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Chatani, Keiji; ; ; Masui, Tomohiko*; Nagai, Akinori; ;
PNC TN9410 92-186, 63 Pages, 1992/06
Dose rates around UGT (Upper Guide Tube) of CRDM (Control Rod Drive Mechanism) have been measured in Experimental Fast Reactor "JOYO" during the 9th periodical inspection in order to reflect the study on the shield thickness of UIS (Upper Internal Structure) cask, which has been planned to be used for a Large Fast Reactor. Absolute amount of radioactive corrosion products (CP) is evaluated by gamma spectra analysis for waste water from cleaned UGT. The results on this study are summarized as follows: (1)Measured dose rates distribution around UGT before and after clean-up show the same reduction. The affection of CP is not clearly observed for the dose rate distribution. (2)The relative values of dose rate, which are evaluated by considering the inside structure of UGT, show the attenuation of 10
from bottom to sodium level of UGT. The above relative distribution agrees well with that of measurement data using U-235 fission chamber, which was conducted at MK-I core start-up tests, except the stellite region. (3)As to the relative values of dose rate, calculation by "DOT3.5" and estimation by measured dose rate agree within factor 3 for the attenuation of 10. It is confirmed that the calculation can predict well the measurement. (4)Absolute amount of CP estimated by gamma spectra analysis and waste water analysis is 180 MBq. 
Co dominates 92 % of CP. This value agrees with the prediction by corrosion product behavior analysis code "PSYCHE" within factor 2.
*; Haraguchi, Tetsuharu*; *; Tanigawa, Shingo*; Yamaguchi, Katsuhisa
PNC TN9410 88-107, 121 Pages, 1988/09
In the studies using PLANDTL, it would be planned to valid the thermal-hydraulic analysis codes which were developed each for whole system, plenum and subassembly, and also to evaluate the reactor plant in the future using these codes. SSC-L is to be as the main code in these studies and is used for design analysis through test analysis. In the first step of this study, model development and modification of SSC-L has been achieved for PLANDTL and predictive analyses have been applied as to validate the models and examine the design of PLANDTL. The estimated transient curves have been obtained about flow rate and temperatures at subassembly and loop of PLANDTL. As a result, the design conditions have been given to be able to perform the programmed tests. It have been validated that the conditions of tests would be within the design value, and the characteristics of PLANDTL and operational conditions have been obtained from the predictive analyses using design data of the plant. The modification and validation of SSC-L will be applied using the results of various kinds of functional tests, and test analyses will be performed in future.
Muramatsu, Toshiharu; *
PNC TN9410 88-064, 95 Pages, 1988/06
The existed experimental data for natural convection in an annulus between a reactor vessel and a rotating plug were reviewed to understand its phenomena for a prevention measure of sodium mist deposition. The generalized keypoints are as follows; (1)Natural convection in overall circumference of the annulus is decreased by reducing the annulus gap width. (2)In evaluation of sodium mist into the annulus, the effect of sedimentation by gravity for its transport can be neglected, and (3)Sodium mist transport due to natural convection was dominated by axial and circumferential temperature distributions in the annulus. A calculation using the mass transport version of three-dimensional thermal-hydraulic analysis code, AQUA was carried out for the experiments of convection barrier. In the calculation results, sodium mist concentration at the top of annulus agrees well with the experiments.
