Masao Chaki

Thermal Hydraulic Test of Advanced Fuel Bundle With Spectral Shift Rod (SSR) for BWR: Steady State and Transient Test Results and Analysis

Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed.SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect.This paper presents the steady state test with varied SSR geometry parameters, the transient test, and the simulation analysis of these steady state and transient tests. The steady state test results showed that the basic functioning principle such as the controllability of SSR water level by flow rate was maintained in the possible range of geometry parameters. The transient test results showed that the change rate of SSR water level was slower than the initiating parameters. The simulation analysis of steady state and transient test showed that the analysis method can simulate the height of SSR water level and its change with a good agreement.As a result, it is shown that the SSR design concept and its analysis method are feasible in both steady state and transient conditions.Copyright

Evaluation of the Sensitivity of a Two-Phase Flow Model for the Steam Separators Analysis

Reducing of the pressure losses of steam separator systems of boiling water reactor (BWR) plants is useful to reduce the required pump head and enhance core stability design margin. The need to reduce the pressure losses of steam separator systems is especially important in BWR plants that have high power density cores and natural circulation systems. The core flow rate of a BWR plant with a natural circulation system is affected by the pressure losses of steam separator systems. In BWR plants with high power density cores, the core stability design margin is affected by these pressure losses. Generally, reducing the pressure losses of the steam separator systems leads to increased carry-under and carryover. Reducing the pressure losses while keeping the characteristics of both carry-under and carryover is desired, so many studies have been done. The steam separator of a BWR plant consists of a standpipe section, a swirl vane section and three-barrel sections. Two-phase flow of steam and water enters the steam separator through the standpipe section and reaches the swirl vane section. In the swirl vane section, the two-phase flow is given centrifugal force and is basically separated into steam and water. Therefore investigating the two-phase flow characteristics of the swirl vane section is very important. After the swirl vane section, the two-phase flow enters the barrel sections. Each barrel has a pick-off ring. The water in the barrel section is mainly removed by these pick-off rings because the water mainly flows upward as a liquid film in the barrel section due to the centrifugal force given in the swirl vane section. We researched the effect of using the drag force model of the swirling two-phase flow in analyzing a steam separator and we found that the drag force model greatly affects the results of the analysis.Copyright

Thermal Hydraulic Test of Advanced BWR Fuel Bundle With Spectral Shift Rod (SSR): Overview and Pre-Test Analysis by TRACG Code

Japanese national project of next generation light water reactor (LWR) development started in 2008. As one of its development items, the thermal-hydraulic test of spectral shift rod (SSR) is planned. A new component called SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. In this paper, its test plan overview and pre-test analysis by TRACG code is presented. The test plan was developed with the purpose of evaluating SSR thermal-hydraulic characteristics at the actual BWR operating condition (7MPa), such as the controllability of SSR water level, and obtaining data for the validation of calculation method. In the test plan, several types of SSR simulation which covers SSR design in both next generation BWR and conventional BWR were designed. Also test operating conditions such as thermal-hydraulic parameters are determined. In order to evaluate these test specifications, pre-test analysis by TRACG code was conducted. Analysis results of each parameter’s effect on SSR characteristics are consistent with SSR mechanism, which shows that the actual operating condition for SSR fuel is simulated well.Copyright

Development of BWR Power Uprate Method Based on “Heat Balance Shift” Concept

We have developed an innovative power uprate method for boiling water reactors (BWRs) that will increase thermal power by more than 5% without having to replace the high-pressure turbine. Reactor power uprate of nuclear power plants is an efficient plant operating method. Most BWR plants need to replace high-pressure turbines when thermal power is increased to over 5% because the main steam flow rate exceeds the inlet steam flow rate limit of the high-pressure turbine. A conventional power uprate method increases the feedwater and main steam flow rate in proportion to increase in thermal power. We examined a decrease in feedwater temperature instead of an increase in the feedwater and main steam flow rate. Since a decrease in feedwater temperature leads to a smaller main steam flow rate, a power uprate of over 5% can be achieved without replacing the high-pressure turbine. We call this power uprate method the “heat balance shift” method. In the present study, we evaluated the heat balance shift method to determine if it can increase electric power to a level higher than that of the conventional power uprate method without replacing the high-pressure turbine.Copyright

Development of a Thermal-Hydraulic Analysis Code in the Distributed Parameter System for Furnace Water Wall Tubes of Fired Boilers(Fluids Engineering)

For design optimization of the furnace water wall of pulverized coal-fired boilers, a thermal-hydraulic analysis code to calculate the flow rate and wall temperature distribution in the water wall has been developed. The code can be applied to complex flow networks for two-phase flow consisting of many tubes and headers for both steady and transient states in the distributed parameter system. The detailed flow conditions at any tube position, such as temperature, quality, velocity, etc. can be calculated stably and quickly by the developed algorithm using a matrix solver. The results of 3-D combustion analysis are used as one of the boundary conditions for this code. The performance of the code was verified using steady and transient states data of temperature distribution in the furnace water wall tubes of actual boilers. The results of both steady and transient states analyses showed good agreement with measured data. The new algorithm decreased the calculation time for steady-state analysis significantly compared with the previous algorithm. For transient-state analysis with a high performance computer, it can calculate tube temperature faster than the real transient time.

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.