Yoshihiko Ishii

Subchannel analysis to investigate the fuel assembly for the supercritical-water-cooled power reactor

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs. The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.

Verification of a Three-dimensional Multi-energy Group Kinetic Analysis Model by Using BWR Critical Operation Data

A time-dependent diffusion model of a three-dimensional, multi-energy group kinetic analysis program STAND has been verified by using neutron detector responses in the subcritical condition. The responses were obtained from a BWR start-up examination. STAND is based on a polynomial nodal expansion method and an improved quasi-static model. It can simulate transients and reactivity accidents of boiling water reactors in critical and subcritical conditions, and works efficiently on a vector computer. This paper describes the core neutronic model including treatment of external neutron sources and verification results tracing about a 10-min control rod operation during the BWR start-up examination. Calculated transient detector responses had good agreement with experimental ones. The CPU time necessary to calculate Is of behavior for a full core (460 radial meshes×30 axial meshes) with three energy groups was about 13 s on the HITACHI S-3800 vector computer when the shape function time step was set to 0.1–0.2 s.

Development of Inherently Safe Technologies for Large Scale BWRs: (5) Operation Support System for Plant Accidents

When accident events are caused by a large-scale natural disaster, conditions beyond those at the plant site may affect the accident. As well, quick diagnosis and recognition of damaged equipment are necessary. We have been developing inherently safe technologies for boiling water reactor (BWR) plants in response to these. An operation support system for plant accident events is one of these technologies. Our operation support system identifies accident events and predicts the progression of plant behavior.The system consists of three main functions: sensor integrity diagnosis, accident event identification, and plant simulation functions.The sensor integrity diagnosis function diagnoses whether sensor signals have maintained their integrity by correlating redundant sensors with the plant design information.The accident event identification function extracts a few of candidate accident events using alarm and normal sensor signals received by the sensor integrity diagnosis function. The scale and position of the accident event are determined by comparing plant simulation results with normal sensor signals.The plant simulation function uses a detailed three-dimensional model of the nuclear reactor and plant. This simulation can predict future plant behavior on the basis of identified accident events.This proposed operation support system provides available results of accident event identification and plant condition prediction to plant operators. This system will reduce the occurrence of false identifications of accident events and human errors of operators.Copyright

Plant Simulation System for Developing ABWR Automatic Power Regulator System

We developed a plant simulation code system to design and verify an automatic power regulator (APR) system for an advanced boiling water reactor (ABWR). The point kinetic plant simulator AUTSIMU is an off-line and real-time online plant simulation code that can simulate the start-up process from a subcritical condition of cold water to a 100% rated power condition as well as the reactor shutdown process. The three-dimensional effect of neutron detector signals in the core was simulated by combining predetermined three-dimensional analysis and point kinetic analysis results. The AUTSIMU code was tested using ABWR start-up experimental data. The calculated results were in good agreement with the measured ones. Less than one minute was needed in the off-line mode to simulate a ten-hour transient with an ordinary personal computer. We confirmed that simple AUTSIMU models could effectively simulate ABWR start-up behaviors. The APR hardware was effectively verified in the online mode through factory tests before being shipped to the nuclear site.

Multi-Channel Thermal Hydraulic Model for LOCA Analysis of Heterogeneous BWR Core

A multi-channel thermal hydraulic model for LOCA analysis of a heterogeneous core such as a HCBWR has been developed. This model solves integral formulations for basic equations based on a one-dimensional drift flux model. The core region is divided into several fuel channel groups which differ in their thermal power or geometry. The various flow patterns in the core are determined by calculating the redistribution of vapor generated in the lower plenum into the fuel channel groups. In order to verify the multi-channel model, a computer program FLORA was developed based on the multi-channel model and large and small break LOCA experiments conducted in the Two Bundle Loop (TBL) facility were analyzed by the FLORA program. As a result, the difference in thermal hydraulic behavior between two bundles with different power in the various break LOCA experiments were well simulated.

Analysis of Isolation Condenser Behavior by TRACG Code

For countermeasure of the station black out, an isolation condenser (IC) with enhancing direct-current power supply system is one of effective devices. But not so much transient heat transfer data were obtained for full scale ICs. The IC performance at the accident of the Fukushima daiichi nuclear power plant unit 1 was evaluated by TRACG code. The pressure of reactor pressure vessel was well simulated to measurements.Copyright

The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)

A new concept of a small and medium sized light water reactor, named the double MS: modular simplified and medium small reactor (DMS) was developed. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts, and effective use of proven technology. The decrease in the primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about 2 m compared with 3.7 m in the conventional boiling water reactor (BWR). The short active fuel length reduces the drop in core pressure and overcomes the natural circulation system. By using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to confirm transient performance, the DMS plant performance under transient conditions was evaluated using the TRACG code. TRACG code, which can treat multidimensional hydrodynamic calculations in a RPV, is well suited for evaluating the DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As a result, the maximum change in the minimum critical power ratio of the DMS was 0.14, almost the same as for the current advanced boiling water reactor (ABWRs). In order to improve safety efficiency developing an emergency core cooling system (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following: (1) eliminating highpressure injection systems, (2) adopting passive safety-related systems, and (3) optimizing distribution for the systems and power source for the ECCS. In this way, the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of the loss of coolant accident analysis, we confirmed that the core can be covered by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

Analysis of ABWR Critical Control and Heat-Up Control Operation by TRACG Code

Most startup and shutdown operations in advanced boiling water reactors (ABWRs) are automated by an automatic power regulator (APR). Hitachi and Hitachi-GE utilized the three-dimensional transient analysis code TRACG to design and verify the APR control algorithms. To verify the algorithms, an external neutron source model that makes it possible to simulate a sub-critical initial core, a water temperature reactivity model, a startup range neutron monitor (SRNM) model, and the APR system models were developed and coded onto the TRACG code. The improved TRACG code has been tested and verified with ABWR startup test data. In the test, the criticality was achieved 40 min after beginning of control rod (CR) withdrawal. The code results, for example, CR operation timing, CR withdrawal length, and signals of the neutron sensors agreed well with the test data. In the heat-up control mode, the measured increasing rate of the reactor water temperature was well simulated for a period longer than six hours.Copyright

Analysis of ABWR Critical Control Operation by TRACG Code

Hitachi utilized three-dimensional transient analysis to design and verify a critical-control mode algorithm of an automatic power regulator (APR). TRACG has a three-dimensional neutron kinetics model based on diffusion theory and a six-equation two-phase flow model. To verify the APR critical-control mode algorithm, an external-neutron-source model that makes possible to simulate a sub-critical initial core, and an APR system model were developed and added on TRACG. The code was verified by comparison of measurements and calculation results of ABWR start-up operation under the critical-control mode. The modified TRACG could simulate neutron count rates of start-up-range neutron monitors (SRNMs), reactor period, control rod operation timing, CR withdrawal length, and time of criticality declaration, well.Copyright

Transient Performance of Medium Small LWR “DMS-400” Evaluated Using TRACG Code

A new concept of a small and medium sized light water reactor, named the DMS (double MS: modular simplified & medium small reactor) has been developed. The DMS features significantly simplified plant systems realized by adoption of a natural circulation system of coolant and a free surface separation (FSS) system that is based on the gravitational separation of steam and water. With these systems, reactor internal pumps and steam separators are not needed, reducing plant cost. In this study, the DMS plant performance under transient conditions has been evaluated using TRACG code. TRACG code, which can treat multi-dimensional hydrodynamic calculations in a reactor pressure vessel (RPV), is well suited for evaluating DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As critical transient cases, generator load rejection with total turbine bypass failure (LRNBP) and loss of feedwater heating (LFWH) were chosen to evaluate. LRNBP and LFWH are the most severely recognized events as a pressure increase event and a thermal power increase event, respectively. In case of LRNBP, heat flux increased to about 110% of rated power, and the natural circulation flow barely changed, resulting in a lower ΔMCPR than that of LFWH case. The reason that heat flux only increased to 110% was because the RPV of the DMS has a large steam region volume in the chimney region compared to the thermal power. As a result, the change in the void fraction with a pressure increase in the core was small. In case of LFWH, the maximum heat flux, calculated using the neutron flux, was 121% of rated power when a scram occurred, and ΔMCPR was 0.14, almost the same as for current ABWRs. Since the analysis conditions were set conservatively, these results show that the DMS performs as well for transient events as conventional BWRs.© 2008 ASME