Session: 04-02: SMRs, Advanced Reactors and Fusion

Paper Number: 133805

133805 - Development of Low-Pressure Loss Separator for Advanced Bwrs 

Abstract: 

In BWRs, a mixture of steam and water is produced by the reactor core. Steam separators are used downstream the reactor core to separate the steam from the water. The separated water, which still contains a small amount of carried-under steam, is recirculated back to the core inlet. The separated steam, which contains a small amount of carried-over liquid, undergoes additional separation in steam dryer before leaving the reactor pressure vessel for the turbine. The steam separators are designed to provide good steam-water separation ratio (limit the carry-over ratio of liquid water in steam and carry-under ratio of steam in water) without inducing too large pressure loss on the recirculation flow. The steam separators work by forcing the incoming two-phase flow into rotation within the cylindrical separator barrel. The centrifugal force separates the heavier water from the steam by forcing the water to the outskirts of the barrel, from where it is redirected towards water outlet by so called pick-off rings. Good steam-water separation requires stronger rotation within the barrel but inducing stronger rotation increases the pressure loss presenting a trade-off. Many studies have been conducted aiming to both improve the separation performance and lower the pressure loss. This work introduces recent activities of Hitachi-GE Nuclear Energy. A previously developed design prototype was optimized to further improve performance and also improve the manufacturability by taking the advantage of quick prototyping with CFD modelling. The performance of selected design prototypes was further confirmed using half-scale air-water testing at room temperature and atmospheric pressure. For a final confirmation of the performance of the final design prototype, a testing at nominal BWR operating conditions of 7 MPa and 290 °C is being planned in Hitachi Utility Steam Test LEading facility (HUSTLE). The CFD modelling faced the challenge of complex two-phase flow behaviour within the steam separators with regions inside the separator that have very different two-phase flow regimes. This makes it very challenging to develop a model capable of accurately predicting the behavior of the flow inside a steam separator and predict the performance. A CFD methodology was developed to allow a comparative evaluation of various design options with a reasonable computational cost adopting a simplified model. The design prototypes optimized with the use of CFD were manufactured using 3D printing and the prototypes were tested in a half-scale test using room temperature atmospheric pressure with air and water two-phase flow. The test conditions were scaled from the actual BWR operating conditions, and the test results were used to confirm the performance of separator design prototypes. The final testing at nominal BWR operating conditions at HUSTLE is being planned. The test section was designed to be similar to BWR configuration with additional equipment to measure the separator performance. This includes carry-over measurement system designed to confirm the ratio of liquid water in the separated steam and the carry-under measurement system designed to confirm the ratio of steam in the separated water. The validity of the testing methodology was confirmed by conducting a test with a reference separator design. The final test will confirm the performance of the developed low-pressure loss separator and will allow adopting the separator in future advanced BWR builds. In addition, a potential replacement of older separators in operating BWR plants would lead to improved plant performance and safety.

Presenting Author: Antonin Povolny Hitachi-GE Nuclear Energy

Presenting Author Biography: Antonin Povolny got his MS in Nuclear Engineering at FNSPE CTU in Prague and a PhD in Nuclear Engineering from Tokyo Institute of Technology. He has been working in Hitachi-GE Nuclear Engineering as engineer responsible for safety simulations and reactor island design (including the design of BWRX-300 SMR) focusing on thermohydraulical simulations (CFD, etc.) and separated effect testing.

Authors:

Antonin Povolny Hitachi-GE Nuclear Energy
Kenichi Katono Hitachi-GE Nuclear Energy
Naoyuki Ishida Hitachi-GE Nuclear Energy
Kiyoshi Fujimoto Hitachi-GE Nuzlear Energy
Kenichi Yasuda Hitachi-GE Nuclear Energy
Kazuaki Kito Hitachi-GE Nuclear Energy