆. Based on the experimental findings, dominant o verall reaction of uranium dioxide in CF₄/O₂ plasma is determined: 8UO₂ + 12CF₄ + 3O₂ = 8UF₆ + 12CO_2-x where CO_2-x represents the undetermined mix of CO₂ and CO