[en] LOTIS is an automated wide field-of-view telescope system capable of responding to GRB events as early as 10s after a trigger from the GCN which rapidly distributes coordinates from the Beppo/SAX, BATSE and RXTE instruments. Measurements of optical activity at these early times will provide important clues to the GRB production mechanism. In over two year's of operation, LOTIS has responded to 40 GCN triggers including GRB971217 with l10s and GRB980703 within 5 hours. We report results from these events and constraints on simultaneous optical signals during these GRB's

[en] Conceptual design of DUPIC irradiation pellets with double cladding was carried out. And the preliminary study of the temperature effect on the design and manufacturing parameters of DUPIC pellets through HEATING and GENGTC was performed. The analysed results of the newly designed DUPIC pellets to be irradiated in HANARO, were, 1) thermal conductivity of fuel, linear power of fuel and axial gap affected greatly the temperature of fuel, 2) thickness of sheath, gamma heating rate and thermal transfer coefficient affected little the temperature of fuel. 3) the centerline temperature calculated by HEATING was evaluated higher than that by GENGTC such presented to be desirable for using GENGTC in the view point of safety GENGTC, 4) by transient thermal analysis, after 160 seconds, the temperature of fuel reaches its equivalent temperature. (author). 3 tabs., 16 figs

[en] A series of experiments are performed to verify the manufacturability of DUPIC fuel and its performance by use of HANARO test reactor. Major works performed during this research period are : analysis of manufacturing process of DUPIC fuel, fabrication technology development such as development of disassembly and decladding method of spent PWR fuel, study on the OREOX process using simulated high burnup fuel, weldability of end cap weld, and development of fabrication equipment including the conceptual and detailed design of DUPIC equipment mainly for the powder preparation, pelletization and fuel element fabrication. A study on the material properties of DUPIC fuel and performance analysis method using irradiation of test fuel was also performed. (author). 91 refs., 274 tabs., 254 figs

[en] We are studying the feasibility of utilizing Kα x-ray sources in the range of 20 to 100 keV as a backlighters for imaging various stages of implosions and high areal density planar samples driven by the NIF laser facility. The hard x-ray Kα sources are created by relativistic electron plasma interactions in the target material after a radiation by short pulse high intensity lasers. In order to understand Kα source characteristics such as production efficiency and brightness as a function of laser parameters, we have performed experiments using the 10 J, 100 fs JanUSP laser. We utilized single-photon counting spectroscopy and x-ray imaging diagnostics to characterize the Kα source. We find that the Kα conversion efficiency from the laser energy is ∼ 3 x 10^-4

[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] The goal of this project is the safe and successful decommissioning of the inside reactor hall of the Korean Research Reactor No 2 (KRR-2) and convert to temporary storage facility for the radioactive waste produced during decommissioning. It's necessary to manage the overall decommissioning and decontamination project for the man-power, the progress of the work, etc. needed to works and to develop the suitable technology and equipment in order to perform the decommissioning works for the purpose of securing the safety and minimizing the radiation exposure for works. Also, the large amount of the liquid and solid wastes were generated from the dismantling works. The radioactivity of the waste was not high but the amount was large and the properties was very diverse, and therefore unique management technologies were required for the decommissioning waste. The operation experience of the uranium conversion plant as a nuclear cycle facility was contributed to the localization of nuclear fuels for HWR. It was shut down in 1993. And, in 2001 the decontamination and dismantlement program for the conversion plant has been launched to achieving radiation safety and environment restoration. Conversion plant environment restoration project will be contributed to developing the decontamination and dismantlement technologies related to other domestic nuclear facilities and to set new criteria in the D and D areas

[en] In 2004, the dismantling of the reactor pool, including the equipment in the pool, was started. The radioactivity of the equipment, caused by activation was much higher than the expected activity and 7 remotely operated devices were developed to dismantle them. Project management including man power control, quality control was carried out and radiation protection programs and environmental surveillance was also conducted. The waste from dismantling sites was classified according to the physco-chemical properties and radioactivity level, each was further treated or stored according to the relevant procedures. The issues from the safety evaluation by KINS was analyzed and answered. Some preparative works, such as a supply of new electric source and installation of a new ventilation system, were conducted. A process for the treatment of the sludge waste in lagoon was developed and the basic requirements for the process were established

[en] Irradiation characteristics examination technology development of irradiated nuclear materials and high burn-up fuels will support not only technical bases of high performance nuclaer fuels and structural materials, which are developed by domestic technologies in the future, but also research of new conceptual design research reactor and fuels. Main objective of this project is to maintain chemical hot cell facility and essential instruments for chemical examination and analysis KAERI has and is to support other nuclear R and D groups and nuclear industries with producing examination data for the specimens requested using these facility and instruments. The objective of the project is to collect, treat and store those radioactive wastes safely for the preservation of the environment from being contaminated by the radioactive wastes generated at KAERI. Therefore, radwastes are disposed of in a disposal site as solidified waste forms for its complete isolation from the human environment. The physicochemical properties of waste forms and the radionuclide concentration in waste forms should be evaluated for the radiological and structural safety of a disposal site, radionuclide type and solidification matrix, and it is difficult to carry out tests(for example, compressive strength, leaching rate, etc)with a full-scale waste forms. The waste classification and acceptance criteria is the result of technology development for characterization of waste and solidified waste forms. This treatment is carry out to low-cost and low-absorbed dose

[en] The computer program SAFE has been used to size and analyze the performance of a steam generator which has two types of heat transfer regions in Korean Standard Nuclear Power Plants (KSNP) and Korean Next Generation Reactor (KNGR) design. The SAFE code calculates the analytical boiling heat transfer area using the modified form of the saturated nucleate pool boiling correlation suggested by Rohsenow. The predicted heat transfer area in the boiling region is multiplied by a constant to obtain a final analytical heat transfer area. The inclusion of the multiplier in the analytical calculation has some disadvantage of loss of complete correlation by the governing heat transfer equation. Several comparative analyses have been performed quantitatively to evaluate the possibility of removing the multiplier in the analytical calculation in the SAFE code. The evaluation shows that the boiling correlation and multiplier used in predicting the boiling region heat transfer area can be replaced with other correlations predicting nearly the same heat transfer area. The removal of multiplier included in the analytical calculation will facilitate a direct use of a set of concerned analytical sizing values that can be exactly correlated by the governing heat transfer equation. In addition this will provide more reasonable basis for the steam generator thermal sizing calculation and enhance the code usability without loss of any validity of the current sizing procedure. (author)

[en] A new ion exchanger with porous silica as a supporting material and diphosphonic acid as a functional chelating group has been developed at ANL for the effective removal of transition metals and actinide ions from very acidic radioactive liquid wastes. The applicability of this resin for the treatment of low- and/or intermediate-level aqueous waste from nuclear power plants (NPP) has not been reported in scientific literature, but is under study now in Korea. The major radioisotopes in NPP radioactive liquid waste are Cs and Co in neutral pH ranges. This study on the thermal stabilization of metal-loaded waste resin has been carried out in parallel with the sorption experiment. Thermal treatment of metal (Co, Cs or U) loaded resin was accomplished to see the possibility of enhancing the safety and stability of the final product during transportation and disposal. In this paper, characteristics of the metal-loaded resins before and after heat treatment at three different thermal conditions were investigated and compared with each other to see the effectiveness of the thermal treatment method