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Journal Articles

The R&D goal of Monju

Hiroi, Hiroshi*; Arai, Masanobu; Kisohara, Naoyuki

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06

The purpose of fast breeder reactors (FBR) and the role of Monju were discussed in Ministry of education, culture, sports science and technology-Japan (MEXT) after the Fukushima NNP accident. The discussion has concluded that FBRs contribute to energy security and reduction of high-level radioactive waste, and that Monju is to be utilized to demonstrate these usefulness and to implement international contributions. This paper addresses anticipated R&D results from Monju on the basis of the enforcement of new nuclear regulation, the energy situations in Japan and the international status of FBR development and collaborations.

JAEA Reports

Preliminary conceptual design of the secondary sodium circuit-eliminated JSFR (Japan Sodium Fast Reactor) adopting a supercritical CO$$_{2}$$ turbine system, 2; Turbine system and plant size

Kisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji*

JAEA-Research 2014-016, 60 Pages, 2014/09

JAEA-Research-2014-016.pdf:22.38MB

JAEA has performed a design study of an S-CO$$_{2}$$ gas turbine system applied to the JSFR. In this study, the S-CO$$_{2}$$ cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. This report describes the system configuration, heat/mass balance, and main components of the S-CO$$_{2}$$ turbine system, based on the JSFR specifications. The layout of components and piping in the reactor and turbine buildings were examined and the dimensions of the buildings were estimated. The study has revealed that the reactor and turbine buildings could be reduced by 7% and 40%, respectively, in comparison with those in the existing JSFR design with the secondary sodium circuit employing the steam turbine. The cycle thermal was also calculated as 41.9-42.3%, which is nearly the same as that of the JSFR with the water/steam system.

JAEA Reports

Preliminary conceptual design of the secondary sodium circuit-eliminated JSFR (Japan Sodium Fast Reactor) adopting a supercritical CO$$_{2}$$ turbine system, 1; Sodium/CO$$_{2}$$ heat exchanger

Kisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji*

JAEA-Research 2014-015, 33 Pages, 2014/09

JAEA-Research-2014-015.pdf:27.33MB

JAEA has performed a design study of an S-CO$$_{2}$$ gas turbine system applied to the JSFR. In this study, the S-CO$$_{2}$$ cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. The Na/CO$$_{2}$$ heat exchanger is one of the key components, and this report describes its structure and the safety in case of CO$$_{2}$$ leak. A Printed Circuit Heat Exchanger (PCHE) is employed to the heat exchanger. A SiC/SiC ceramic composite material is used for the PCHE to prevent crack growth and to reduce thermal stress. The Na/CO$$_{2}$$ heat exchanger has been designed in such a way that a number of small heat transfer modules are combined in the vessel in consideration of manufacture and repair. CO$$_{2}$$ leak events in the heat exchanger have been also evaluated, and it revealed that no significant effect has arisen on the core or the primary sodium boundary.

Journal Articles

Evaluation of Earthquake and Tsunami on JSFR

Chikazawa, Yoshitaka; Enuma, Yasuhiro; Kisohara, Naoyuki; Yamano, Hidemasa; Kubo, Shigenobu; Hayafune, Hiroki; Sagawa, Hiroshi*; Okamura, Shigeki*; Shimakawa, Yoshio*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.677 - 686, 2012/06

Evaluation of Earthquake and Tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong Earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.

Journal Articles

Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

Kisohara, Naoyuki; Suzuki, Hiroshi; Akita, Koichi; Kasahara, Naoto*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.621 - 630, 2012/06

Journal Articles

Conceptual design study for the demonstration reactor of JSFR, 5; Reactor cooling system design

Kisohara, Naoyuki; Ishikawa, Hiroyasu; Futagami, Satoshi; Xu, Y.*; Shimoji, Kuniyuki*; Kawamura, Masaya*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

The cooling system of the JSFR adopts an integrated primary sodium pump/intermediate heat exchanger (IHX), dual structure straight tube steam generator (SG) and short elbow sodium piping layout. Since, however, this is the first experience applying these technologies to SFRs in Japan, design approaches have been evaluated and R&D has been undertaken. This paper addresses the design study of the cooling system of the demonstration reactor JSFR in terms of thermal-hydraulic and structural integrity. Recent studies have shown that these new technologies have potential to be applied to the JSFR.

Journal Articles

Steam water pressure drop under 15 MPa

Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Journal of Power and Energy Systems (Internet), 5(3), p.229 - 240, 2011/04

For a steam generator with straight double-walled heat transfer tubes that used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multiplier models were checked and then, it was found that both two-phase flow multiplier models of Chisholm and homogeneous can predict the present experimental data in high accuracy.

JAEA Reports

Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

Kisohara, Naoyuki; Sakamoto, Yoshihiko

JAEA-Review 2010-036, 26 Pages, 2010/09

JAEA-Review-2010-036.pdf:3.05MB

The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted. Owing to these work methods, the modification work proceeded close to schedule without incident.

Journal Articles

Thermal-hydraulic experiments under high pressure condition

Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05

For steam generator with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important issues need researching. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube with a similar inner diameter as that in the designed SG. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15 MPa. Six models for the prediction of two-phase multiplier were evaluated. The results showed the Chisholm correlation and homogeneous model gave best predictions. Note that in the homogeneous model verification, the homogeneous model was only used in the friction loss calculation. In the calculation of void fraction, which is necessary for static head, drift flux model, instead of homogeneous model, was used.

Journal Articles

Steam-water pressure drop under high pressure condition

Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 10 Pages, 2009/09

For the Steam Generator (SG) in a commercialized sodium cooled faster breeder reactor, flow instability in water side is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments using water as test fluid were performed under high pressure condition at JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper focuses on the discussion to steam - water pressure drop. We evaluated existing correlations for two-phase flow multiplier under high pressure. As a result, Chismholm correlation was confirmed being the best one for the present high pressure data.

Journal Articles

Flow instability research on steam generator with straight double-walled heat transfer tube for FBR; Pressure drop under high pressure condition

Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 5 Pages, 2008/11

To discuss the feasibility of Steam Generator (SG) with a straight double-walled heat transfer tube that used in the Fast Breeder Reactor (FBR) system, we need to construct thermal hydraulic design method that can predict the flow instability accurately. To verify and to improve the correlations that used in the thermal-hydraulic design of the SG, Japan Atomic Energy Agency has started experiments under high pressure conditions. Detailed thermal hydraulic data including pressure drop data have been derived. This research does the analysis to the performed experiments with using TRAC-BF1 code. The pressure drop under high pressure condition is verified. It is found that with using the drift flux model in Track code for the void fraction calculation, Pffan's correlation for the friction pressure drop calculation in single phase flow and Martinelli-Nelson two-phase multiplier, the pressure drop can be predicted conservatively.

Journal Articles

Temperature and flow distributions in sodium-heated large straight tube steam generator by numerical methods

Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki

Nuclear Technology, 164(1), p.103 - 118, 2008/10

 Times Cited Count:3 Percentile:73.51(Nuclear Science & Technology)

A sodium heated steam generator (SG) for the Japanese future commercialized fast reactor is a straight double-wall-tube type. The SG is large-sized by economics of scale. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These mal-distributions cause tubes thermal expansion mismatch and it might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs. The temperature profiles in the SG are examined by numerical methods, and the flow distribution devices are designed to prevent these issues. Multi-dimensional thermal-hydraulic codes "FLUENT" and "MSG" are used to predict tube temperature distributions, and the thermal loads on tubes are obtained by the structural code "FINAS". These codes have revealed that the sodium flow is distributed uniformly by the flow distributors, and that the tube thermal loads remain within the allowable range for the tubes and the junctions structural integrity.

JAEA Reports

Studies of super-critical CO$$_{2}$$ gas turbine power generation fast reactor(Contract research, Translated document)

Kisohara, Naoyuki; Kotake, Shoji; Sakamoto, Yoshihiko

JAEA-Review 2008-040, 67 Pages, 2008/08

JAEA-Review-2008-040.pdf:3.93MB

(1) Preliminary design of an SFR that adopts a super-critical CO$$_{2}$$ turbine has been made. This SFR system eliminates secondary sodium circuits. The power generation efficiency of the SFR is 42%. Compared to a SFR that adopts a steam cycle with secondary sodium circuits, the reactor building volume of the SC-CO$$_{2}$$ SFR is reduced by 20%. (2) A super-critical CO$$_{2}$$ cycle test loop was fabricated. The high efficiency of a compressor is confirmed near the super-critical region. The temperature efficiencies of PCHE recuperators are 98-99% (hot leg). No flow instability is observed in the loop operation. (3) Na/CO$$_{2}$$ Reaction tests were executed. Continuous reaction occurs more than 570-580$$^{circ}$$C. Reaction products are Na$$_{2}$$CO$$_{3}$$ and CO. The reaction heat is 50-75kJ/Na-mol. (4) Safety calculation was done for 1 DEG tube failure in Na/CO$$_{2}$$ heat exchanger. The maximum pressure in the primary circuit is below the allowed level. The void reactivity of the reactor core also has no affect. The reaction product brought no sodium flow blockage in fuel assemblies. (5) After an exposure of 600$$^{circ}$$C-5000 hours in super-critical CO$$_{2}$$ environment, the corrosion of 12Cr steel and 316FR were 170g/m$$^{2}$$ and 5g/m$$^{2}$$, respectively. 316 FR shows a good corrosion proof property.

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

Flow and temperature distribution evaluation on sodium heated large-sized straight double-wall-tube steam generator

Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki

Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 9 Pages, 2006/06

The sodium heated steam generator (SG) of the commercialized FBR being designed in the Feasibility Study is a straight double-wall-tube type and it is large-sized to reduce the manufacturing cost by economics of scale. This paper addresses the temperature and flow multi-dimensional distributions at steady state to obtain the prospect of the SG. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These phenomena might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs. The flow adjustment devices installed in the SG are optimized to prevent these issues, and the temperature distribution properties are uncovered by analysis methods. The analysis model of the SG consists of two parts, a sodium inlet distribution plenum (the plenum) and a heat transfer tubes bundle region (the bundle). The flow and temperature distributions in the plenum and the bundle are evaluated by the three-dimensional flow code "FLUENT" and the two dimensional thermal-hydraulic code "MSG". The MSG code is particularly developed for sodium heated SGs in JAEA. These codes have revealed that the sodium flow is distributed uniformly by the flow adjustment devices, and that the lateral tube temperature distributions remain within the allowable temperature range for the structural integrity of the tubes and the tube to tube-sheet junction.

JAEA Reports

Design study on sodium-cooled reactor; Results of the studies in 2004 (Joint research)

Hishida, Masahiko; Murakami, Tsutomu*; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato*; Hayafune, Hiroki; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu; Uno, Osamu; et al.

JAEA-Research 2006-006, 125 Pages, 2006/03

JAEA-Research-2006-006.pdf:11.55MB

In Phase I of the "Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase II, design improvement for further cost reduction and the establishment of the plant concept has been performed. In this study, reactor core design and large-scale plant design have been performed by adopting the modified fuel assembly with inner duct structure and double-wall straight tube steam generator (SG), which concepts were chosen at the interim review of FY 2003. For this SG, safety logics have been studied and the structural concept has been established. And the plant designs improving the in-service inspection (ISI) and repair capability have been performed. Furthermore, elaborate confirmation of the design has been performed reflecting the development of elemental technology, back-up concepts have been proposed. Besides, cost reduction measures have been studied by reducing reactor grade materials, introducing autonomous standardizations, simplifying the design due to deregulation and adopting systemized standards for BOP and NSSS. From now on, reflecting the results of elemental experiments, in-depth design studies and examination of critical issues will be carried out and the plant concept will accomplish in preparation for the final evaluation in Phase II.

Journal Articles

Feasibility study of a compact loop type fast reactor without refueling for a remote place power source

Chikazawa, Yoshitaka; Kisohara, Naoyuki; Usui, Shinichi; Konomura, Mamoru; Sawa, Naoki*; Sato, Mitsuru*; Tanaka, Toshihiko

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

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 MW-electric 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.

JAEA Reports

Conceptual design study of small sodium cooled reactors

Chikazawa, Yoshitaka; Kisohara, Naoyuki; Usui, Shinichi; Konomura, Mamoru; Tanaka, Toshihiko

JNC-TY9400 2005-004, 189 Pages, 2005/06

JNC-TY9400-2005-004.pdf:9.33MB

A conceptual design of various small metal fuel sodium cooled reactors has been studied in the feasibility study on commercialized fast breeder reactor cycle system. In FY2004 study, a 50 MWe power plant for remote places with a long life core without refueling and a 300 MWe modular reactor which pursues standardization for learning effect and reduction of capital risks.In the small reactor with a long life core, the reactor vessel is minimized without a permanent fuel handling system and the cooling system is simplified adopting 1 loop. The total mass of the reactor vessel and the cooling system is dramatically reduced and the concept has a potential to be an attractive power source for remote places.In the 300 MWe modular reactor, the cooling system adopts 1 loop and the ex-vessel fuel storage tank for spent fuels is eliminated adapting the in-vessel storage (IVS) which has a capacity for a 4 year storage. The reactor building is minimized without the ex-vessel storage This concept has a potential to be an power source for key grids with modular constructions and a first plant with a small fuel cycle facility can demonstrate the metal fuel fast reactor cycle

JAEA Reports

Designing Study on large-sized Steam Generators of Sodium cooled FBR; Measures to diminish the influence of water leak

Kisohara, Naoyuki; Hori, Toru; Soman, Yoshindo;

JNC-TN9400 2004-062, 69 Pages, 2004/10

JNC-TN9400-2004-062.pdf:4.73MB

In case of heat transfer tubes failure in sodium heated Steam Generators (SG), the secondary sodium pressure rapidly increases due to the hydrogen generated by Na/water reaction. The Na/water reaction mitigation system terminates this phenomena immediately without any influence on the integrity of the sodium circuit. Therefore, the water leak has no effect on the reactor and plant safety. However, not only safety but also economy and public acceptance are required for practical FBRs. As for economy, it is necessary to protect the investment of SGs and to avoid the decline of plant availability due to water leak. In addition, decreasing the possibility of Na/water reaction accidents is also taken into consideration in order to promote the public acceptance of FBRs. For these purposes, a double wall straight tubes (DWT)-SG and a single wall helical coil tubes SG are selected as the candidates for the FBR's SG. The DWT-SG has the potential to exclude Na/water reaction by its dual boundaries between Na and water. Although the helical coi1 SG provides single wall tubes, this SG have proven to be developed in Japan and the high reliability can be attained by a lot of knowledge and experience of the SG test facility and the prototype reactor. In terms of the avoidance of Na/water reaction, the DWT-SG is the first candidate. However, there are many issues to be solved for the DWT-SG, the single wall helical coil tubes SG is regarded as the second candidate for an alternative. This report describes the method to prevent or minimize Na/water reaction for both of the SGs. The ultra sonic test (UT) method during the periodical plant inspection is applied to DWT-SG to prevent inner and outer tube simultaneous failure. The preliminary ISI test and failure analysis of the DWT indicate the potential of avoiding the penetrated failure of DWT. However, crack detection tests by UT and crack development analysis due to DNB are indispensable to confirm this methodology to exclude ...

JAEA Reports

Design Study on Sodium-Cooled Large-Scale Reactor

Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki; Hayafune, Hiroki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Uno, Osamu; Saigusa, Toshiie; et al.

JNC-TY9400 2004-014, 78 Pages, 2004/07

JNC-TY9400-2004-014.pdf:7.97MB

This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared.

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