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; Hirakawa, Yasushi; ;
JNC TN9410 2001-023, 84 Pages, 2001/08
JNC-TN9410-2001-023.pdf:6.77MB
NaK (an alloy of sodium and potassium) Flow and Heat Transfer Test Loop was dismantled from January to Februaly 2001. This report shows experience and results obtained in the dismantling. NaK has a low melting point, -12.6
C and has high chemical activity even comparing to sodium because of containing potassium of 78wt%. In addition, the dismantling of the NaK loop was the first experience in OEC. Therefore before dismantling NaK loop, evaluation of dismantling method, trainings of NaK handling and experiments of NaK reaction were carried out by all concerned to ensure the safety. As a result, the test loop was safely dismantled and NaK handling method was obtained. Followings are major results of this experience: (1)Fpllowings were confirmed by experiments. (a)When NaK was exposed in atmosphere, NaK super oxide was produced by exothermal reaction with oxygen. However, the effect of thermal diffusion to NaK and a metal tray scarcely leads the ignition. (b)Exposing NaK alternately in low and high oxygen atmosphere increases the possibility of ignition. (c)As nitrogen gas may react exothermally with burning NaK and formulate KNO
, nitrogen gas does not have instant extinguish capability. (2)Cutting NaK pipe in the vinyl-bag filled with argon gas was effective to avoid NaK ignition. (3)Because there were no solidified crevice in components like sodium at the room temperature, NaK equipments were more easily dismantled than sodium equipments.
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JNC TN9410 99-013, 72 Pages, 1999/04
Fast breeder reactors use metallic sodium as a coolant, therefore, sodium removal is necessary in the dismantling. An intermediate heat exchanger (IHX) exchanges heat between primary and secondary sodium. The dismantling and sodium removal of IHX is difficult because of the future of IHX such as residing of sodium on both primary and secondary surface, existing of the cover gas region, large amount of bulk residual sodium. In the dismantling and sodium removal of the 50MWt IHX, the effective and safe procedure of dismantling and sodium removal was carefullny examined to prevent of sodium ignition and large sodium water reaction and to store safely during the dismantling. Sodium carbonation had been carried out by introducing carbon dioxide in the IHX at the 50MW Steam Generator Test Facility (50MWSGTF) to prevent sodium ignition. After separation to inner shroud and outer shell, each part was transported to the sodium processing facility where each part was dismantled and sodium was removed by steam cleaning device in the atmosphere. Followings are the major results of this experience. (1)Expelimentally obtained sodium shearing force of 0.3 MPa was confirmed by the separation of inner shroud and outer shell (2)No sodium ignition was observed during the dismantling. Therefore introducing carbon dioxide to IHX could be effective. (3)When quantity and condition of residual sodium cannot be confirmed visually, it is important to control steam volume in nitrogen gas and control reaction between sodium and steam. And also it is important to avoid quick approach for the sodium removal observation due to time delay of sodium and steam reaction after steam injection. (4)Sodium removal weight was about 60kg. The residual sodium of the sodium dipped area was about 0.23mg/cm
and it was about l3.7mg/cm in the cover gas region. (5)The tube bundle rotation was effective for the improvement in safety and efficiency of the cleaning and decommissioning. A newly ...
; Hirabayashi, Masaru; ; Oki, Shigeo; ;
JNC TN9400 99-030, 123 Pages, 1999/04
In order to develop a numerical code simulating sodium fires initiated frame dispersion of droplets, measured data of droplet diameter as well as its distribution are needed. In the present experiment the distribution of droplet diameter was measured using water, oil and sodium. The tests elucidated the influential factors with respect to the droplet diameter. In addition, we sought to develop a similarity law between water and sodium. The droplet size distribution of sodium using the large diameter droplet (Elnozzle) was predicted. The results are as follows. (1)Verification of existing method to determine droplet size distribution. Using a phase Doppler system the droplet size distribution to water spray from a binary fluid nozzle was measured. We found that there was not a large difference between the measured distribution and Nukiyama-Tanasawa distribution function. (2)Characterization of inferential parameters with respect to droplet size distribution. Here we note that the droplet size distribution using a nozzle with binary fluid was different from the one using a compression nozzle was used. It was clarified that the viscosity and surface tensjon are the primary factors which influence the volumetric average diameter. The correlation equation between droplet diameter containing viscosity and surface tension terms, was derived. (3)Evaluation on the droplet size distribution to sodium spray from the El nozzle. A relationship between the volumetric average diameter and the pressure was studied and the volumetric average diameter was formulated as a function of the physical properties (viscosity, surface tension) in the case of the E1 nozzle. Finally, the droplet size distribution to sodium spray from the E1 nozzle was estimated using the developed correlation equation.
Gunji, Minoru; Yamamoto, Shimpei; Onojima, Takamitsu
PNC TN9450 98-009, 150 Pages, 1998/06
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PNC TN9410 97-087, 142 Pages, 1997/07
Computer data analysis is planned as an essential process to facilitate and speed up the ISI of MONJU steam generator tubes using the ECT technique. This process compares the phase and amplitude of the signal in a vector window in order to identify and categories defects. The categorization of the inspection signal requires a high level of precision. The analysis test was carried out taking the best operational conditions for reference. From this, the most accurate classification conditions were established. The MONJU PSI signal data was used to check the effectiveness of the process. The results are as follows. (A) Verification of the set parameter for off line processing. Automatic classification is possible for almost all the support plate signals. Classification of all the weld and bend signals was not possible. Therefore, the set parameter was selected for the category in which there were the largest number of signals was established. (B) Verification of the analysis processing conditions. The established analysis conditions allow automatic classification for about 80 to 85% of the signal comparison factor cases. Furthermore, it is possible to classify all the signals by additional operator intervention. In this way it is possible to analysis and evaluate all the MONJU steam generator tube ISI data. (C) Improvement of the data base. Evaluation of MONJU PSI flaw detection data was carried out by set parameter analysis. FOllowing these results the necessary data base for ISI signal evaluation was created.
Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.
PNC TN9410 96-279, 51 Pages, 1996/08
Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.
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Proceedings of Australasian Universities Power Engineering Conference (AUPEC '96), 0 Pages, 1996/00
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Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR '91), ,
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Technical committee meeting on "Sodium removal and disposal from LMFR's in normal operation and in, ,
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