Development of natural circulation analytical model in Super-COPD code and evaluation of core cooling capability in Monju during a station blackout

Yamada, Fumiaki; Fukano, Yoshitaka; Nishi, Hiroshi; Konomura, Mamoru

Nuclear Technology, 188(3), p.292 - 321, 2014/12

https://doi.org/10.13182/NT13-56

The capability of natural circulation for core cooling has been evaluated in detail for a station blackout (SBO) event induced by an earthquake and a subsequent tsunami hit. The evaluation was prompted by the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code Super-COPD was used for the evaluation, which has been validated by analyses of preliminary test results on the natural circulation in Monju. As a result, it was concluded that natural circulation of the sodium coolant will enable the decay heat from the core to be removed under such an SBO condition.

Recent progress and status of Monju

Kondo, Satoru; Deshimaru, Takehide; Konomura, Mamoru

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 6 Pages, 2013/03

http://www.iaea.org/nuclearenergy/nuclearpower/Downloadable/Meetings/2013/2013-03-04-03-07-CF-NPTD/T9.1/T9.1.deshimaru.pdf

Monju, a 280-MWe prototype sodium-cooled fast reactor of, restarted its test operation in 2010. The zero-power tests were successfully conducted. It's major achievement was accurate prediction of reactor physics parameters with a core including americium-rich fuel. The reactor, however, has been put into a stand-by mode again since the 3.11 Fukushima-Daiichi accident. The roles shall not change: demonstrating stable power generation and actinide burning; providing technology and knowledge base for future SFRs; and using the plant as an international research facility.

Future R&D programs using Monju

Konomura, Mamoru; Ichimiya, Masakazu; Mukai, Kazuo

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

Japanese prototype fast breeder reactor, Monju, will soon restart. Monju has three sodium loops with steam generators and a turbin-generator, 280 MWe. Monju is expected to demonstrate the sodium technology and the sodium-water heat exchange technology in Japan. It will be operated at full power operation during around 10 years after restart in order to accumulate operation & maintenance experience and to evaluate its design technology. After the system start-up test (SST), Monju will be operated under full power. In this stage, the main object of Monju operation will be to achieve its initial targets which were fixed when its construction was decided around thirty years ago. The targets are to demonstrate a safe and reliable operation, that is, accumulation of the operation & maintenance experience and evaluation of the design technology, and to establish sodium handling technology. For example, inspection and diagnosis technologies are important for maintenance of a sodium cooled reactor. An out-of-pile sodium test facility will be constructed near Monju in order to make many tests of inspection devices and research many chemical tests. At the same time, the activities for the performance improvement, for example, a new licence, will be prepared in order to utilize Monju as a R&D facility. After the accumulation of operating experience, Monju will be enhanced the performance as a R&D facility in order to demonstrate innovative technologies, for example, irradiation of advanced fuel, longer operation cycle, higher burnup. For this purpose, Monju will be needed to get a new licence and core modification. And Monju on-site non-destructive Post Irradiated Evaluation facility will be expected at this stage. There were many R&D works in Japan with sodium out-of-pile facilities. All the experience were reflected in the design of Monju. Monju will demonstrate a handling of sodium technologies under power plant operation.

Design challenges for sodium cooled fast reactors

Konomura, Mamoru; Ichimiya, Masakazu

Journal of Nuclear Materials, 371(1-3), p.250 - 269, 2007/09

https://doi.org/10.1016/j.jnucmat.2007.05.012

no abstracts in English

Conceptual design study of small sodium cooled reactors; 300MWe modular reactor (Joint research)

Aizawa, Kosuke; Chikazawa, Yoshitaka; Usui, Shinichi; Konomura, Mamoru; Ando, Masato*

JAEA-Research 2007-042, 105 Pages, 2007/06

JAEA-Research-2007-042.pdf:3.64MB

Various conceptual design studies of sodium cooled small reactor have been performed in the feasibility study. In FY2005 study, a 300MWe modular reactor which adopts metal-fueled and 1 loop cooling system in order to pursues economical competitiveness was investigated. Transient analyses were performed to show core safety under the large pipe break accident that was found to be the severest accident in case of a 1 loop cooling system. From the analysis results, it was showed a possibility to maintain the core safety without the reactor scram. Transient analyses of the natural circulation decay heat removal system were also performed, and it was showed a possibility to maintain the core safety after the reactor scram. The in-vessel storage (IVS) was adopt to eliminate the ex-vessel storage and the design of the distribution flow adjustment device in IVS is studied. It was showed that the IVS could be introduced without modifying in the reactor design in FY2004. The rationalization of the fuel handling system is investigated taking advantage of co-location plant in which the reactor, the recycle plant and the fuel fabrication plant are co-located in a site. As a result of rationalization, the volume of the reactor building becomes 0.85 times as much as that of the design in FY2004.

Development of a new fuel handling machine suitable for an upper inner structure with a slit; The Performance test of a large-sized bearing in Argon gas atmosphere

Chikazawa, Yoshitaka; Usui, Shinichi; Hayafune, Hiroki; Konomura, Mamoru

JAEA-Research 2007-001, 91 Pages, 2007/02

JAEA-Research-2007-001.pdf:9.42MB

In Feasibility Study on Commercialized Fast Breeder Reactor Cycle Systems, Large and Middle scale sodium cooled reactors which have an upper inner structure (UIS) with a slit have been studied for the size reduction of reactor structure. A new fuel handling machine (FHM) which is suitable for the UIS with a slit has been developed in this study. The FHM is required not to contact to UIS under the earthquake condition when its arm is extended into the slit. In the previous study, it is confirmed that the reduction of clearance in bearing of FHM is effective to reduce the deflection of FHM unit. But the general lubricant such as grease can not be used for the bearing because that of FHM will be set in argon gas atmosphere of the reactor vessel. In FY2005, the performance test of large-sized bearing with the selected lubrication specification has been performed in the high temperature argon gas atmosphere.

A System design study of a fast breeder reactor hydrogen production plant using thermochemical and electrolytic hybrid process

Chikazawa, Yoshitaka; Nakagiri, Toshio; Konomura, Mamoru; Uchida, Shoji*; Tsuchiyama, Yoshihiko*

Nuclear Technology, 155(3), p.340 - 349, 2006/09

https://doi.org/10.13182/NT06-A3766

Hydrogen production with a fast breeder reactor is attractive as a long term energy source with nuclear fuel breeding. The thermochemical and electrolytic hybrid process is one of the hydrogen production methods using sulfuric acid cycle with the maximum temperature 500C which can be supplied by a sodium cooled fast breeder reactor. In this study, a hydrogen production plant with the thermochemical and electrolytic hybrid process has been designed and the hydrogen production efficiency has been evaluated. In the present concept, components in hydrogen production system are designed to be made of steels such as high Si cast iron which has good toughness against sulfuric acid. High hydrogen production efficiency of 44% (high heating value) is achieved assuming development of high efficiency electrolysis.

Development of a slim manipulator type fuel handling machine for a commercialized fast reactor

Chikazawa, Yoshitaka; Usui, Shinichi; Konomura, Mamoru; Sadahiro, Daisuke*; Tozawa, Katsuhiro*; Hori, Toru*; Toda, Mikio*; Kotake, Shoji*

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07

A seismic analysis has been performed showing that the seismic interaction between the UIS and the FHM can be avoided adopting gapless bearings at the FHM arm joint. An angular contact ball bearing is suitable for the new FHM since it can eliminate gaps by preload pressure. A major problem of the FHM bearings is lubrication since the contact pressure between steel rings and ball increases because of ball bearing. Additionally, FHM operating temperature is about 200 deg-C and normal grease is not applicable under argon gas with sodium vapor. A endurance test with 1/10 scale bearings in the air has been performed to show applicability of angular contact ball bearings to the FHM arm joint. The results with 20,000 cycle showed that bearings with combination of MoS coating steel rings and ceramics balls can be tolerable as the FHM operating condition. A real scale bearing test in argon gas with sodium vapor has also been demonstrated to reveal bearing size and sodium vapor effects.

A Conceptual design study of a small natural convection lead-bismuth cooled reactor without refueling for 30 years

Chikazawa, Yoshitaka; Konomura, Mamoru; Mizuno, Tomoyasu; Mito, Makoto*; Tanji, Mikio*

Nuclear Technology, 154(2), p.142 - 154, 2006/05

https://doi.org/10.13182/NT06-A3724

Small fast reactors have a potential to be utilized as a power source applicable to diversified social needs and reduce capital risks. In remote sites such as Alaska and Hawaii, power sources without refueling whose capacities are lower than 50 MW-electric are required, because fuel transfer cost is expensive in such sites. In the case of power sources for developing countries, capacities are required in a range 50-300 MW-electric and proliferation resistance must be enhanced. A solution of a power plant without refueling is a fast reactor plant with a very long life core and proliferation resistance in such a plant is supposed to be tough. We have studied concepts of small fast reactor with various attractiveness such as economical competitiveness, reactor safety and very long lived core. In this study, a design concept of a small lead-bismuth cooled reactor, which pursues long operating cycle and inherent safety has been studied. The core is nitride fueled two regions homogeneous core cooled by natural circulated lead-bismuth eutectic. The burnup reactivity is kept 0.93 %dk/k and without refueling 30 years operation with 7.8 GWd/t average burnup. Since lead-bismuth eutectic is chemically inert in lead-bismuth-water contact, a simple tank type design with the steam generator in the reactor vessel is adopted eliminating an intermediate circuit system. The transient analysis shows the reactor has possible capability to keep the reactor system and fuel integrity in UTOP and ULOHS events. The target construction cost for remote sites is achieved by the first plant with 50 MW-electric capacity. The economical target of power sources for developing countries cities is also achieved in the condition of 150 MW electric power regarding series effect.

Study of hydraulic behavior for reactor upper plenum in sodium-cooled fast reactor; Verification analysis of water experiment and applicability of vortex prediction method

Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*

JAEA-Research 2006-017, 113 Pages, 2006/03

JAEA-Research-2006-017.pdf:14.98MB

A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.