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Tobita, Kenji; Hamamatsu, Kiyotaka; Harano, Hideki*; Nishitani, Takeo; Kusama, Yoshinori; Kimura, Haruyuki; Takizuka, Tomonori; Fujieda, Hirobumi*; Shoji, Teruaki; Senda, Ikuo*; et al.
Proc. of 5th IAEA Technical Committee Meeting on Alpha Particles in Fusion Research, p.45 - 48, 1997/00
no abstracts in English
; ; ; ; Ozawa, Kenji; ; Terunuma, Seiichi
PNC TN9410 91-187, 41 Pages, 1991/07
A reactor vessel in JOYO provides three induction type level meters which is defined in the safety protection system. They have two kinds of measuring range and display the sodium level below to the discharge nozzle of the primary cooling system. One is from 350mm about the normal sodium level to 1,600mm below it and other two sets are from 350mm above to 350mm below it. This report describes a special monitoring method of sodium level in the occurrence of the abnormal sodium level incident which reaches it more than 1600㎜ below the normal sodium level in the reactor vessel. The special monitoring method uses the discharge sodium pressure of the primary auxiliary cooling pump. A discharge sodium pipe from the primary auxiliary cooling pump is located in the bottom of the reactor vessel and it's discharge pressure is correlated with the reactor vessel sodium level which works back pressure to the pump. Therefore, it was assumed that abnormal sodium level which reaches it more than 1600mm below the normal sodium level can be monitored using this discharge sodium pressure. A verification test was conducted to measure the correlation of the discharge sodium pressure and the reactor vessel sodium level. Main results obtained from this test were as follows. (1)Validity of this special level monitoring method was confirmed in the sodium level range from normal to 3,390㎜ below it and in case of sodium level changing which is decreased at the rate of 47.5m
/h by this test during the system sodium drain work. (2)A correlation equation is obtained using parameters of discharge sodium pressure, flow and temperature of the primary auxiliary cooling system to gain sodium level of reactor vessel. (3)Parametor chart of the reactor vessel sodium level was made using multi regressive analysis.
*; *; *; *; *; *; *
PNC TN9410 87-001, 22 Pages, 1987/01
This report presents an operation experience of the auxiliary cooling system from January, 1982 to September, 1986. Outline of the operation experience is following ; (1)There has not been any serious trouble on the primary auxiliary cooling system in the term. Operation time of the primary auxiliary cooling system to remove reactor core residual heat during the reactor shut down outage (refueling outage) was about 3400 hours. The primary auxiliary circulating pump has been automatically started 65 times on the scheduled tests. Total operation time of the primary auxiliary cooling system was about 3900 hours after initial sodium filling of the cooling system. (2)Operation time of the secondary auxiliary cooling system with 100% flow rate was about 34000 hours. The secondary auxiliary circulation pump was tripped 11 times during about 34000 hours operation by; (a)Loss of commercial power and various tests, -10 times (b)High temperature of circulation pump coil, - 1 time. High temperature of circulation pump coil was caused by loss of cooling air with the obstacles in the cooling pump filter in May, 1982. This caused no influence to the plant at that time since the reactor decay heat had been sufficiently cooled down. The secondary auxiliary cooling system has been in no trouble except for the trip by high temperature of circulation pump coil. Total operation time of the secondary cooling system after initial sodium filling of the system was about 73000 hours.
Nakano, Makoto*; *; *; Inoue, Teruji*
PNC TN942 84-01, 50 Pages, 1984/04
This report describes incidents occuring during the Period from April 1982 to March 1983. During this period Mark-II core installation was carried out and Mark-II initial criticality was attained on 22nd of November successfully. 290 incidents required repair. Of these 13.4% were associated with the secondary sodium cooling system and 11.7% with the primary sodium cooling system. From the analysis of these incidents, the frequency of incidents involving instrumentation is highest, with that involving valves and control panels also high. These highest three systems occupy larger portion than previous year.
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PNC TN941 83-27VOL2, 456 Pages, 1983/02
PNC-TN941-83-27VOL2.pdf:20.9MB
None
*; *; *; *; *; *; *
PNC TN941 83-27VOL1, 827 Pages, 1983/02
PNC-TN941-83-27VOL1.pdf:25.12MB
In the Experimental Fast Reactor "Joyo", Core Conversion from the Breeding Core (MK-I Core) to the Irradiation Core (MK-II Core) was begun in January 1982. Core Conversion required the refueling of 290 core elements, reconstruction of the control rod drive mechanism, and change-out of both the upper guide tube and lower guide tube of the control rods. The schedule for these activities was planned carefully to achieve 100MWt power in March 1983. Joyo achieved MK-II core initial criticality on November 22, 1982, on schedule and immediately began core characterzation. As the result of the Core Conversion activities, we handled many core elements, twice as many as before this core conversion, and in so doing obtained many kinds of experience and data for maintenance and operation of the refueling system. This report describes the considerable pre-conversion activities and planning, and the accomplishments and results of these Core Conversion activities.
Miyamoto, Seiji; Nakamura, Yukiharu*; Fujieda, Hirobumi; Hamamatsu, Kiyotaka; Oikawa, Toshihiro; Sugie, Tatsuo; Kusama, Yoshinori; Yoshino, Ryuji
no journal, ,
Recently, we developed a simulation model in which an ECRF ray tracing and current drive calculations are combined with TSC. This model is applied to the evaluation of magnetic flux consumption in the ITER current ramp-up scenario. In this model, real geometry of PF/CS coils and EC launcher is taken into account, and EC deposition/current drive profile are calculated in self-consistent with the plasma profile evolution. Central current drive (present ITER design) and off-axis current drive (test case) is compared. Resistive flux is lowered in both cases. Internal flux is also reduced by the off-axis EC due to reduction of internal inductance. It is although shown that, even in the case of central EC, comparable reduction of internal flux is expected due to the skin effect of inductive current.
Takechi, Manabu; Fujieda, Hirobumi; Sakurai, Shinji; Matsukawa, Makoto; Masaki, Kei; Shibama, Yusuke; Higashijima, Satoru; Shibanuma, Kiyoshi; Sakasai, Akira
no journal, ,
no abstracts in English
Takechi, Manabu; Fujieda, Hirobumi*; Sakurai, Shinji; Masaki, Kei; Shibama, Yusuke; Matsunaga, Go; Shibanuma, Kiyoshi; Sakasai, Akira
no journal, ,
We performed the evaluation of over current of poloidal coils and EM force of in-vessel coils with DINA code during disruption for design of in-vessel components of JT-60SA. We performed disruption simulation for the case of upward and downward VDE and major disruption of 5.5MA plasmas at end of burn. The maximum induced current of poloidal coils is about 1.7kA. It is found that amplitude of induced current does not depend on the direction of the plasma movement at disruption and current quench time. EM force of the fast position control coil with 24 turns is 1.2 times larger than that with 16 turns.