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Ohgama, Kazuya; Takegoshi, Atsushi*; Katagiri, Hiroki; Hazama, Taira
Nuclear Technology, 208(10), p.1619 - 1633, 2022/10
Times Cited Count：3 Percentile：68.71(Nuclear Science & Technology)Yabe, Takanori; Murakami, Makio; Shiota, Yuki; Isobe, Yuta; Shiohama, Yasutaka; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro
JAEA-Technology 2022-002, 66 Pages, 2022/07
JAEA-Technology-2022-002.pdf:10.45MB
In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the Ex-Vessel fuel Storage Tank. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the Ex-Vessel fuel Storage Tank are all transferred to the storage pool. In fiscal 2018, as "Fuel Treatment and Storage", 86 fuel assemblies were transferred to the storage pool. As "Fuel Unloading", 76 dummy fuel assemblies were stored in the Ex-Vessel fuel Storage Tank. In fiscal 2019, as "Fuel Unloading", 60 fuel assemblies and 40 blanket fuel assemblies were unloaded from the core. These assemblies were stored in the Ex-Vessel fuel Storage Tank, and dummy fuel assemblies were loaded into the core instead. During these operations, a total of 38 cases of alarming or equipment malfunctions classified into 24 types occurred. However, no significant events that menaces to safety have occurred. The operations were continued safely by removing the direct factors for the malfunctions in the equipment operation and performance.
Shiota, Yuki; Yabe, Takanori; Murakami, Makio; Isobe, Yuta; Sato, Masami; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro
JAEA-Technology 2022-001, 117 Pages, 2022/07
JAEA-Technology-2022-001.pdf:25.55MB
In the first stage of Monju decommissioning project, fuel unload work began to be carried out. There are two tasks in this work. One is fuel treatment and storage work that gets rid of sodium on the fuel assemblies unloaded from Ex-Vessel fuel Storage Tank (EVST) and carries it in the storage pool, and the other is fuel unloading that the fuel assemblies in the reactor core is replaced with dummy fuels and stored in EVST. Fuel treatment and storage work and fuel unloading work are performed alternately, and 370 bodies in the core and 160 pieces in EVST are all carried to the storage pool. 86 fuel assemblies was carried to the storage pool in fuel treatment and storage work in 2018 and 76 dummy fuels were stored in EVST for fuel unloading work. During the work, 86 types and 232 alarms / malfunctions occurred, but there was no impact on safety. There was one equipment's failure at gripper's claw open / close clutch of ex-vessel fuel transfer machine B, but it was repaired and restarted. Also it was eliminated the cause of problem or concession that the equipment failure due to the sticking of the sodium compound and continuous use of the equipment. Some problems related to system control occurred, but the work was done after checking the safety. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.
Ito, Kenji; Kondo, Tetsuo; Nakamura, Yasuyuki; Matsuno, Hiroki; Nagaoki, Yoshihiro; Sakuma, Yuichi
Dekomisshoningu Giho, (63), p.1 - 26, 2022/05
The prototype advanced thermal reactor Fugen entered into the decommissioning phases with the approval of the decommissioning plan in February 2008. The prototype fast breeder reactor Monju entered into the decommissioning stage with the approval of the decommissioning plan in March 2018. In April 2018, the head office of Tsuruga decommissioning demonstration was newly established to oversee the decommissioning operations in Tsuruga area, and decommissioning projects for two unique reactors have progressed safely and steadily.
Sotsu, Masutake
Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08
Thermal expansion behavior was investigated for evaluation of the core support plate expansion reactivity in the Unprotected Loss of Heat Sink reactor trip failure event. A possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype fast breeder reactor Monju. The reactor core expansion was simulated in a three-dimensional finite element analysis model of the reactor vessel considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model. It was found that the thermal expansion of the core was not restrained in the ULOHS evert, although part of the core structure is mechanically restrained.
Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru
JAEA-Technology 2019-020, 167 Pages, 2020/03
JAEA-Technology-2019-020.pdf:21.06MBJAEA-Technology-2019-020-high-resolution1.pdf:47.3MBJAEA-Technology-2019-020-high-resolution2.pdf:34.99MBJAEA-Technology-2019-020-high-resolution3.pdf:48.74MBJAEA-Technology-2019-020-high-resolution4.pdf:47.83MBJAEA-Technology-2019-020-high-resolution5.pdf:18.35MBJAEA-Technology-2019-020-high-resolution6.pdf:49.4MBJAEA-Technology-2019-020-high-resolution7.pdf:39.78MB
The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.
Koga, Kazuhiro*; Suzuki, Kazunori*; Takagi, Tsuyohiko; Hamano, Tomoharu
FAPIG, (196), p.8 - 15, 2020/01
The prototype fast breeder reactor Monju has already started (from June 2017) the unloading operation period (about 5.5 years: until the end of 2022) of the fuel assembly, which is the first stage of decommission. Among them, the first "Processing of fuel assembly" operation (86 in total) was conducted from August 2018 to January 2019 as the first handling of the fuel assembly. Fuji Electric provided technical support, such as dispatching technicians throughout the period, in cooperation with Japan Atomic Energy Agency for the "Processing of fuel assembly" operation, and contributed to the completion of the operation while experiencing various troubles. This manuscript introduces the contents of the first "Processing of fuel assembly" operation and the overview of the trouble status. This manuscript is a sequel to FAPIG No.194 "Prototype Fast Breeder Reactor Monju Decommissioning and Unloading Operation of the Fuel Assembly from the Core", please refer to it.
Sotsu, Masutake; Hazama, Taira
Journal of Energy and Power Engineering, 13(11), p.393 - 403, 2019/11
This paper describes methods and results of the uncertainty evaluation of the significant plant response analysis of the reactor trip failure event, i.e. anticipated transient without scram of the Japanese prototype fast breeder reactor Monju. Unprotected loss of heat sink has a relatively large contribution to the core damage frequency due to reactor trip failure. The uncertainty of the allowable time to core damage in this event by plant transient response analysis, so far, has been estimated with considering the range of reactivity coefficients. There are some cases where core damage is considered to be avoided. Specifically, it is assumed that the core damage due to the ULOHS event would be avoided if the sodium temperature at the pump inlet stays below 650
C for 1 h; otherwise the possibility of cavitation occurring in the hydrostatic bearing increases. In this study, a method is developed to search for a solution as an inverse problem of multiple input variables that satisfy the temperature condition. This paper, as a first step, describes input conditions and probability to satisfy the temperature are evaluated through analyses treating input parameters, reactivity coefficients and kinetic parameters, as variables within the design range.
Muto, Keitaro; Hamano, Daisuke; Kawabata, Mamoru; Tabakoya, Yasuhiro; Yanai, Chisato
E-Journal of Advanced Maintenance (Internet), 11(2), p.86 - 91, 2019/09
Gas sampling type sodium small leak detector, SID, Sodium Ionization Detector, a gas sampling detector installed to monitor small sodium leak in the Prototype Fast Breeder Reactor Monju, had been operated since it's system start up test through to the 40% reactors output test, It's use was terminated in April, 2018. SID showed some indication variations during its operation period, and necessary measures were implemented. As a result, the SID system maintained its functions without any critical malfunction until the end of its operation.
Tsuruga Comprehensive Research and Development Center
JAEA-Technology 2019-007, 159 Pages, 2019/07
JAEA-Technology-2019-007.pdf:19.09MBJAEA-Technology-2019-007-high-resolution1.pdf:42.36MBJAEA-Technology-2019-007-high-resolution2.pdf:33.56MBJAEA-Technology-2019-007-high-resolution3.pdf:38.14MBJAEA-Technology-2019-007-high-resolution4.pdf:48.82MBJAEA-Technology-2019-007-high-resolution5.pdf:37.61MB
This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.
Aizawa, Kosuke; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*
Nuclear Technology, 204(1), p.74 - 82, 2018/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Development of inspection technique in opaque liquid metal coolant is one of the important issues to ensure the safety of Liquid Metal Fast Breeder Reactor (LMFBR). Performance tests of an Under Sodium Viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of LMFBR Monju, have been carried out. The ultrasonic sensors and reflectors are located across the core inside of the Monju's RV. The USV can detect an obstacle existing in between the core top and the Upper Core Structure (UCS) bottom by identifying differences of echo signals. This report describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated and the signal to noise ratio successfully exceeded the target value. Measured signals clearly differed from with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in the specified place.
Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka
JAEA-Technology 2018-003, 97 Pages, 2018/07
JAEA-Technology-2018-003.pdf:12.54MB
The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.
Kitano, Akihiro; Nakajima, Ken*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04
The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.
Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin
Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04
Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the 
decay of 
Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month (
10
fissions/cm
). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.
Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin
Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11
Times Cited Count：2 Percentile：19.65(Nuclear Science & Technology)The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.
Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.
Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*
JAEA-Research 2017-001, 40 Pages, 2017/03
JAEA-Research-2017-001.pdf:5.19MB
In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (
2.04C) required by the JEAC4203-2008.
Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*
JAEA-Research 2016-021, 32 Pages, 2017/02
JAEA-Research-2016-021.pdf:5.0MB
In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.
Kitano, Akihiro
Bunri Henkan Gijutsu Soron, p.215 - 222, 2016/09
Prototype fast breeder reactor Monju has restarted in May 2010 after the long shutdown since the sodium leak accident. Before the restart the safety authority has examined the influence of the long shutdown, such as the characteristics of Am accumulated in MOX fuel. The integral data of the core with about 1.5% was acquired in the core confirmation test conducted in 2010 and the newest analysis method shows the enough accuracy. The irradiation test for the MA transmutation is planned, called GACID project by Japan, France and United States. The research of MA transmutation is one of the most important issues in Monju research plan included in the basic research plan decided by the cabinet council. It is expected that Monju would be utilized to demonstrate MA transmutation in the fast reactor.
Kitano, Akihiro; Takegoshi, Atsushi*; Hazama, Taira
Journal of Nuclear Science and Technology, 53(7), p.992 - 1008, 2016/07
Times Cited Count：8 Percentile：60.26(Nuclear Science & Technology)A feedback reactivity measurement technique was developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K
) and reactor vessel inlet temperature (K
). This technique was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties revealed that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The C/E value of K showed good agreement between calculated and measured values within the established uncertainty, and the value of K was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2C.
