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
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.
Kisohara, Naoyuki; Hori, Toru; Soman, Yoshindo;
JNC TN9400 2004-062, 69 Pages, 2004/10
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 ...
Hayafune, Hiroki; Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu; Igawa, Kenichi*
JNC TN9400 2004-054, 339 Pages, 2004/08
Concepts of the reactor, SG and main coolant pump have been studied considering maintainability and aseismic capability, which is a medium size pool type lead-bismuth cooled reactor. The results are following.(1) Reconsideration of reactor design concepts concerning maintainability. In pursuit of good reactor maintainability, the structural concepts of SG, UIS and core support structures have been changed to be drawn up above the upper area of the reactor system. After a few decade of interval, lead-bismuth inventory in the reactor vessel shall be fully drained for easy ISI operation of in-vessel main components such as core support structures. From the viewpoint of the reactor aseismic capability, the axial length of reactor vessel was reduced and the reactor vessel support location was changed from the top hanging to the circumference of the vessel.(2) SG concept selection in conjunction with a compact reactor vessel.The concept of SG consisting of a once through type with helical coil tube is selected. 6 units of a small scale SG are arranged on a reactor roof deck along the peripheral direction, in addition to 3 units of a centrifugal mechanical pump.(3) Aseismic structural integrity of the reactor components. Aseismic structural integrity of the reactor vessel, core support structures, UIS, FHM, SG and the main pumps has been vigorously examined respectively. These components besides FHM could keep the aseismic structural integrity for strong S2 earthquake under the design condition FHM could also keep the integrity for S1 earthquake.(4) Safety evaluation. Thermal translents following loss of flow type accident due to plant total blackout and typical manual reactor trip incident, have been evaluated to assure the plant safety design, by analyzing thermal hydraulic behavior of transients concerning core flow rate and temperatures of the plant cooling system. *Loss of flow accident due to plant total blackout. The reactor coolant pumps shall be tripped and the
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
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.
Hishida, Masahiko; Murakami, Tsutomu; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato; Hayafune, Hiroki; Chikazawa, Yoshitaka; Hori, Toru; Saigusa, Toshiie; Uno, Osamu; et al.
JNC TY9400 2004-012, 97 Pages, 2004/07
Based on the concept of a plant consisting of four modules with a capacity of 750 MWe each, which has been established by the end of FY2002, a concept of the entire plant was proposed, reflecting the modifications related to the high internal conversion type core, the double-wall straight tube steam generator (SG), and the fuel storage system. Concept studies were also performed to overcome the drawbacks of the sodium and to achieve in-service inspection and repair as easily as in light water reactor. Furthermore, feasibility studies were carried out to confirm the design, which included safety, thermal-hydraulics and the structures of the primary reactor auxiliary cooling system and the double-wall straight tube SG. A prospect for realization of this plant concept has been obtained through the evaluation results. In addition, as the interim evaluation of the candidate concepts 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 medium-scale reactor candidate concepts were prepared.
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Kisohara, Naoyuki; Soman, Yoshindo; Nishiguchi, Yohei; Konomura, Mamoru
JNC TN9400 2003-090, 76 Pages, 2003/10
The conceptual design study of sodium-cooled FBRs is in progress in the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)". In this plant system design, a primary sodium pump are placed into an intermediate heat exchanger (IHX) and these two components are installed in one vessel to reduce plant construction cost. The primary sodium pump is located in the center of IHX tube bundle and this pump-IHX component includes primary/secondary sodium flow shrouds, pump-IHX boundary shroud and bellows. Although the pump and IHX are installed in the same vessel, these two components are structurally separated, because they are connected through a bellows and their weights are supported by different floors. However, the vibration caused by pump rotation has possibility to induce the vibration of heat transfer tubes via sodium and it leads the tubes to fretting wear against their support plates. Therefore, the tube fretting wear has been evaluated by both a simple beam vibration analysis model and a detailed shell vibration analysis model. Since the pump-IHX component consists of many different parts such as shrouds and tube bundle, analysis tools cannot reveal the vibration phenomena precisely. Then the analysis model requires to be validated by vibration tests. A 1/4-scale vibration test equipment of the whole pump-IHX component has been planned to confirm the occurrence and transmission of vibration. This test equipment is precisely minimized based on the actual pump-IHX component structure except the tube bundle. Acceleration sensors are installed on the shrouds to measure beam and shell vibration phenomena, and the signal will be used to reveal vibration modes and pump response properties. The fabrication of the test equipment was completed in FY2002 except the pump. In FY2003 the vibration tests are being executed by using electromagnetic vibrator instead of the pump. In FY2005, the pump will be settled into the test equipment, and the vibration tes
Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Saigusa, Toshiie; Uno, Osamu; Soman, Yoshindo; et al.
JNC TY9400 2003-015, 103 Pages, 2003/09
In Phase I of the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowingdown candidate concepts at the end of Phase 2.
Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki; Hori, Toru; Fujii, Tadashi; Uchita, Masato; Chikazawa, Yoshitaka; Saigusa, Toshiie; Uno, Osamu; Soman, Yoshindo; et al.
JNC TY9400 2003-014, 52 Pages, 2003/09
In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2002, which is the second year of Phase 2. In the JFY2002 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.As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects.
Sakai, Takaaki; Enuma, Yasuhiro; Soman, Yoshindo; Nishi, Yoshihisa*; Kinoshita, Izumi*
JNC TY9400 2003-012, 51 Pages, 2003/09
In System analysis has been performed to evaluate the thermal-hydraulics effect of the tube rapture accident. In addition, lifted flow rate by gas injection in the lead-bismuth has been measured to confirm the applicability of existing void fraction correlations based on the drift-flux model by Zuber-Findlay. As a result, it is clarified that the cooling capability is successfully maintained also in case of the tube rapture accident, because the flow reversal is limited to only 1/8 sector of downward flow area due to the tube support wall structure in the steam generator. And also, the applicability of existing void fraction correlations to Lead-Bismuth is confirmed by the gas injection experiment.
Enuma, Yasuhiro; Hayafune, Hiroki; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu; Sakai, Takaaki
JNC TN9400 2003-081, 234 Pages, 2003/09
Main issues of structural integrity, seismic integrity and operation are studied to confirm feasibilities of the medium size lead-bismuth cooled reactor as a candidate for one of the concepts of fast reactor in the feasibility studies on commercialized FBR cycle system. The results are following. (1) Integrity under loading of earthquake (a) Reactor system: Seismic integrity of the reactor is confirmed. (b) Upper core structures and fuel exchanging machine: Those structures are necessary for increasing stiffness by bonding. (2) Structural integrity for thermal loading Characteristics of both the medium size lead-bismuth cooled reactor and the large size sodium cooled reactor in the feasibility studies are arranged and structural integrity for thermal loading is evaluated about particular phenomena of thermal stratification and thermal striping. (3) Operation Feasibility of the natural circulation operation is confirmed. (a) Water-steam supply system concerned with control system (b) Control characteristics and transient characteristics of lead-bismuth reactor (c) Initial start-up operation concerned with preheating and charging of coolant (4) Safty Core integrity on SG tube rupture is evaluated to be confirmed its safety, by analyzing thermal hydraulic behavior of transient of power, flow and temperature, by calculating characteristics of core and by evaluating additional reactivity by steam coming into core region. (5) Internal core structures design and system power balance design.
Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu
JNC TN9400 2003-073, 272 Pages, 2003/08
In the feasibility studies of commercialization of an FBR fuel cycle system,the targets are economical competitiveness to future LWRs, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation, besides ensuring safety. Both medium size pool-type lead-bismuth cooled reactor with primary pumps system and without primary pumps system are studied to pursue their improvement in heavy metal coolant considering design requirements from plant structures. The design of plant systems are reformed, and the conceptual design is made and the commodities are analyzed. (1)Conceptual design of lead-bismuth cooled reactor with pumping system Electrical output 750 MWe and 4-module system The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (2)Structural analysis of main components (3)Conceptual design of natural circulation type lead bismuth cooled reactor Electrical output 550 MWe and 6-module system The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (4)Study of R&D program
Enuma, Yasuhiro; ; Soman, Yoshindo; Konomura, Mamoru
GENES4/ANP2003, P. 1085, 2003/00
; ; Soman, Yoshindo; Konomura, Mamoru
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
Hayafune, Hiroki; Enuma, Yasuhiro; Mizuno, Tomoyasu; Soman, Yoshindo; Konomura, Mamoru; Mito, Makoto*; Tanji, Mikio*
Roshia Kosokuro Kokusai Kaigi, 0 Pages, 2003/00
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru; Kasai, Shigeo; Soman, Yoshindo; Shimakawa, Yoshio; Hori, Toru; Chikazawa, Yoshitaka; Miyahara, Shinya; Hamada, Hirotsugu; et al.
JNC TN9400 2003-002, 109 Pages, 2002/12
Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. (1)Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow down system must be evaluated to realize its performance. (2)In-service inspection (ISI&R). The viewpoint of the commercialized plant's ISI&R was organized by comparing with the prototype reactor's ISI&R method. We also investigated short-term ISI&R methods without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize them with the present technology. Hereafter, the ISI&R of the commercialized plants must be defined by considering its characteristics. (3)Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was ...
; Soman, Yoshindo; ; ;
JNC TN9400 2002-029, 92 Pages, 2002/05
A primary sodium pump is installed in the center of an integrated component and heat transfer tubes surround the pump. Then, the pump rotation induces the vibration of heat transfer tubes and it leads the tubes to fretting wearing against support plates. Therefore, the tube wearing must be evaluated to confirm its integrity during the plant life span(60 years). This report describes the results of the tube wearing analysis by using vibration and wearing calculation models. In the first place, the vibration analysis of a pump shaft, shells, tube bundle etc. of the integrated component reveals its properties such as frequency, amplitude and vibration mode. In the second place, based on the above mentioned vibration analysis, the wearing analysis model shows the frequency and amplitude of the fretting wearing between tubes and support plates and the wearing depth of tubes. The amplitude of the pump vibration, vibration transmission paths and the contact condition between tubes and support plates especially affect the tube wearing, then the wearing evaluation needs that conservative calculation conditions must be found out by surveying these parameters. This calculation result indicates that the tube abrasion does not affect the tube integrity during the plant life time. However further evaluation by more detailed analysis and vibration and wearing tests are needed to acqire more accurate results.
; ; Soman, Yoshindo
JNC TN9400 2002-024, 55 Pages, 2002/05
Compared to a sodium cooled FBR, the cost ratio of a steam generator(SG) to the total cost of Pb-Bi cooled FBR plant is relatively high, because the main piping, pumps and IHXs are eliminated in the natural circulation Pb-Bi FBR. Then, reducing the size of the SG has a greatly influence on the total construction cost in Pb-Bi cooled FBR. In this study, the size of the SG of the natural circulation Pb-Bi cooled FBR was evaluated by using one dimensional steady thermal-hydraulic calculation code (POPAI-6) to obtain more benefit of the plant construction cost. Raising Pb-Bi temperature and adopting heat transfer promoting tubes (which provide grooves inside) to the SG are considered to be the effective method of reducing the heat transfer area. However, another study revealed that Pb-Bi is inferior to sodium in the wetting to the tube surface and this property leads to the less thermal convection. Thus computer code POPAI-6 indicated these effect on the heat transfer area of the SG. Raising Pb-Bi temperature and adopting heat transfer promoting tubes achieve a 23% and 15% decrease of heat transfer area respectively. And both methods enable to diminish 34% of the heat transfer area. However, the less wetting of Pb-Bi to the tubes surface considered, the heat transfer area was estimated to increase by 14%. This effect must be considered in designing the SG and Plant system.
Sato, Kazuyoshi; Neyatani, Yuzuru; Maruo, Takeshi; Mukai, Satoru*; Uchida, Shoji*; Soman, Yoshindo*
no journal, ,
no abstracts in English