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Kaneko, Yoshihisa*; Nishikimi, Masakazu*; Shirato, Seiichi*; Morimoto, Makoto*; *
PNC TN941 85-25, 347 Pages, 1985/02
Nickel membrane type hydrogen meter is used to detect earlier water leakage from heat transfer tubes in the steam generator for Fast Breeder Reactor. The nickel membrane is one of the most important pieces for the hydrogen meter and is used under the severe condition such as high temperature of 500
C so that it may happen that corrosion, conversion in microstructure and shortage of mechanical strength etc. Various material tests for 11 kinds of nickel membranes used in the test facility for a longterm were done and following results were obtained. (1)A little corrosion was observed for the nickel membrane used over a long term. (2)The microstructure of the nickel membrane have been converted to recrystal microstructure. The microstructure for the weld is dendritic structure and it has atendency of the growth of crystal grain by the welding heat. (3)The tensile strength of the nickel membrane is decreased slightly, but, it keeps sufficiently the permissible tensile strength. The hardness of the nickel membrane have been decreased to the hardness of annealing material. (4)Iron, chromium and manganese etc. were attached to the surface of the nickel membrane and especiallyy the manganese soaked into several micron in the nickel membrane. But, these deposits haven't any effect for the hydrogen meter. (5)The leak position on the nickel membrane was found in the welding side between the nickel membrane and the vacuum pipe (made of stainless steel). Then, one of important factors having an influence on the membrane's life may be the welding.
Kaneko, Yoshihisa*; Nishikimi, Masakazu*; Shirato, Seiichi*; Morimoto, Makoto*; *
PNC TN941 85-24, 194 Pages, 1985/02
Series of tests on the water leak detection system have been conducted in the 50MW Steam Generator Test Facility since December 1974. This report is data of water leak simulation tests (water or hydrogen injection tests) which were carried out until November 1983 since May 1978. These data (response of water leak detector) are recorded in magnetic tape and it is easy to change scale and soon.
Water leak alarm systemShirato, Seiichi*; Kaneko, Yoshihisa*; Nishikimi, Masakazu*; Morimoto, Makoto*; Matsumoto, Shigeharu*; Motonaga, Tsutsuji*; Imai, Hiroshi*; Fukuda, Toru
PNC TN941 85-23, 193 Pages, 1985/02
To develop the small leak sodium-water reaction detection system in the steam generator for Fast Breeder Reactor, the series studies on system evaluation tests have been carried out at the 50MW Steam Generator Test Facility. This report deals with hydrogen and injection tests into the sodium outlet pipe from the superheater, test plans and test equipments for hydrogen injection tests into the evaporator, the response performances for the hydrogen meter and oxygen meter, detection ratio for hydrogen and oxygen and the water leak alarm system etc. The test results are summarized as follows: (1)Hydrogen injection tests & water injection tests into the sodium outlet pipe from the superheater were done and the hydrogen & oxygen behaviors were understood. (2)Hydrogen injection tests at the annulus, down-comer and cover-gas region in the evaporator were done and hydrogen behaviors were understood. (3)Hydrogen detection rates depending on hydrogen injection tests & water injection tests were arranged and factors giving the effect on the rate were evaluated. Also, hydrogen transfer times were evaluated. (4)Hydrogen detection rates at the water leakage in the evaporator were evaluated by the comparison of hydrogen injection test results & water injection test results into the sodium out1et pipe from the superheater and hydrogen injection test results into the evaporator. (5)The water leak alarm system by the computor was completed and the validity was confirmed.
Kaneko, Yoshihisa*; Nishikimi, Masakazu*; Shirato, Seiichi*; *
PNC TN941 85-22, 142 Pages, 1985/02
Series of tests on water leak detection system for Monju steam generators have been excuted in the 50MW Steam Generator Test Facility (50MW SGTF). This paper describes the test results about the removal efficiency of hydrogen by the secondary cold trap, the hydrogen background concentration and hydrogen flux through heat transfer tubes to secondary sodium in steam generator (SG) in order to make clear the hydrogen behavior during ordinary plant operation from October in 1976 to July in 1983. Main results from these tests are: (1)The removal efficieney of hydrogen by the cold trap is 70
80 percents, but it shows tendency to decrease when difference between hydrogen concentration at cold trap and that in secondary main loop is small. (2)Hydrogen flux through tubes is smaller than the value which is reported in the first report of these series reports. (3)It is found that the hydrogen background concentration in the secondary main loop of 50MW SGTF is thoroughly low. (4)It is estimated that the hydrogen background concentration in the secondary main loop in Monju plant is lower than 169 ppb.
Shirato, Seiichi*; Kaneko, Yoshihisa*; *; Nishikimi, Masakazu*; *; *; *
PNC TN941 84-136, 258 Pages, 1984/10
It is necessary to detect the water leak (sodium-water reaction) as fast as possible in a steam generator for liquid metal fast breeder reactor, and hydrogen meter has been developed for detection of water leak. In-sodium hydrogen meter and in-cover gas hydrogen meter have been developed in 50 MW steam generator test facility for the purpose of leak detection system of LMFBR "Monju" plant. This paper reports the development of each hydrogen meter, operational experience and evaluation of performance. Some proposals for design, manufacture, operation and maintenance of "Monju" plant 1eak detection system are also described.
Kaneko, Yoshihisa*; *; Nishikimi, Masakazu*; *; *
PNC TN941 84-93, 139 Pages, 1984/06
Rupture disks were tested for a purpose of examination of burst characteristic, corrosion, change of metal formation and mechanical characteristic. Those rupture disks were installed at evaporator (reverse buckling type), superheater (reverse buckling type) and reaction product vessel (tension loaded type) and were used from September, 1975 to January, 1983 at 50 MW steam generator test facility (50 MW SGTF). Main results are the following; (1)Burst pressure of used and unused rupture disks of steam generator and reaction product vessel was lower than the value measured at producted time. But it satisfied specification in purchase. (2)Burst characteristic of rupture disk for steam generator did not Change even after they were used for 1013812387 hours through at the condition of 200430C and 0.7 1.4kg/cm
g. (3)Rupture disks of steam generator and reaction product vessel opened perfectly at burst test. (4)Corrosion was scarcely found. (5)Metal formation and mechanical and physical characteristics of material (Inconel X-750) of rupture disk has changed (age hardnening) through environment effect. (6)If we use Inconel X-750 as rupture disk material for steam generator, we must put in operation after ageing treatment. Otherwise it is desirable to use Inconel 600 or 625 which is not effected by age hardnening and has heatproof and corrosion-resistance characteristics.
