Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

Development of advanced loop-type fast reactor in Japan, 5; Adoption of self-actuated shutdown system to JSFR

Nakanishi, Shigeyuki; Kubo, Shigenobu*; Takamatsu, Misao; Ikarimoto, Iwao*; Kato, Jungo*; Shimakawa, Yoshio*; Harada, Kiyoshi*

Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.519 - 525, 2008/06

A self-actuated shutdown system (SASS) is a passive safety feature which inserts control rods by the gravity force, where the detachment of the rods would be achieved by the coolant temperature rise under anticipated transient without scram (ATWS) conditions. Various out-of-pile tests have already carried out to investigate the basic characteristics of SASS, and a demonstration test of holding stability under the reactor operation condition has been performed, where a function test of the driving system to re-connect and of pulling out the control rod have been done in the experimental reactor JOYO. The element irradiation tests have been also conducted to confirm that no impact will be foreseen by the irradiation. The effectiveness of SASS for a reference core design of JSFR has been evaluated through all types of ATWS. As a result, it is ensured that JSFR will have a reliable passive shutdown system.

None

*; Ikarimoto, Iwao*; *

JNC TJ4440 2001-005, 55 Pages, 2002/03

JNC-TJ4440-2001-005.pdf:2.41MB

no abstracts in English

Predictions and Experience of Core Distortion Behaviour 発表論文の要約集; IWGFR Specialists Meeting 1-4 Oct.1984 SOUTH MANCHESTER,ENGLAND

*; *; Ikarimoto, Iwao*

PNC TN943 84-12, 85 Pages, 1984/12

PNC-TN943-84-12.pdf:1.84MB

None

Development of seismic assessment method toward the demonstration reactor of JSFR, 1; Vibration test of 1/2.5 scale test assembly in triangular arrangement

Iwasaki, Akihisa*; Monde, Masatsugu*; Sawa, Naoki*; Ikarimoto, Iwao*; Taniguchi, Yoshihiro*; Kitamura, Seiji

no journal, , 

The three-dimensional seismic analysis technology of a reactor core is developed. The vibration test of 1/2.5 scale test assembly in triangular arrangement was carried out, and the three-dimensional response behavior has been grasped.

Development of core seismic assessment method toward the FBR

Murakami, Hisatomo; Kitamura, Seiji; Yamazawa, Tomoyuki*; Ikarimoto, Iwao*; Kan, Taro*

no journal, , 

no abstracts in English

Thermal-hydraulic evaluation for fuel assembly design of next generation FBR in Japan, 3; Numerical analysis for entrance nozzle of fuel assembly

Watanabe, Osamu*; Hayakawa, Satoshi*; Ikarimoto, Iwao*; Ohshima, Hiroyuki

no journal, , 

Numerical investigation using commercial CFD code was performed to estimate flow characteristics of the entrance nozzle of the fuel subassembly of sodium-cooled fast reactor. A three-dimensional numerical simulation of the entrance nozzle test section simulating the fuel subassembly of Monju reactor was carried out. Magnitude of the pressure loss coefficient obtained by the numerical results agreed to that of the experimental result within the uncertainty range required in reactor design. As the result, potential applicability of the numerical simulation using the CFD to design of entrance nozzle of the fuel subassembly was confirmed.