[en] For developing the sodium_cooled fast reactor (SFR), an ex_core neutron flux monitoring system (ENFMS) should be designed to maintain the reactor safely and to regulate the reactor power. This report summarizes the results of study on types of detector for monitoring of neutron flux in SFR, measuring ranges, location of detectors within the reactor considering the environmental conditions, digital signal processing technologies for the neutron flux signals, design concept for configuration of the system, and a few of technical issues to be considered along with the reactor core design and safety analysis field design advance in the future. There are two design features in the ENFMS of SFR. One is that detectors are located within the reactor vessel, due to that the diameter of SFR reactor is larger than that of the conventional pressurized water reactor (PWR). The second is that a wide_range neutron flux detectors are applied, which are able to measure not only the wide_range output from source range to power_range being used for the reactor protection and monitoring but also the narrow_range output being used for the reactor control. But unfortunately, there is no commercialized detector in which it satisfies the system design requirements. Therefore the detector has to be developed through the joint research with the detector supplier. Accordingly, this report will be able to be used for the design and fabrication of wide_range neutron detectors to be performed in the near future, and will be referred to the design of ENFMS at preliminary and detail design stages in the next future

[en] A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident

[en] The rise of temperature and pressure, the release of aerosol in the buildings as a result of sodium fire must be considered for the safety measures of LMR. Therefore for the safety of the LMR, it is necessary to understand the characteristics of sodium fire, resulting from the various type of leakage. ASSCOPS(Analysis of Simultaneous Sodium Combustion in Pool and Spray) is the computer code for the analysis of the thermal consequence of sodium leak and fire in LMR that has been developed by Japan Nuclear Cycle Development Institute(JNC) in Japan. In this study, a preliminary analysis of sodium leak and fire accidents in S/G building of KALIMER is made by using ASSCOPS code. Various phenomena of interest are spray and pool burning, peak pressure, temperature change, local structure temperature, aerosol behavior, drain system into smothering tank, ventilation characteristics at each cell with the safety venting system and nitrogen injection system. In this calculation, the dimension of the S/G building was chosen in accordance with the selected options of LMR name KALIMER(Korea). As a result of this study, it was shown that subsequent effect of sodium fire depended upon whether the sodium continued to leak from the pipe or not, whether the ventilation system was running, whether the inert gas injection system was provided, whether the sodium on floor was drained into the smothering tank or not, whether the building was sealed or not, etc. Specially the excessive rise of pressure into each cell was prevented by installing the pressure release plates on wall of the building

[en] The TSEE and TL from Al₂O₃ single crystals, subjected to X-ray and UV irradiation in a vacuum, were studied. Four TSEE and PL peaks were observed after X-ray irradiation. For UV excitation, two peaks were observed in a TSEE glow curve and only one prominant peak for TL. TSEE and TL glow curves for Al₂O₃ showed much similarity regarding the position of peaks and temperature shifts, regardless of the type of excitation. Thus, there is a close relation between TSEE and TL mechanism concerning Al₂O₃ single crystals. (author)

[en] As a treatment method of the waste sodium which was produced from a sodium facility, an investigation for a reaction procedure of the waste sodium with the sodium hydroxide has been developed. The sodium was injected into a reaction vessel filled with a caustic soda through an atomizing nozzle to maintain the reaction uniformly. There were complex reacting phenomena in the system to observe with a naked eye. Therefore, a water mock-up was carried out for a practical use the data got in the waste sodium treatment test. The major experimental parameters are the flowrate of water through an atomizing nozzle and the recirculation rate. In addition, the positions and flow directions of the nozzles are important parameters, also. From this experiment, 300 sets of data were obtained by analyses of the phenomena of the photographic records, and the optimum flowing conditions

[en] The successful protection of a water/steam into a sodium leak in the LMR SG at an early phase of a leak origin depends on the fast response and sensitivity of a leak detection system. The control time for the protection of the LMR SG is several seconds. Subject of this study is to introduce the detection performance of the acoustic leak detection system discriminated by a back-propagation neural network according to a preprocessing of the FFT power spectrum analysis and the Octave band analysis, and to introduce the status of the development of the acoustic leak detection at KAERI. It was used for the acoustic signals from the injected Argon gas into water experiments at KAERI, the acoustic signals injected from the water into the sodium obtained in IPPE, and the background noise of the PFR superheater

[en] In a sodium cooled fast reactor there exists a free surface in the upper plenum of the reactor vessel where the sodium coolant contacts with the cover gas. Fluctuation of this free surface causes two important phenomena. One is to secure the structural integrity of the reactor vessel due to a thermal striping. Another is the gas entrainment at the free surface. An experimental study has been performed to measure the fluctuation phenomena in a vessel with an internal structure (UIS). The effects of the vessel geometries on a free surface fluctuation are studied to develop a correlation for a free surface fluctuation

[en] In a fast reactor there exists a free surface in the upper plenum of the reactor vessel where the sodium coolant contacts with the cover gas. Fluctuation of this free surface causes two important phenomena, which are a thermal striping on a vessel wall and a gas entrainment at a free surface. An experimental study has been performed to investigate the frequency of a free surface fluctuation in a vessel. There are two dominant frequency regions which are generated by a standing wave and a jet. In this paper, the effect of the geometry on the frequency generated by a jet is described by a dimensionless number based on the central velocity of a jet

[en] It is important to safely treat the waste sodium which is produced from sodium cooled fast reactors and sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from experiments which have been carried out since 1990. Furthermore, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed by France for the treatment of waste produced from sodium facilities and reactors. Through the NOAH process, a small amount of sodium waste is continuously injected into the upper area which is formed on a free surface of the aqueous and slowly reacts with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slower or suddenly increases. An improvement process was proposed by KAERI. The final product of the process is salt through a neutralization of the sodium hydroxide with the nitric acid. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on the lower level than that of the aqueous. The nitrogen gas was injected simultaneously into the reaction vessel through the annular channel of the atomizing nozzle to maintain a consistent reaction. There are complex reacting phenomena in the system to observe with the naked eye. A water mock-up has been carried out already for the practical use of the data in the waste sodium treatment test. Based on the data, an experimental facility was designed and installed for the sodium waste process to find out the effect of the sodium injection rate. But the neutralization process which is not an important parameter in the experiment was eliminated to simplify the process in the experimental facility

[en] The Sodium-cooled Fast Reactor (SFR) is a very promising candidate for the development of fast neutron reactors. Sodium is a very attractive coolant due to its high heat transfer properties, on the other hand, if a water/steam leaks into the sodium, a severe problem can occur on account of a violent sodium-water reaction (SWR). To resolve the difficulties in deploying a SRF for a power generation, an elimination of the IHTS by using a new steam generator (SG) was suggested by Kim et al. The concept of a double tube bundle steam generator (DTBSG) is to eliminate a SWR possibility in a sodium-cooled fast reactor, fundamentally, and a thermal-hydraulic performance analysis code, Integrated Steam Generator Analyzer (ISGA) was developed for three candidate types of DTBSGs that are the integrated single-region, integrated double-region and radially separated bundle types. Heat transfer tube bundle configuration of the DTBSG: (a) integrated single-region; (b) integrated double region; (c) radially separated. In this paper, a heat transfer experiment in an integrated double-region type DTBSG, one of the three tube bundle types was addressed