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Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro
JAEA-Technology 2014-043, 36 Pages, 2015/02
JAEA-Technology-2014-043.pdf:8.94MB
The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.
Advisory Committee on Monju Safety Requirements
JAEA-Evaluation 2014-005, 275 Pages, 2014/11
JAEA-Evaluation-2014-005.pdf:134.4MB
In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of "Monju" based on TEPCO Fukushima Daiichi Nuclear Power Plant accident occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up "Advisory Committee on Monju Safety Concept" consists of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to prototype FBR "Monju" considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee.
Yamada, Fumiaki; Hashimoto, Akihiko*; Kato, Mitsuya*; Arikawa, Mitsuhiro*
no journal, ,
In this report that review on the safety evaluation for the consequence of Large Pipe Break in Primary Heat Transport System on the Monju used experimental data.
Kato, Shinya; Kato, Yuko; Kitano, Akihiro; Ueyama, Masahiko*; Fukuchi, Ikuo*
no journal, ,
no abstracts in English
Chiba, Yusuke; Ichikawa, Shoichi; Ono, Shimpei; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Abe, Hisashi*; Kitano, Hiroshi*
no journal, ,
The lithium chloride-type dew point detector was used for reactor containment vessel the entire leak rate test (CV-LRT) in fast-breeder reactor Monju (Monju) which needs the maintenance every three months. But when the plant process is considered, it's desirable for the maintenance periods beyond 12 months. Thereupon the capacitance-type dew point detector made in VAISALA Corporation was nominated as the alternative lithium chloride-type dew point detector in Monju. The lithium chloride-type dew point detector, when it's adopted, for, to have to be satisfied a request of the regulations, it's necessary to estimate the performance of the lithium chloride-type dew point detector. In the condition of the CV-LRT, the dew point data were measured by the capacitance-type dew point detector and the lithium chloride-type dew point detector for 24 hours. These data were compared and the performance of the capacitance-type dew point detector was estimated. Furthermore, the dew point data of Monju under the atmosphere ware continuously measured by the capacitance-type dew point detector and the mirror surface-type dew point detector for 2 years. These data were compared and the performance of the capacitance-type dew point detector was estimated. As a result, the capacitance-type dew point detector was confirmed that had the instrument precision that JEAC4203-2008 required.