| 1. |
Data Report of a tight-lattice rod bundle thermal-hydraulic tests, 1; Base case test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests as one of essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor which aims to achieve a high breeding ratio and super high burn-up by innovative performance-up of water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition(BT)(Subjects:BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the base case test section. The thermal-hydraulic characteristics using the large scale test section were obtained for the critical power, the pressure drop and the wall heat transfer under a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Effects of local peaking factor on the critical power were also obtained.. |
| 2. |
Data report of tight-lattice rod bundle thermal-hydraulic tests, 2; Gap-width effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT)(Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one.. |
| 3. |
Data report of tight-lattice rod bundle thermal-hydraulic tests, 3; Rod-bowing effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests were performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3 mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0 mm) and Rod-bowing one)). In the present report, we summarize the test results from the rod-bowing effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the rod-bowing effects were also investigated based on the comparison between the results using the base case test section and the rod-bowing effect one.. |
| 4. |
An Improved Critical Power Correlation for Tight-Lattice Rod Bundles
Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced:Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration with a gap of about 1 mm between rods. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced: Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration with a gap of about 1 mm between rods. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region ( 300 kg/m^2s), it is written in local critical heat flux-critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m^2s and pressure from 2 to 11 Mpa.. |
| 5. |
Tight-Lattice 37-rod Bundle Thermal-Hydraulic Tests and Model Experiments
Main purposes of the tight-lattice 37-rod bundle thermal-hydraulic tests (37-rod tests) and model experiments are to investigate the scale effect (rod-number effect) on critical power and to obtain the detailed thermal-hydraulic data for understanding the phenomena and estimating the advanced numerical analysis codes, respectively. The 37-rod bundle test section and some test sections for the model experiments simulate the Reduced-Moderation Water. Reactor core. It was found from the comparison of 37-rod test data with existing 7-rod test data that critical quality increase with increasing the rod number. Using the spacer-effect fundamental neutron radiography experiment, void fraction distribution around the object, which simulates the spacer, in a heated tube was discussed. From the 14-rod bundle neutron 3D tomography experiments, it was found that vapor tends to move the center region of the flow channel.. |
