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Ishikawa, Koki; Takamatsu, Misao; Kawahara, Hirotaka; Mihara, Takatsugu; Kurisaka, Kenichi; Terano, Toshihiro; Murakami, Takanori; Noritsugi, Akihiro; Iseki, Atsushi; Saito, Takakazu; et al.
JAEA-Technology 2009-004, 140 Pages, 2009/05
JAEA-Technology-2009-004.pdf:2.0MB
Probabilistic safety assessment (PSA) has been applied to nuclear plants as a method to achieve effective safety regulation and safety management. In order to establish the PSA standard for fast breeder reactor (FBR), the FBR-PSA for internal events in rated power operation is studied by Japan Atomic Energy Agency (JAEA). The level1 PSA on the experimental fast reactor Joyo was conducted to investigate core damage probability for internal events with taking human factors effect and dependent failures into account. The result of this study shows that the core damage probability of Joyo is 5.0
10
per reactor year (/ry) and that the core damage probability is smaller than the safety goal for existed plants (10 ry) and future plants (10
/ry) in the IAEA INSAG-12 (International Nuclear Safety Advisory Group) basic safety principle.
Ito, Chikara; Ito, Hideaki; Ishida, Koichi; Hatoori, Kazuhiro; Oyama, Kazuhiro; Sukegawa, Kazuya*; Murakami, Takanori; Kaito, Yasuaki; Nishino, Kazunari; Aoyama, Takafumi; et al.
JNC TN9410 2005-003, 165 Pages, 2005/03
JNC-TN9410-2005-003.pdf:12.66MB
At experimental fast reactor JOYO, appraisal of detection efficiency of behavior and FFD and FFDL of the fission product which is discharged inside the furnace as one of safety research of the country, is carried out. In MK-II core, the slit in the gas plenum part of the test sub-assembly, the test which irradiates this(1985 April, FFDL in-pile test(I)), providing the slit in the fuel column part of the test sub-assembly, the test which it irradiates(1992 November, FFDL in-pile test(II)) were carried out.MK-III reactor core replacement was completed and started in 2004. That the behavior in the system of FP with the reactor core replacement and so on changes in the MK-III reactor core and to have an influence on the sensitivity and the replying of FFD and FFDL are thought of. Therefore, behavior of FP in the fuel failure in the MK-III reactor core, the performance of FFD and FFDL must be confirmed beforehand. Moreover, to prepare for the fuel failure and the RTCB test which is doing a future plan, and to confirm a plant operation procedure in the fuel failure in MK-III reactor core operation and to attempt for the correspondency to improve are important.Therefore, in the period from 2004 November 11th to November 29th, it carried out the FFDL in-pile(III). It did a series of plant operation to stop a nuclear reactor after loading a reactor core center with the fuel element for the test which provided an artificial slit for the fuel cladding in the MK-III reactor core and irradiating it and detecting fuel damaging and to take out fuel. And it confirmed the operation procedure of the fast reactor in the fuel failure.Also, the improvement items such as the improvement of the operation and the procedure and the remodeling and the service of the facilities could be picked up. In the future, it attempts these compatible, and it prepares for the MK-III reactor core operation and it incorporates a final examination result by the improvement of the safety of FBR.
; Aita, Tsuyoshi; Murakami, Takanori; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro
JNC TN9410 2004-006, 36 Pages, 2004/03
Joyo Plant Operation Management Expert Tool system named JOYPET has developed with the aim of confirming the stable and safety operation of JOYO and improving operational reliability in future FBR plants.New JOYPET system was designed and manufactured in 2002, and began or operation in 2003, because the former system, which was designed in 1988 and operated from 1991 to 2002, was superannuated, and it was difficult to obtain alternative hardwares and replace parts.The difference between the former one and the later new one was adopted the web-online system to use lan(Lacal area network) instead of the host and the terminal computer processing system.Then the new system enabled to take unitary document management for reactor operation, and each person in one's rost was able to search, refer and wake document on line directly.This document reported new JOYPET system design, manufacturing, system constitution and operation actual result.
Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro; Sekine, Takashi; Ariyoshi, Masahiko; Ito, Chikara; Nemoto, Masaaki; Murakami, Takanori; Isozaki, Kazunori; Hoshiba, Hideaki; et al.
JNC TN9410 2003-011, 197 Pages, 2004/03
JNC-TN9410-2003-011.pdf:10.26MB
MK-III performance tests began in June 2003 to fully characterize the upgraded core and heat transfer system. Then, the last pre-use inspection was finished in November 2003.This report summarize the result of each performance test.
; Terano, Toshihiro; ; Onuki, Osamu; ; Okubo, Toshiyuki;
PNC TN9410 96-103, 88 Pages, 1996/03
This paper describes operational experiences of JOYO Secondary Sodium Purification System which were obtained from April 1990 to March 1995. And, the operational problem of Secondary Sodium Purification System with MK-III core conversion were mentioned in this report. The experience results were as follows. (1)The number of cases of the secondary purification system trouble was 12. Those troubles, however, did not affect to the operation of the plant. (2)As the result of investigation about possibility of using secondary cold trap after MK-III core conversion, amount of trapped impurity was estimated 9.4kg as of January, 1994. In addition, 25.2kg was the estimated amount of trapped impurity at sodium initial purification after MK-III core conversion. It was subtantial excess of original design(18kg). (3)The change of heat transfer characteristics of secondary cold trap economizer occurred about two years after the modification of cold trap control tempereture. It seemed to be led by condition change of sticking impurity to heat transfer tubes which was followed by the modification of set up temperature. When set up temperature had been kept low, heat transfer characteristics was better. (4)A large quantity of sodium vapor was identified on the secondary argon gas system pressure control header. The influence of engulfing of argon gas at over flow line of Secondary Sodium Purification System is concerned.
*; ; *; *; *; *;
PNC TN9410 89-138, 50 Pages, 1989/08
Cover gas monitorring system was installed for obtaining the information of impurities concentration in the primary and secondary cover gas on JOYO, and it has been operated to these days. The experience results are as follows; (1)Increasing of impurities concentration in the cover gas were detected quickly by this system. (2)This system was able to detected the impurities concentration continuously, so it is very useful for investigation the cause of abnormal condition and observation the effect of cover gas purge. (3)This system was modified some points, consequently the measurement performance and reliability of data were advanced more than old one.
Sato, Kazujiro; ; Kamide, Hideki; Murakami, Takanori;
Donen Giho, (60), p.53 - 57, 1986/12
None
*; Hayashi, Kenji; *; Sato, Kazujiro*
PNC TN941 85-152, 48 Pages, 1985/10
A simple geometry model test was performed to study natural circulation thermohydraulics in the Monju EX-Vessel Storage Tank (EVST), using water as a working fluid. Thermohydraulic data in steady state natural circulation were obtained from measurements of temperature, velocity and flow pattern profiles. Also, analytical results by COMMIX-DRACS were compared with experiments to examine applicability of the code to the buoyancy dominated flow. The flow in the model is extremly complex and varies in space and time, however, it is comfirmed that temperature variations in circumferential directions are nearly uniform. COMMIX-DRACS results indicate relatively good agreement with the basic physical effects shown by the experimental data. Both analytical flow and temperature profiles match with experiments. Thus, the COMMIX-DRACS code is capable of predicting thermohydraulics in the Monju EVST by the use of the proposed analytical model in the present paper.
*; *; Sato, Kazujiro*
PNC TN941 85-98, 51 Pages, 1985/07
Hydraulic tests of the corse control rod assembly were performed to collect experimental data required for an evaluation of core flow distribution in the Monju reactor by using the Water Test Loop. For simulating reactor operating conditions, pressure loss and intra-flow distribution characteristics were measured with varying an axial location of the control rod bundle, which is movable in the range from 0 mm to 1000 mm. In addition, effects of the FIV (Flow-Induced Vibration) restraint buttons were examined by mounting two accelerometers at the lower end of the bundle protect tube. Pressure loss characteristics are independent of axial locations of the control rod bundle in the range from 300 mm to 900 mm. Intra-flow distributions are also constant above 200 mm and about 45% of the total flowrate flows into the control rod bundle. The axial location of the bundle varies from 465 mm to 765 mm during the first and last core operations of Monju. Thus, it is indicated that hydraulic charaeteristics of the control rod assembly can be treated as to be constant in working ranges. Amplitudes of FIV at the measurement points do not decrease with mounting the FIV restraint buttons. However, the buttons are effective in decreasing anisotropic rolling of the protect tube. Experimental data obtained in the present tests will be reflected on the control rod assembly and core thermo-hydraulic designs of Monju.
*; *; Sato, Kazujiro*
PNC TN945 85-07, 85 Pages, 1985/06
The 1/2-scale model of the LMFBR Monju reactor was constructed to study core flow distribution using water as a working fluid. This paper presents the preliminary test results for core elements (core subassemblies, control rods, etc.) in the model concerning with hydraulic characteristics. Calibrations of the instrumental turbine flow-meters in the subassemblies have been also performed in the test. Hydraulic characteristics of three core elements in the each flowrate zone were measured and then experimental equations of flow resistance coefficients, 
, were obtained as a function of Reynolds Number, Re. Also, correlations between the pulse count number and the flowrate were obtained in the calibration tests. The above basic data about the 1/2-scale reactor model are essential for both the experiment and the analysis of the follow-on core flow distribution test. The results will be reflected on the core thermo-hydraulic design of Monju through the follow-on test.
