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Kubo, Shigenobu; Payot, F.*; Yamano, Hidemasa; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; Shibata, Akihiro*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Rouault, J.*; Le Coz, P.*; Garnier, J.-C.*; Hamy, J.-M.*; Hayafune, Hiroki; Iitsuka, Toru*; Mochida, Haruo*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.832 - 837, 2015/05
The French and international industrial partners already joined the project from 2010 to 2013 and many others are also effective in the Research and Development in support of ASTRID. A new partnership is now effective on both topics with Japan. This collaboration on the ASTRID Program and Sodium Fast Reactor is now fully integrated in the ASTRID program organization. In addition a specific Joint Team, CEA, AREVA, JAEA, MHI and MFBR, has been created to follow specifically Japanese contribution and develop evaluations of a common interest to orientate future work and contribute to ASTRID options confirmation and be of an interest for the future Japanese Fast Breeder reactor.
Rouault, J.*; Abonneau, E.*; Settimo, D.*; Hamy, J.-M.*; Hayafune, Hiroki; Gefflot, R.*; Benard, R.-P.*; Mandement, O.*; Chauveau, T.*; Lambert, G.*; et al.
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.824 - 831, 2015/05
The Preconceptual Design phase of the ASTRID Project ended late 2012, the main goal was to evaluate innovative options. It is now followed by the AVP2 phase planned until the end of 2015 whose objectives are both to focus the design in order to finalize a coherent reactor outline and to finalize by December 2015 the Safety Option Report. The CEA acts as the industrial architect of the project. In 2014, Japan which participates now in the design studies and also in Research and Development in support of the ASTRID Project and VELAN are the latest partners to join the Project. The next important milestone is at the end of 2015 with the release by the Project team of a convincing and coherent Conceptual Design file.
Kikuchi, Hirohiko*; Mochida, Haruo*; Ide, Akihiro*; Iitsuka, Toru*; Hayafune, Hiroki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05
Hishida, Masahiko; Konomura, Mamoru; Uchita, Masato; Iitsuka, Toru*; Kamishima, Yoshio*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), P. 5112, 2005/05
An innovative concept of medium-scale sodium-cooled modular reactor, named M-JSFR, has been created as based on the large-scale advanced loop type fast reactor concept. M-JSFR employs other concepts such as standardization and learning effects by designing as a modular plant and reduction of secondary loop number for the purpose of dissolving the scale-demerit. On this M-JSFR, some improvements are performed to overcome the weak point (strong chemical activity of sodium) of a sodium-cooled reactor and to achieve in-service inspection (ISI) and repair as easily as in light water reactors. Based on the ISI guidelines for light water reactors, the ISI procedures are reviewed reflecting the characteristics of M-JSFR. A guideline for ISI with the same grade of that of the light water reactors is established and major components subjected to ISI are selected. Moreover, suitable ISI procedures for each selected major component are proposed, and a plant concept amenable to ISI is studied. As the result of these studies, the construction cost of ISI&R reinforcement M-JSFR increases about 3% mainly because of the diameter expansion of reactor vessel.
Sumino, Kozo; Oshima, Jun; ; ; ; Ozawa, Kenji
JNC TN9410 2001-008, 47 Pages, 2001/03
JNC-TN9410-2001-008.pdf:2.66MB
The nitrogen gas evaporator using steam heating is a main component of the nitrogen gas supply system of JOYO and had been operated without the maintenance in order to supply nitrogen gas to the plant continuously. However, the necessity of replacing the nitrogen gas evaporator occurred due to the corrosion of the tank which involved water and steam for the heating in the recent years. Therefore, the nitrogen gas evaporator using steam was replaced with a new one that has a tank made of stainless steel, and the nitrogen gas evaporator using air heating was newly installed in order to supply the nitrogen gas during the maintenance of the nitrogen gas evaporator using steam heating. In addition, thermometers were installed at water in a tank and supply nitrogen gas in order to monitor these temperatures from the main control room. The main results of a preoperation function test were as follows; (1)It was confirmed that the performance of the new nitrogen gas evaporator using steam heating was more than it of the old one. (2)The nitrogen gas evaporator using air heating could successfully maintain the nitrogen gas atmosphere (within 4% oxygen concentration) in the lower part of the reactor containment vessel. (3)The correlation between the water temperature in a tank and the supply nitrogen gas temperature were confirmed for the normal and maximum operations.
; *; ; Kamide, Hideki
JNC TN9410 99-004, 66 Pages, 1999/01
Instability analysis was carried out using BOST code for a steam generator in a large scale sodium test facility of fast reactors. However, it was found that BOST code gave stable characteristics under the conditions of higher pressure in water-steam system than MONJU conditions, even if the flow ratio of sodium to water was increased as expected to give unstable condition. Here, modification of BOST code was considered and we found some points to be modified. However, main reasons of stable calculation were not resolved. In this report, the current status of BOST code was summarized especially for the stable calculation under the higher pressure condition for further modification and a new code based on current knowledge and coding technique.
; ; Kamide, Hideki
JNC TN9400 99-012, 120 Pages, 1998/10
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a fast reactor with power of 600MWe. The test facility consists of components from a reactor vessel to a steam generator. Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the large scale fast reactor with two primary cooling loops. However the primary piping length and plenum volume are larger than those of the 1/3 scale of the fast reactor, because a sodium is heated by a gas burner which is placed between pump and reactor vessel. And, design of a direct reactor auxiliary cooling system (DRACS) in the test facility is also of interests. Therefore, dynamics analyses of the thermal transition tests have been done in which parameters were the primary cold leg piping length, the plenum volume and Euler numbers of the DRACS system. It was shown that the shortening of the primary cold leg piping length was not important to improve the transient response, and the reduction of the lower plenum volume in the reactor vessel was effective. And, it was found that the Euler numbers of the DRACS should be 1/3 scale condition, and the heat capacity of an air cooler, i.e. its tube size, reduction was of importance. Analyses have been done in which only electrical heaters in the core were used and a temperature difference between hot and cold legs was set 1/3 of the fast reactor. It was shown that the thermal transition just after the scram agreed fairly well between the test facility and the reactor. Analyses of the primary pump stick tests have been done. It was shown that the thermal transition of the fast reactor could be simulated roughly, and a prevention of reverse rotation of the primary pump due to the reverse flow was not very influential to improve the response.
; Kamide, Hideki;
PNC TN9410 98-083, 118 Pages, 1998/07
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a demonstration fast reactor. The test facility is consisted of components from a reactor vessel to a steam generator (SG). Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the demonstration fast reactor with two primary cooling loops and two into one secondary loop. The secondary piping length of the test facility is longer than the 1/3 scale of the demonstration fast reactor. The tests facility has the branch and junction of the secondary piping because of two primary loops and one SG. There is a possibility of flow and temperature unbalance if a buoyancy force were large and pressure loss were small. Therefore, dynamics analyses of the thermal transition tests had been done in which the secondary piping length. To examine the unbalance occurred or not, the natural circulation analysis had been performed providing different heat transfer area of the IHX or presser loss of the primary loop between A loop and B loop. It was shown from the analyses that the temperature response during the transition was delayed in the test model compared to the real reactor. Main cause of the delay was due to the real scaled SG. Other parameters, the length of piping etc., were not very influential to the response. The analysis such predicted that there wasn't large difference of global behaviors between the loops. Therefore, it was shown that there would be no problem, if the difference were made between the loops due to a manufacturing error.
Iitsuka, Toru; Oki, Yoshihisa; Kawashima, Shigeyo*; Nishimura, Motohiko; Isozaki, Tadashi; Kamide, Hideki
PNC TN9410 98-022, 58 Pages, 1998/03
Assessment of the maximum temperature and the position of the hot spot is the most important issues on the reactor safety when the local subchannel porous blockage is occurred. From these background, authors are going to perform a sodium experiment with 37-pin bundle test rig simulating the porous blockage, to understand the phenomena and acquire data for thermal-hydraulic analysis code validation. Before the execution of sodium test, one basic experiment and some using subchannel analysis code ASFRE-III had been done. The basic experiment was a water test to examine the pressure loss characteristics of the porous blockage. The pressure loss correlation derived from the water test was applied to the subsequent subchannel analysis of the 37-pin bundle sodium test rig. The analysis such predicted that the difference between the maximum temperature and the inlet temperature would be in propotion to the power to flow rate ratio, within the condition of the power=100
400 W/cm and the flow rate =200480 
/min. And it was also shown that the maximum subchannel temperature would not over the operational limit temperature 650 
C, if the power to flow rate ratio were kept lower than 0.75(W/cm)//min). The map was made to predict the maximum temperature from the experimental conditions.
Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.
PNC TN9410 96-279, 51 Pages, 1996/08
Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.
Kisohara, Naoyuki; Ikeda, Hirotsugu; Sato, Mitsuru*; Iitsuka, Toru*
no journal, ,
Safety, economy and public acceptance are required for commercialized FBR plant systems. Sodium heated steam generators of the FBRs must satisfy the plant safety and increase the plant availability and the safety impression to the society by decreasing the possibility of Na/water reaction as much as possible. For this purpose, the steam generators of the FBR provide double-wall-heat transfer tubes, and the basic structure of the SG was designed by taking account of Na/water reaction prevention and thermal hydraulic and structural viewpoints.
