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Shono, Akira; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto
JNC TN9400 2003-074, 401 Pages, 2003/08
JNC-TN9400-2003-074.pdf:48.95MB
Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fue1, and the Peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the criticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastjc cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nucljde, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large e
Kimura, Nobuyuki; Miyake, Yasuhiro*; Miyakoshi, Hiroyuki; Nagasawa, Kazuyoshi*; Igarashi, Minoru; Kamide, Hideki
JNC TN9400 2003-077, 96 Pages, 2003/06
JNC-TN9400-2003-077.pdf:3.96MB
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing causes high cycle thermal fatigue in structural components, is of importance for structural integrity and reactor safety.Thermal conductivity of sodium is approximately 100 times larger than that of water. Thus, temperature fluctuation characteristics will be different between sodium, which is used as a coolant of a fast reactor, and water, which is used in general industries. In this study, a comparison of convective mixing among jets was performed in parallel triple wall jets with the same geometries between sodium and water. The discharged velocity in the sodium experiment was experimental parameter and set at the same velocity and the same Reynolds number in comparison with the water experiment. And also, the velocity ratio among the triple jets was varied to change flow pattern. It was seen that the water jets were mixed in slightly closer region to the nozzle than in sodium jets. As for the power spectrum densities (PSD) of temperature fluctuation, the PSD of sodium was similar to the PSD of water under the same discharged velocity condition. At the neighborhood of the wall, the lower frequency component in the PSD of sodium decreased in comparison with the PSD of water. It was shown that the amplitude and frequency characteristics obtained by rain-flow method, which was important to evaluate structural damage by the thermal fatigue, were identical between sodium and water overall. These experimental results show that water experiment could simulate the frequency and the amplitude in temperature fluctuation characteristics in the sodium cooled reactor.