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Ashikagaya, Yoshinobu; Nakazawa, Takashi; Yoshino, Toshiaki; Yasu, Katsuji
JAERI-Tech 2001-092, 76 Pages, 2002/01
JAERI-Tech-2001-092.pdf:11.72MB
no abstracts in English
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JNC TN4400 99-002, 192 Pages, 1999/03
The tritium transport analysis code, TTT, has been validated using data from the low power test of Monju, and then its behaviour at along term full power operation of Monju in future has been estimated, when the estimated transport and distribution of tritium in the reactor system has been also compared with the result in Joyo and Phenix, which had been already experienced long term operations. The TTT code had been develpped using the tiritium and hydrogen transport model proposed by R. Kumar, ANL, and had been applied to the evaluation in Monju design work. After then, futhermore, the code has been improved using the data from long term operation of Joyo with MK-II core, and in this work the code has been validated for the first time for Monju data. The results from this work are as follows; (1)Comparison of the best fitted tritium source rates from cores in Joyo, Phenix and Monju makes an estimation of the major source from control rods, (2)The calculated tritium concentration in each medium for cooling and its change is a reasonable agreement to the measured, C/E=1.1, (3)The cover gas transport model cosidering isotopic exchange of H and H
can reproduce reasonably the measured concentration distirbution of tritium in sodium and cover gas, (4)The tritium concentration in secondary sodium of Monju was about l/50 times as much as the primary one, which shows the acceraration effect on cold tarapping of tritium due to coprecipitation with permeated hydrogen through Evaporater (EV) heat conduction tube walls. The tritium cold trapping efficiency was estimated to be 1 for coprecipitation with hydrogen and 0.3 for isotopic exchange, respectively, (5)Tritium transport and distribution for along term full power operation of Monju in future was estimated, which could involve a excess factor to 4 at the maximum. The tritium concentration in sodium and Steam Generator (SG) water will be substantially saturated after somthing like 10 years full power operation, ...
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PNC TN9410 98-053, 43 Pages, 1998/04
In the Power Reactor and Nuclear Fuel Development Corporation (PNC), the following are examined as part of an application technology using a high power electron linac : monochromatic gamma ray sources, free electron lasers, and intense positron sources. This report describes an adaptive estimate of a superconducting magnet in order to efficiently converge a positron beam for the development of an intense positron source. By comparing the intensity of a positron beam using a superconducting magnet with a normal conducting magnet, the intensity obtained was more than double. In addition, a small magnet was manufactured in order to examine the characteristics of the superconducting magnet as a solenoid coil. An excitement test was carried out with rated current. As a result of measuring the maximum magnetic field on the central axis, we achieved 5.6 Tesla, which was the designed value. Therefore, it was confirmed to function as a focusing device, when the superconducting magnet was used to converge the positron beam.
Otaka, Masahiko; Ohshima, Hiroyuki; Ninokata, Hisashi;
PNC TN9410 96-212, 36 Pages, 1996/06
A single phase subchannel analysis code ASFRE-III has been developed at PNC for predicting behavior of coolant and fuel pin temperature distributions in a fast reactor fuel subassembly under various operation and accident conditions such as a local flow blockage event. Salient features of the code are: a distributed resistance model of wire-wrap spacers, a porous blockage model, and an efficient matrix solver suitable for a large vector/parallel computation. In this study, ASFRE-III was applied to the thermal-hydraulic analysis of the two out-of-pile experiments using sodium performed at PNC for the purpose of the code validation. The one was performed around rated flow and heat flux conditions and the other was decay heat removal conditions. The computational results obtained under various flow and heat flux conditions were compared with the experimental data. The predicted coolant temperatures in subassemblies were agreed well with the measured data within 5 
6% in the wide range from low to high Reynolds number regions.
Muramatsu, Toshiharu; *
PNC TN9410 92-106, 354 Pages, 1992/04
A natural circulation analysis in the upper plenum of the MONJU reactor was conducted for transient simulating a pump coast down and reactor scram to a full-power operation condition using a multi-dimensional code AQUA. In the analysis, full options of the AQUA code (higher-order differencing schemes, an algebraic stress turbulence model, an adaptive Fuzzy control system, etc.) were used to obtain a refined numerical result. From the analysis, the following results have been obtained. (1)In a steady-state calculation simulating the full-power operation condition, maximum axial temperature gradient 154
C/m was calculated at the region between the upper and the lower flow holes. Therefore detailed measurements are necessary for thermal stress evaluation of internal components due to the axial temperature gradient at various power operation conditions. (2)In a transient caluculation simulating a natural circulation phenomenon, it was confirmed that a rising speed of the thermal stratification interface is delayed due to the decrease of a effective mixing volume in the upper plenum region. And the AQUA code calculated a discontinuity temperature transient (a hot shock continued from a cold shock) at the outlet nozzle of the reactor vessel due to the change of locally flow patterns in the upper plenum. Therefore it was concluded that detailed investigation is necessary using experimental data in various power operation conditions. (3)A gentle temperature transient was calculated with the AQUA code in comparison with a one-dimensional code. It is concluded that the one-dimensional code yields a conservative numerical result.
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 91-002, 110 Pages, 1991/08
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 91-001, 83 Pages, 1991/01
no abstracts in English