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; Moro, Satoshi; ; Kakehi, Isao; ;
PNC TN9410 98-033, 284 Pages, 1998/03
System engineering division of OEC has being carried out a design study of the advanced nuclear fuel recycle system using electro-metallurgical process, aiming for improvements in safety, reliability, economy and a1so in environmental burden and nuclear non-proliferation. But the public criticism against nuclear power is more severe recently, and the situation is changing as seeing in the conclusion of the round-table conference on FBR. The researcher's meetings, in which researchers in PNC and from other organizations attended, were held during December, 1997 and March, 1998 in order to discuss on the advanced nuclear fuel recycle system and technology for FBR to be aimed in the future, and how to execute its research & development, etc. The conclusions of this meeting are as follows: (1)The future advanced FBR fuel cycle system shall be the system which has high potential for maximum utilization of uranium resources, and also for revolutionary improvements of economy, safety, environmental burden, etc. so as to be accepted in the society. (2)Regarding to the process of the future fuel eycle system, electro-metallurgical process that is able to apply for reprocessing of different types of fuel (oxide, metal and nitride) and is flexible for technical progress is recommended. Research & development of this system and technology shall be carried out. (3)The mission of PNC (new organization) is to select the most appropriate advanced FBR fuel cycle system from the viewpoint of the long-term FBR age in the future, and to conduct development of its system. It is expected for the new organization to execute its research and development steadily in cooperation with other research institutes, etc. under the nation-wide assessment and agreement. According to the above conclusions, the system engineering division will enhance the design study of the advanced FBR fuel cycle system and establish the definite concept of the system in cooperation with concerned in and ...
Hunter
PNC TN9410 98-015, 81 Pages, 1998/02
The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only 
30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...
; Funasaka, Hideyuki; ; Koyama, Tomozo
PNC TN8600 97-007, 109 Pages, 1997/11
no abstracts in English
Kawabe, Ryuhei*; Himeno, Yoshiaki; Kawada, Koji*; Miyaguchi, Kimihide
PNC TN941 85-104, 17 Pages, 1985/06
Flow and combustion test of low temperature sodium (250
C) on a simulated for liner has been conducted to give an answer to the possible flow blockage or flow plugging. The simulated floor liner used for this purpose was 2.4m in length and 1.2m in width having liner gradient of l/100. The bottom surface of the liner was well thermally insulated. In the test, 160kg of sodium was slowly spilled from a nozzle having a wide opening at flow rate of 1 
/sec for 200 sec. The nozzle was attached to the side of the liner. Flow pattern and combustion characteristics of sodium have been monitored during the test, and temperatures of the flowing sodium and a liner steel have also been measured. In the post-test examinations, distribution of residual sodium and sodium oxide on the floor liner as well as that in a drain pipe was determined. The results thus obtained were summarized as follows. (1)At beginning of the test, although the spilled sodium froze for a certain period of time due to its heat transfer to the liner, it remelted by taking heat from a successive flowing sodium at higher temperature. Therefore, on the liner sodium flowed continuously without being blocked its flow path. (2)Heat flux from sodium to the liner was less than 80kw/m
, while related heat transfer coefficient was 300 500w/mC. The latter value was almost the same to that obtained from the similar test with hot sodium (505C). (3)Post-test examination revealed that the distribution of residual sodium and sodium oxide on the floor liner was almost uniform with the average value of 1kg/m. No massive combustion products that may cause flow plugging was found in a sodium drain pipe.
