Thermophysical properties of stainless steel containing 5mass%-BC in the solid phase

Takai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1007 - 1013, 2018/04

Evaluation of feedback reactivity coefficients by inverse kinetics in Monju

Kitano, Akihiro; Nakajima, Ken*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04

The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.

ASTRID nuclear island design; Update in French-Japanese joint team development of decay heat removal systems

Hourcade, E.*; Mihara, Takatsugu; Dauphin, A.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.556 - 561, 2018/04

In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving an update concerning ASTRID DHR strategy with description of reference architecture evolution and project objectives. In particular, new developments were made for DHR during normal shutdown and role of Ex-Vessel system. A special focus is made on design process of automatic shutter to hydraulically connect Hot Plenum and cold plenum to enhance primary vessel natural convection.

Seismic evaluation for a large-sized reactor vessel targeting SFRs in Japan

Uchita, Masato*; Miyagawa, Takayuki*; Dozaki, Koji*; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hayafune, Hiroki; Suzuno, Tetsuji*; Fukasawa, Tsuyoshi*; Kamishima, Yoshio*; Fujita, Satoshi*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.380 - 386, 2018/04

It is well-known that pool-type SFRs are the main streams recently in a field of Generation IV reactors. The pool-type encloses primary pumps and IHXs located around the core barrel in a main vessel. Consequently, the main vessel diameter trends to be larger than that of loop-types. From the viewpoint of commercialization in the future, a target of the vessel diameter and its weight including Sodium coolant will increase further. In this paper, the prospects are described in terms of seismic design and structural integrity for the thermal loadings to prevent buckling of the reactor vessel based on parameter studies with diameters of the vessel. In addition, the seismic isolation device which will be effective as a countermeasure is proposed in order to secure a margin against buckling of a large reactor vessel.