Takafumi Anegawa

An experimental study on rewetting phenomena in transient conditions of bwrs

Abstract It is known that rod temperature rise after boiling transition (BT) is not excursive and that the peak cladding temperature (PCT) is suppressed by rewetting to return to nucleate boiling, even if BT occurs under severe conditions exceeding abnormal operational transients for a BWR. The purpose of this study is to develop and verify the rewetting correlation. The rewetting correlation was developed based on single rod data, as a function of quality, mass flux, pressure and heat flux. The transient thermal-hydraulic code used in the BWR design analysis (SCAT) with this rewetting correlation was compared with transient rod temperature result after the occurence of BT obtianed by the 8×8 and 4×4 rod bundle. It is concluded that the transient code with the developed rewetting correlation predicts the PCT conservatively, and the rewetting time well.

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

Abstract An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor.

Simulation of boiling water reactor one-pump trip transient by TRAC/BF1-ENTRÉE

Abstract The three-dimensional plant simulator TRAC/BF1-ENTRÉE was validated based on a one-pump trip test. Trends in major plant process parameters and three-dimensional power distributions were studied with regard to in-core flow reduction, insertion of control blades, and neutron spectrum mismatch. An improved moderator direct heating model was proposed by separately modeling the neutron slowing down and the gamma-ray absorption mechanism. The delayed heat conduction caused by the gamma heating in metallic regions was implemented. Sensitivities of water level and three-dimensional power were studied by varying the core power history, the dryer loss coefficient, and the neutron kinetics solution approach.

Qualification of a Three-Dimensional Core Dynamics Analysis Program Coupled with a Detailed Mesh Division for Commercial Boiling Water Reactors-I

In the stability licensing analysis and evaluation of boiling water reactors (BWRs), frequency-domain stability analysis programs have been used in Japan. To back up the licensing analysis and evaluation, time-domain, multiregional analysis programs have been used because more detailed analytical results can be obtained by these programs with little more computer time than that used by the frequency-domain stability analysis programs. In the backup calculation by time-domain, multiregional analysis programs, many trial-and-error experiments and much expertise on the reactor core radial regional division and on the initial disturbance input are necessary to analyze properly the stability of the BWR core, particularly its regional nuclear thermal-hydraulic stability. A three-dimensional time-domain core dynamics analysis program called SUPER-STANDY was developed with a detailed mesh division that makes various trial-and-error procedures and experience-based expertise unnecessary and that can treat the stability peculiar to the BWR core accurately. The program was applied to a plant where regional instability was observed, and the results were qualified. They showed that BWR stability can be analyzed using SUPER-STANDY by adding only the core uniform initial disturbance input without considering the reactor core radial regional division. It was determined that core regional mode instability can be properly analyzed by the multiregional analysis program (a) by dividing the core into six or more radial regions, (b) by specifying the hot fuel bundle as one region, and (c) by specifying the surrounding fuel bundles around the hot fuel bundle as one region. A visual display system was also developed for a huge number of stability data and core nuclear thermal-hydraulic characteristics, which are connected to each other in a complex way. These are obtained by the SUPER-STANDY program.