Results 1 - 10 of 157
Results 1 - 10 of 157. Search took: 0.016 seconds
|Sort by: date | relevance|
[en] Cooperation with IAEA on SFR SDC, SDGC: Contributions of IAEA Specialists: IAEA expert (Nuclear Safety and Security) has been a member of SDC TF; IAEA Department of Nuclear Energy, joins GIF, a yearly Workshop of safety of SFR and later of LMFR; and GIF Contributions to IAEA safety standards for SMRs, most of which overlap with Gen IV design tracks. Summary: Long and successful history of cooperation between GIF and INPRO IAEA; Messages of peaceful and sustainable use of Nuclear toward the world; International standards of safety design for Gen IV reactors.
[en] We have investigated cell calculation models to be used in the analysis of neutronic characteristics of a heterogeneous fast critical assembly. As cell models we have considered a single drawer model with a critical buckling, a single drawer model with group dependent bucklings and a multidrawer model which consists of some fuel and blanket drawers. We have compared the cell averaged cross sections obtained from these cell models with the results of a reference transport calculation and estimated the effects of the cell models on ksub(eff), reaction rate ratios, reaction rate distributions and sodium void worths in the heterogeneous fast critical assembly ZPPR-13A. The multidrawer model and the single drawer model with group dependent buckling give reasonable cell averaged cross sections and have large effects on ksub(eff) and 238U fission rate distribution. (author)
[en] A large-scale core fuel subassembly in a real reactor was analyzed under the conditions including natural circulation in order to evaluate thermo hydraulics where it was cooled by inter-wrapper flow or heat transfer among the subassemblies. The core fuel subassembly of a 217-pin bundle was analyzed by using a three-dimensional thermal-hydraulic analysis code, AQUA. The analysis method is the combination of staggered half-pin mesh arrangement and correlations of axial flow resistance, the former approach assigning a control volume to a subchannel surrounded by three fuel pins and the latter one developed for subchannel analysis codes. This method has been validated, based on experimental data on the 37, 61 and 169-pin bundles, including the conditions of sidewall cooling. After implementing the prediction analysis, the following conclusions were obtained: (1) Even if the 217-pin subassembly is cooled through a wrapper tube wall, the peak temperature in the subassembly decreases. A decreasing trend of temperature, including the effects of power/flow rate conditions, can be correlated with buoyancy parameter, Gr*/Re, and heat flux ratio of wrapper wall to pin surface, q''wall/pin. (2) The horizontal temperature distribution in the subassembly varies, dependent on the power/flow rate and heat removal. The peaking factor, the ratio of peak temperature to cross-sectional average temperature, can be evaluated by using buoyancy parameter Gr*/Re, and q''wall/pin considering the dependency on power/flow rate. On the other hand, the temperature near the wrapper tube wall is represented by a wall subchannel factor. It is correlated with q''wall/pin, by grouping the buoyancy parameters. (3) Comparison with the past sodium experimental results, the 217-pin large bundle has higher peaking factors and lower wall subchannel factors than 37-pin bundles. (4) To see an influence of non-homogeneous cooling in axial direction, only the upper-half of subassembly is cooled. As a result, it was found that peaking factors and wall subchannel factors could be correlated with average heat flux on the wrapper wall. (author)
[en] A calculational method of diffusion and drift coefficients for asymmetric slab cells has been derived on the basis of the first-flight collision probability method. The calculated values of the diffusion and drift coefficients for the asymmetric cells used in the fast critical assembly ZPPR-9 are shown. Drift coefficients calculated based on the original Duracz method and Duracz' cell averaging method are numerically compared. Flux distributions for one-dimensional slab models composed of the asymmetric cells such that U-238 plates are bunched at the core center are calculated with and without drift coefficients. It is seen that the Duracz original drift coefficient has well predicted the flux dip at the core center caused by neutron capture in the U-238 plates in lower energy, but the Duracz cell averaged drift coefficient is not able to predict the flux dip. (author)
[en] Wrap-up: • Further progress are underway to increase the technology readiness level of the six Generation IV systems; • New areas – Interest from the private sectors for advanced reactors; – Flexibility requirements • Advanced nuclear energy systems and innovative applications of nuclear technologies can provide solutions underpinning economic growth; • Generation IV systems offer additional features in terms of performance and sustainability compared to existing concepts; • The GIF calls on policymakers to acknowledge the real contribution that nuclear energy is making today to the mitigation of carbon emissions from the power sector, and to consider supporting the deployment of advanced reactors and innovative applications of nuclear technologies.
[en] Reduction of construction cost is one of key issues to develop sodium cooled fast reactors. A compact reactor vessel is useful design approach. When vessel diameter is reduced under constant power condition, fluid velocity increases. The reactor vessel has a free surface. Thus, increase of velocity may cause gas entrainment at the free surface. Then two types of water model experiments were carried out to find adequate geometry in the reactor vessel. One is a 1/10th full sector scaled model of an upper plenum of the reactor vessel. The other is 1/1.8th partial model. Flow field was optimized by using the l/10th model. Scale effect was examined by the l/1.8th model. (author)
[en] A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. Flow visualization, measurement of flow velocity using particle image velocimetry, and detailed temperature measurement using a movable thermocouple tree were carried out. The flow patterns of jet exiting from the branch pipe could be classified into (1) wall jet (2) deflecting jet and (3) impinging jet according to the inflow condition. The velocity measurement showed that, in the wall jet case, the vortices were generated in the wake region behind the branch pipe jet like Karman vortex. This vortex was correlated with prominent frequency of the temperature fluctuation. As for the transfer of temperature fluctuation from fluid to structure, higher frequency component was attenuated significantly. A constant heat transfer coefficient was applied to the prediction of transfer function. (author)
[en] Sodium cooled fast reactor is designed in a feasibility study. PRACS (Primary Reactor Auxiliary Cooling System) is adopted as a decay heat removal system of middle class reactor. In this system heat exchange tubes are installed in the IHX inlet plenum. During the decay heat removal operation cold fluid cooled by PRACS tubes and hot fluid which bypasses the tubes mixes in the inlet plenum and temperature fluctuation will occur. This temperature fluctuation is significant on a point of high cycle thermal fatigue (thermal striping) in a complex component which has a weld line. Sodium experiment was carried out for such mixing phenomena in IHX with the PRACS cooling tubes using PLANDTL test loop in Oarai engineering center. Experimental parameters were flow rate in the primary side in IHX, temperature difference in IHX primary side, and removal heat of the PRACS. Temperatures at the cooling tubes of the PRACS and a upper tube plate of IHX were measured. Following findings were obtained. 1) Temperature fluctuation intensity was large at the tube plate than at the PRACS tube. 2) Temperature fluctuation intensity depended on flow velocity in a IHX inlet plenum including the PRACS tubes, it increases as the flow velocity got larger (1 to 3 cm/s). In the design of module type middle class reactor, average flow velocity in the inlet plenum of IHX is 16 to 4 cm/s during the decay heat removal system. It is needed to pay attention to the velocity dependency of temperature fluctuation at downstream region of PRACS tubes. (author)
[en] A design study of an advanced sodium-cooled fast reactor (Advanced-SFR) has been conducted in Japan Atomic Energy Agency (JAEA). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and the bottom of the UIS may cause high cycle thermal fatigue on the structure around the bottom of the UIS. A water experiment using a 1/3 scale 60deg-sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. By the velocity measurement using the particle image velocimetry (PIV), it was known that the swirling flow in vertical direction from the upper to the lower regions of the CIP was formed around the flow-hole of the back-up control rod (BCR) channel. The large intensity of the temperature fluctuations was observed near the cold fluid outlets simulating the BCR channel. The countermeasure to mitigate the temperature fluctuation intensity was applied and its effectiveness was confirmed. (author)
[en] A quantitative evaluation on high cycle thermal fatigue due to temperature fluctuation in fluid is of importance for structural integrity in the reactor. It is necessary for the quantitative evaluation to investigate occurrence and propagation processes of temperature fluctuation, e.g., decay of fluctuation intensity near structures and transfer of temperature fluctuation from fluid to structures. The JSME published a guideline for evaluation of high-cycle thermal fatigue of a pipe as the JSME guideline in 2003. This JSME standard covers T-pipe junction used in LWRs operated in Japan. In the guideline, the effective heat transfer coefficients were obtained from temperature fluctuations in fluid and structure in experiments. In the previous studies, the effective heat transfer coefficients were 2 - 10 times larger than the heat transfer coefficients under steady state conditions in a straight tube. In this study, a water experiment of T-junction was performed to evaluate the transfer characteristics of temperature fluctuation from fluid to structure. In the experiment, temperatures in fluid and structure were measured simultaneously at 20 positions to obtain spatial distributions of the effective heat transfer coefficient. In addition, temperatures in structure and local velocities in fluid were measured simultaneously to evaluate the correlation between the temperature and velocity under the non-stationary fields. The large heat transfer coefficients were registered at the region where the local velocity was high. Furthermore it was found that the heat transfer coefficients were correlated with the time-averaged turbulent heat flux near the pipe wall. (author)