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Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Miura, Kuniaki*
Nihon Kikai Gakkai 2002-Nendo Nenji Taikai Koen Rombunshu, p.273 - 274, 2002/09
Corrosion-erosion properties of type 316 austenitic stainless steel were investigated at 450oC. The study aims at developing ADS, accelerator driven system, for nuclear transmutation of long lived activated nuclei to shorter ones. After 3000 hrs flow of eutectic 45Pb-55Bi loop tubes were cut and analyzed by optical microscope, SEM, EDX, WDX and TEM. It is concluded that corrosion-erosion depth is maximally 0.1mm per 3000 hrs and Cr-Fe crystals were precipitated in the lower temperature of flowing channel. Further more inspection results of Electro Magnetic Pump, Electro Magnetic Flow meter and controlling valve were also reported. Output signal from EMF was stable after certain time duration. In this experiment oxygen content in Pb-Bi was not actively controlled but Pb-Bi was covered by 4N Argon gas.
Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Nakamura, Fumihito*; Ohashi, Masahisa*
JAERI-Tech 2000-044, 25 Pages, 2000/06
JAERI-Tech-2000-044.pdf:4.64MB
no abstracts in English
; Nemoto, Masaaki; Saikawa, Takuya*; Sukegawa, Kazuya*
JNC TN9410 2000-008, 66 Pages, 2000/03
JNC-TN9410-2000-008.pdf:1.39MB
The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.
Soroi, Masatoshi; Miyakawa, Shunichi
PNC TN9410 98-050, 57 Pages, 1998/05
This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor "JOYO". FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. The main results and problems of the development of FLORA are as follows; (1)The results of the development: (a)The neutron detector in FLORA can detect the delayed neutron which is emitted from failed fuel. (b)Out-of-pile A-LIP tests in sodium conditions has been completed. (The length of the tested A-LIP is half the actual size.) Out-of-pile test results showed that the A-LIP achieved a 300
/min flow rate and 265kPa pressure in 550
C sodium. This pump performance satisfied the FLORA requirements. (c)By controlling the sodium flow rate from 40 to 100% using the A-LIP, we can control the fuel cladding temperature satisfactorily. (2)The problems: (a)In the development of the process detector, it is necessary to miniaturize the neutron detector and test the effect of neutron irradiation and high temperatures on the permanent magnet in the flow meter. (b)The problem which is left about A-LIP is its influence on neutron irradiation. For this purpose, we have to irradiate a small size A-LIP and test its characteristics and electric isolation. (c)To get more accurate results concerning the efficiency of the A-LIP, we have to ...
; ; ; ; 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.
*; 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.