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[en] Full text: Concluding Remarks. Opening Session: Current status of GIF and IAEA are informed for better understanding and consensus was obtained. Gen-IV reactor systems development status: All reactor systems except for VHTR were shown and IAEA provides activities of CRPs and also online Catalogue. High activities of VHTR and SMR, MSR in IAEA are also shown including collaboration between IAEA and GIF. Opportunity and Challenge for advanced reactor: GIF showed activities for harmonization with increasing Renewable energy system and collaboration with NICE future and CEM10 in Vancouver. IAEA showed the sustainable nuclear energy development with 3 dimensions of Economy, Environment and Society. Collaboration on this issue between IAEA and GIF will provide more fruitful results on the same direction. SMR utilization and also Hybrid system for RES harmonization are shown for the direction to sustainable energy system with RES. 2nd Day Safety Issue: Priority on Advanced NPP designs are shown by IAEA. SMR is one of significant issue requested by many of developers. LWR base and HTGR base concepts are picked up for modification of SSR 2/1 together with Risk-informed approach and graded approach. In the activities of TecDoc of SMRs, technology neutral approach and also Risk informed approach was also highlighted. In the discussion of RSWG of GIF, we have a comment that different approach and different outcomes from IAEA, GIF, WGSAR make confuse for member countries. Needs of harmonization, coherent mechanism and no duplication are recognized. From GIF side some proposals were made for SFR SDC and SDG. Review of SDG for key structure, system and component (SSC) was kindly accepted by IAEA. Discussion of IAEA TECDOC for SFR SDC and SDG toward the IAEA safety standards was very active. Let me summarize as follows, TECDDOC is not the step to IAEA standards but harmonization among various aspects of requirements and regulation issues. IAEA will have internal discussion on this issue. It is confirmed to continue the discussion between IAEA and GIF about the SFR SDC and IAEA activity. GIF will provide TOR on this issue. The GIF activities on R&D and Infrastructure TF are discussed together with IAEA Database of advanced reactor experimental facility. Cooperation and supplement by these two activities are expected for effective international cooperation on the experiments. PRPP is significant issues for commercialization of advanced reactors. Activities in GIF and IAEA are introduced. New activities of safeguard by designs are also recognized as one of significant directions of PRPP. Thank you very much for active discussion in each session. I believe these discussion contribute further cooperation between IAEA and GIF for sustainable use of nuclear Energy and decarbonized society. (author)
[en] Last week, Japanese people had a so sad memory of day, March 11th, on the huge earthquake, tsunami, and 1F accident. So many people, more than 40,000, are still evacuated from their hometown due to the influence of the radio active materials released by 1F accident. Nuclear engineers and researchers, not only in Japan but over the world, thought again how the safety of nuclear is significant and our responsibility. Today, it is our great opportunity to discuss not only for safety but also sustainable and clean energy supply systems using nuclear, together with IAEA specialists and GIF members.
[en] After a reactor scram, the decay heat removal (DHR) is of decisive importance for the safety of the plant. A fully passive DHR system based on natural circulation alone is independent of any power source. The DHE system consists of immersion coolers (ICs) installed in the hot plenum and connected to air coolers, each via intermediate circuits. During the postscram phase, the decay heat is to be removed by natural circulation from the core into the hot plenum and via the ICs and intermediate loops to the air coolers. The function of this DHR system is investigated and demonstrated in model tests with a geometry similar to the reactor, though on a different scale RAMONA is such a three-dimensional model set up on a 1:20 scale. It is operated with water. The steady-state tests for natural-circulation DHR operations have been conducted over a wide range of operational and geometric parameters. To study the transition from nominal to DHR conditions, experiments were defined to investigate the onset of natural circulation in the postscram phase (transient tests). The experiments were analyzed using the one-dimensional LEDHER code. LEDHER is a network analysis code for the long-term DHR of a fast reactor developed at Power Reactor and Nuclear Fuel Development Corporation in Japan. The results of the experiments and conclusions are summarized
[en] An experimental investigation was conducted on convective heat transfer to a local blockage in a simulated subchannel of a Liquid Metal-cooled Fast Breeder Reactor. The experiment was performed with a 4-subchannel geometry water test facility. A porous blockage is located at the center subchannel and is surrounded by three unplugged subchannels. The blockages used in this study were solid metal, a porous blockage consisting of metal spheres, and a porous blockage with plates covering the side or top faces of the blockage to intentionally prevent either the axial and/or the lateral flows through the blockage. In the experiment, the heat flux provided by an electrical heater were set at 50(kW/m2) and 20(kW/m2) while the Reynolds number was varied from 3.5 x 103 to 8.6 x 103. Temperature measurements of the water were made inside/outside the blockage. Finally, velocity profiles outside the blockage were measured with a Laser Doppler Velocimeter (LDV) and an Ultrasound Velocity Profile monitor (UVP). Normalized temperature inside the blockage revealed that the influence of buoyancy was negligibly small, and that the temperature depended on the flow rate and the configurations of the blockage. Comparison of temperature and velocity profiles between the blockage types as shown in Fig. A-1, showed that both lateral and axial flow influenced the heat removal from inside the upper part of the porous blockage, as well as the heater surface contacting the blockage. Father, lateral flow had a strong influence on the peak temperature inside the blockage than axial flow. The heat transfer characteristics showed that the predominant mode of heat was not conduction, but convection via lateral flow through the blockage and axial flow through the upper region of the blockage under higher flow rate conditions
[en] Sodium experiments were conducted on core thermal-hydraulics simulating a scram transient of a large scale fast breeder reactor using the test facility PLANDTL-DHX with seven fuel subassemblies. The influence of inter-subassembly heat transfer on temperature distribution in the subassembly was revealed via measurements. The flow in the gap between neighboring subassemblies called inter-wrapper flow (IWF) was also studied in relation to its capability of cooling the subassemblies. A computational model is presented for predicting the transient without IWF. The multi-dimensional numerical analysis model employs an empirical correlation to simulate mixing effects between adjacent subchannels. It was shown that the present computational method could evaluate the transient behavior of thermal-hydraulics in the subassemblies accurately from forced to natural circulation accompanied by inter-subassembly heat transfer and flow redistribution in the subassembly. The cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop attributable to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core.
[en] A proper assessment of core thermohydraulics under natural circulation conditions is important so that the full potential of the inherent, passive feature of a fast reactor can be used. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium exiting the heat exchangers may penetrate into the gap regions between fuel subassemblies; this gap flow between the wrapper tubes of subassemblies is called interwrapper flow (IWF). During natural circulation decay heat removal, IWF will significantly modify the flow and temperature distributions in the subassemblies. Sodium experiments were carried out to investigate these phenomena, using a test section consisting of seven subassemblies housed and connected to an upper plenum. The cooling effect of IWF on the fuel subassemblies was evaluated and a new nondimensional parameter was deduced to characterize this effect. On the other hand, IWF reduced the natural circulation flow in the primary loop due to a temperature decrease in the upper part of the core. A balance between the cooling effect and the flow reduction effect is discussed. Three-dimensional analyses were performed to establish an estimation method for IWF. For the temperature decreases due to IWF at the hottest point in the subassemblies there was good agreement between experiments and predictions
[en] Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor's upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis
[en] In Japan Atomic Energy Agency, an innovative sodium cooled fast reactor of 1500 MWe class, JSFR, has been investigated on the Fast Reactor Cycle Technology Development project. A compact reactor vessel (R/V) and a column type upper inner structure with a radial slit for an arm of a fuel-handling machine are adopted. These result in increase of the spatial-averaged velocity on the horizontal cross section of the R/V by factor of 2.5. These high velocities may cause gas entrainment at the free surface in the upper plenum and also the cavitations. Therefore horizontal dipped plates (D/P) are set below the free surface to prevent the gas entrainment. We performed two water experiments using an 1/10 scaled full-sector model of the upper plenum of R/V and a large scaled partial model. The flow optimization in the upper plenum was performed in the full-sector model. It was observed in the large scaled model that the gas entrainment occurred under the extreme velocity conditions which were far from the rated condition in the reactor design of JSFR. Consequently, there is a chance for this design of the compact reactor vessel to suppress the gas entrainment and cavitation. (authors)
[en] In order to investigate the thermal hydraulic behavior in a mixing tee, fluid temperature and flow velocity distributions were measured by a movable thermocouple tree and Particle Image Velocimetory (PIV). An in-house Direct Numerical Simulation (DNS) code, DINUS-3 was applied to understand mixing phenomena and also to confirm applicability in a mixing tee. The temperature and velocity fields obtained from the DNS were in good agreement with the experimental results. A prominent frequency of temperature fluctuation was calculated in the downstream region in the tee. It was also in good agreement with the experimental data. Measured velocity field showed vortex structure in the wake region behind the jet exiting from the branch pipe. The results of analysis showed that Karman vortex was generated in this wake region and it influenced the major temperature fluctuation in a mixing tee and also the prominent frequency. And the analytical results revealed that the vortices generated in the downstream region had axes in three directions and interacted with each other
[en] In the design of Japan Sodium-cooled Fast Reactor (JSFR), coolant velocity is beyond 9 m/s in the primary hot leg pipe of 1.27 m diameter. The Reynolds number in the piping reaches 4.2x10"7. Moreover, a short-elbow (r/D=1.0, r: curvature radius, D: pipe diameter) is adopted in the hot leg pipe in order to achieve compact plant layout and reduce plant construction cost. Therefore, the flow-induced vibration (FIV) arising from the piping geometry may occur in the short-elbow pipe. The FIV is due to the excitation force which is caused by the pressure fluctuation on the wall. The pressure fluctuation on the pipe wall is closely related with the flow fluctuation. In this study, water experiments using two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0 and 1.5 (short-elbow and long-elbow), were conducted in order to investigate the mechanism of velocity and pressure fluctuation in the elbow and its downstream. The experiments were carried out at Re=5.4x10"5 conditions. Measurement of velocity fluctuation and pressure fluctuation in two types of elbows with different curvature revealed that behavior of separation region and the circumferential secondary flow affected the pressure fluctuation on the wall of the elbow greatly. (author)