Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1250 - 1257, 2016/04
After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.347 - 356, 2016/04
A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of initiating event for local faults (LFs) such as adventitious fuel pin failure, local overpower and flow blockages. An evaluation on the consequences of HTIB was performed in the past study as an enveloping event analysis among a wide spectrum of initiating events for LFs. Although the SAS4A code has been validated by a number of in-pile experiments in the French CABRI and U.S. TREAT reactors, most of them were performed under loss-of-flow (LOF) combined with transient-overpower (TOP) conditions. The changing rate of flow in HTIB is much more rapid than that in these LOF test. Therefore additional and more expedient validation was undergone in this study using the TIB experiments which were performed in the French SCARABEE reactor especially for its modeling of coolant boiling, cladding melting, molten cladding motion, fuel melting and wrapper tube failure anticipated to occur during an HTIB condition. Four TIB experiments were performed with 19 or 37 pin bundles in the SCARABEE reactor. SAS4A analyses on these experiments showed good agreement with those experimental results in the following phenomena which were anticipated to occur during an HTIB condition: (1) Timing and progress of coolant boiling and cladding dryout; (2) Timing of cladding melting and behavior of molten cladding relocation; (3) Timing and progress of fuel melting, disruption and relocation; (4) Timing of wrapper tube melt-through. Therefore it can be concluded that the validity of SAS4A application to safety evaluation on the consequence of HTIB in the past study is enhanced in this study.
Imaizumi, Yuya; Fukano, Yoshitaka
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.357 - 363, 2016/04
SAS4A is the code which has been developed to analyze the initiation phase of the core-disruptive accident in SFRs. The code of which can be adopted in a safety licensing needs to be validated through the experimental results. In this study, the code was validated by the experimental results of CABRI project which was conducted in the framework of international collaboration. The selected three CABRI tests of this validation target were all conducted using annular fuel pellets with middle burn-up (6.4 at%). Severe conditions consisted of loss of flow (LOF) and transient overpower (TOP) was imposed in the tests to reproduce similar conditions when unprotected-loss-of-flow (ULOF) occurred in SFRs. The TOP were imposed when coolant temperature reached around the boiling point or several seconds after the cladding melting. The results of the SAS4A analyses agreed well with the CABRI results such as the timing of coolant boiling, voiding extension during the coolant boiling, and the relocation and refrozen behaviors of the molten fuel. Consequently, the coolant boiling and fuel relocation models of SAS4A were validated by these analyses.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04
Yamano, Hidemasa; Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1029 - 1038, 2016/04
This study conducted the flow-induced vibration evaluation of the primary hot-leg piping in the demonstration reactor design of advanced loop-type sodium-cooled fast reactor in order to confirm the integrity of the piping. Following the description of the primary hot-leg piping design and a design guideline of the flow-induced vibration evaluation, this paper describes mainly the flow-induced vibration evaluation and thereby the integrity assessment. In the fatigue evaluation for the flow-induced vibration, the pipe stresses considering the stress concentration factor and so on, at representative locations were less than the design fatigue limit. Therefore, this evaluation confirmed the integrity of the primary hot-leg piping in the demonstration reactor.
Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohashi, Yuko
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.314 - 320, 2016/04
Temperature and flow rate measurement of each fuel subassembly outlet is effective to detect loss of coolant accident (LOCA) and reactivity initiated accident (RIA) early and to understand a thermal hydrodynamic performance in liquid metal fast breeder reactor (LMFBR). This report shows the data of eddy current type flow meters in Monju. High linearity between the signal intensity of each eddy current type flow meter and flow rate of primary sodium was obtained in the flow rate condition of 10100%. In addition, the linearity was also demonstrated in the low velocity region, approx. 0.25 m/s. Fluctuation shown on each eddy current type flow meter was below 0.2 m/s, which is 5 % of the time averaged velocity at the rated condition. Those experimental results show that the eddy current type flow meter can detect the change of relative flow rate.
Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04
Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.
Hourcade, E.*; Curnier, F.*; Mihara, Takatsugu; Farges, B.*; Dirat, J.-F.*; Ide, Akihiro*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1740 - 1745, 2016/04
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 highlights of ASTRID DHRS current strategy. Focus is made on operating temperature diversification for in-vessel heat exchanger as well as core catcher coolability by an original features such as heat exchanger located within reactor cold pool, whose design was taken over by Japan team since 2014.
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.