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Journal Articles

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

 Times Cited Count:13 Percentile:77.66(Nuclear Science & Technology)

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.

Journal Articles

Evaluation on coolability of the reactor core in Monju by natural circulation under earthquake and subsequent tsunami event

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

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

The core cooling capability by natural circulations at a station black-out event, induced by an earthquake and a subsequent tsunami attack, has been evaluated in detail, referring to the accident of the Fukushima Dai-ichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code: Super-COPD has been used for the evaluation, which has been verified by the analyses of the preliminary test results on the natural circulation in Monju. As a result it was concluded that the natural circulations of the coolant sodium will enable the decay heat removal of the core as far as the sodium coolant flow circuits are intact and secured.

Journal Articles

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

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.

Journal Articles

MONJU; A Center of GACID project

Konomura, Mamoru

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 54(5), p.312 - 314, 2012/05

The GACID project has been performed by collaboration with Japan, France and the United States of America. The objectives of GACID project is to irradiate at least a fuel subassembly with Minor Actinides bearing fuel in Monju which is a Japanese prototype fast breeder reactor. Monju will carry out three functions at the same time: a decrease of Minor Actinides, the generation of electricity and a breed of nuclear fuel. That is the reason why a re-start of Monju will be expected internationally.

Journal Articles

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

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.

Journal Articles

Phenix to Monju; FBR development and Japan role, 1; Incident in development of FBR and its effect on the development

Konomura, Mamoru

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 53(3), p.211 - 214, 2011/03

It is not too much to say that R&D people have common sense to manage incidents in development of FBR with a method of risk analysis. In order to avoid an accident from some incidents, a reactor plant is designed with a way of defense in depth and a concept of multiple function and diversity in the design phase and is constructed and operated with enhancement of its reliability. Actually it is impossible to avoid any mechanical incidents. But the reliability can be enhanced by these incidents with appropriate risk management. In this sense, incidents are valuable for development of FBR. We should pay attention to developer's attitude how he understand incidents and how he improve his system on condition that he keeps transparency of information to public.

Journal Articles

Design challenges for sodium cooled fast reactors

Konomura, Mamoru; Ichimiya, Masakazu

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

 Times Cited Count:17 Percentile:76.46(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

A Modular metal-fuel fast reactor with one-loop main cooling system

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sato, Koji; Sawa, Naoki*; Sumita, Hiroyuki*; Nakanishi, Shigeyuki*; Ando, Masato*

Nuclear Technology, 159(3), p.267 - 278, 2007/09

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

A diversified or modular power source is attractive since it requires a low construction cost per unit and can be demonstrated in small scale experimental facilities. In this study, a new metal fuel sodium cooled reactor with 300MW electric has been developed enhancing cost reduction. And economical potential at demonstration stage with first of a kind (FOAK) is emphasized. A minimum configuration with a compact reactor vessel, a one-loop main cooling system and a simple fuel handling system is adopted enhancing cost reduction. For safety evaluation, reliability of the one-loop main cooling system has been shown by pipe-break transient analyses. Besides, construction cost of a demonstration plant with a first reactor and a small reprocessing and fuel fabrication facility is also evaluated. A major feature of the present concept is that the demonstration reactor and facilities can be directly appropriated for first commercial modules and the power plant can easily increase its capacity adding reactor and electrorefiner modules. A fast reactor cycle commercialization scenario using the present concept is thought to give low R&D or investment risk and high cost performance since the total demonstration plant cost is relatively small and the facilities are directly appropriated to commercial use.

JAEA Reports

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.

Journal Articles

Progress on the plant design concept of sodium-cooled fast reactor

Hishida, Masahiko*; Kubo, Shigenobu; Konomura, Mamoru; Toda, Mikio*

Journal of Nuclear Science and Technology, 44(3), p.303 - 308, 2007/03

 Times Cited Count:16 Percentile:74.96(Nuclear Science & Technology)

An innovative concept of sodium-cooled fast reactor, named JNC Sodium Cooled FR (JSFR) has been created and modified through the Feasibility Study on Commercialized FR Cycle System, aiming at full satisfaction of the development targets. A modified concept of JSFR applied double-wall straight tube type steam generator (SG) which is excelling in safety for sodium-water reaction has been developed. In addition, decay heat removal system suitable for the straight tube SG has been selected and in-service inspection and repair capabilities have been improved. As the result of this study, the potential to realize this plant concept has been obtained through evaluation concerning safety and economics.

JAEA Reports

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.

Journal Articles

A Compact loop-type fast reactor without refueling for a remote area power source

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sawa, Naoki*; Shimakawa, Yoshio*; Tanaka, Toshihiko*

Nuclear Technology, 157(2), p.120 - 131, 2007/02

 Times Cited Count:1 Percentile:11.92(Nuclear Science & Technology)

A small reactor has a potential to be utilized as a power source applicable to diversified social needs and reduce capital risks. In remote sites where the population is small and plants can not be economically connected to a power grid, power sources without refueling whose capacities are lower than 50 MWe are required because fuel transfer cost is expensive in such sites. In the present study, a small sodium cooled core with 30 years lifetime has been developed and a simple plant system without refueling has been sketched. Dimensions of major components are determined to evaluate its economical potential. Transient analyses show that self actuated shutdown system (SASS) enhances the passive safety features to maintain the reactor integrity in anticipate transient without scram events.

Journal Articles

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

 Times Cited Count:5 Percentile:38.34(Nuclear Science & Technology)

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 500$$^{circ}$$C 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.

JAEA Reports

Study of an electromagnetic pump in a sodium cooled reactor; Design study of secondary sodium main pumps (Joint research)

Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru*; Uchida, Akihito*; Nishiguchi, Yohei*; Nibe, Nobuaki*

JAEA-Research 2006-049, 75 Pages, 2006/07

JAEA-Research-2006-049.pdf:4.55MB

In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160m$$^3$$/min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160m$$^3$$/min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160m$$^3$$/min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R&D plan of EMP are reported.

Journal Articles

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$$_{2}$$ 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.

Journal Articles

A Modular metal fuel fast reactor enhancing economic potential

Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sato, Koji; Ando, Masato*; Nakanishi, Shigeyuki*; Sawa, Naoki*; Shimakawa, Yoshio*

Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 8 Pages, 2006/06

A diversified or modular power source is attractive since it requires a low construction cost per unit and can be demonstrated in small scale experimental facilities. In this study, a new metal fuel sodium cooled reactor with 300MW electric has been developed enhancing cost reduction. And economical potential at demonstration stage with first of a kind (FOAK) is emphasized. A minimum configuration with a compact reactor vessel, a one-loop main cooling system and a simple fuel handling system are adopted enhancing cost reduction within safety requirement. Besides, construction cost of a demonstration plant with a first kind of reactor and a small reprocessing and fuel manufacturing facility is also evaluated. A major feature of the present concept is that the demonstration facilities can be appropriated for commercialized ones since they can be easily commercialize by increasing reactor and electrorefiner modules. A FBR cycle commercialization scenario using the present concept is thought to give low risk and high cost performance since the total demonstration plant cost is relatively small and the facilities are directly appropriated to commercial use.

Journal Articles

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

 Times Cited Count:1 Percentile:10.64(Nuclear Science & Technology)

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.

JAEA Reports

Design study on a fuel handling system in a sodium cooled reactor; Study in FY2004 (Joint research)

Chikazawa, Yoshitaka; Usui, Shinichi; Konomura, Mamoru; Ikeda, Hirotsugu

JAEA-Research 2006-032, 202 Pages, 2006/04

JAEA-Research-2006-032.pdf:38.12MB

In the feasibility study on commercialized fast breeder cycle system, fuel handling systems for sodium cooled reactors has been studied. In FY 2004 study, a fuel handling system with an EVST for a twin large scale reactor power plant is designed and key issues about the system are identified. A manipulator type fuel handling machine suitable for the upper internal structure with a slit designed and seismic analyses show that it can treat spent fuels without interaction with upper internal structure in earthquakes. Fuel handling time is reduced adopting a sodium pot which can carry 2 subassemblies in onetime. Spent fuels are stored at an EVST while their decay heat are reduced to be 5kW/subassembly. A new fuel handling system for fuels with minor actinide is designed considering 1kW/subassembly heat and shielding. A innovative concept without an EVST is also studied. A fuel handling system adopting fuel transfer without a sodium pot is constructed to reduce material mass. A fuel handling system for a metal fuel reactor plant has been design. From the result of a survey on a gas storage, a water pool storage with helium cans and EVST, a system with EVST is selected because of its economical and safety advantage. Fuel handling condition is briefly reviewed considering commercialized reactor fuel specifications such as minor actinide content and ODS cladding.

JAEA Reports

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.

JAEA Reports

Conceptual design study of Cu bonded steam generator; Pressurized crack propagation experiments at the room temperature

Chikazawa, Yoshitaka; Aizawa, Kosuke; Konomura, Mamoru

JAEA-Research 2006-007, 114 Pages, 2006/03

JAEA-Research-2006-007.pdf:15.6MB

In the feasibility study of commercialized fast reactor cycle systems of JAEA, we make a concept of a sodium cooled reactor without secondary sodium circuits. And a sodium cooled reactor with Cu bonded steam generator is one of promising concept has been investigated. In the FY2004 study, pressurized tube experiments using the 3$$times$$3 array specimen imitated the reference tube geometry are carried out aiming to clarify crack propagation behavior in the reference steam generator. Compared with the results of the bend experiments in the FY2003, it is understood that the crack initiation and the crack propagation are greatly obstructed by the effect of the specimen shape. The reference tube geometry was optimized to pursue the economic. As a result of optimization, the construction cost of reactor cooling system with Cu bonded steam generators is 0.7 times as much as that of an ordinary sodium cooled reactor with secondary sodium circuits. Condition for crack propagation in the reference tube geometry is analyzed by using the result of the bend experiments in the FY2003. From the results of the bend experiments and the analysis, it was understood that there was a possibility that the crack reaches the Cu layer. But, when the crack reaches the Cu layer, it was shown that the crack propagation stopped.

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