Kazuaki Kitou

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.

Development of Inherently Safe Technologies for Large Scale BWRs: (1) Plant System

The Fukushima Daiichi nuclear accident and their consequences have led to some rethinking about the safety technologies used in boiling water reactors (BWRs). We have been developing the following various safe technologies: a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system to better deal with reactor accidents. The above mentioned technologies are referred to as “inherently safe technologies”.The passive water-cooling system and infinite-time air-cooling system achieve core cooling without electricity. These systems are intended to cope with a long-term station black out (SBO), such as that which occurred at the Fukushima facility. Both these cooling systems remove relatively high decay heat for the initial 10 days after an accident, and then the infinite-time air-cooling system is used alone to remove attenuated decay heat after 10 days.The hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding made of silicon-carbide (SiC cladding) and a passive autocatalytic recombiner (PAR). Since the SiC cladding generates less hydrogen gas than the current zircaloy fuel cladding when core damage occurs, the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B), such as an operating floor, can be reduced because the pressure in the PCV can be kept lower with less hydrogen gas generation. The leaked hydrogen gas is recombined by the PAR.When a large-scale natural disaster occurs, fast event diagnosis and recognition of damaged equipment are necessary. Therefore, the operation support system consists of event identification and progress prediction functions to reduce the occurrence of false recognitions and human errors.This paper describes the following items: the targeted plant system; the heat exchange tests conducted for both water-cooling and air-cooling systems; the air-cooling enhancing technology for air-cooling in a 4700 MW thermal power class reactor; hydrogen generation tests for SiC material; and the concept of the operation support system.Copyright

Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents.One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B.The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs.The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.Copyright

Development of Inherently Safe Technologies for Large Scale BWRs: (2) Passive Water-Cooling System

Working from the lessons of the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents. The objective of the study reported here was development of the passive water-cooling system. The above technologies are referred to as ‘Inherently Safe Technology’.The passive water-cooling system works without electricity for the first 10 days after an event to remove a relatively large mount of decay heat from the core. The system consists of a condenser and a steam turbine-driven pump for transferring water from a suppression pool to the reactor. Steam from the reactor pressure vessel is condensed in the condensation tubes of the condenser, and the condensate flows out into the suppression pool in the primary containment vessel (PCV). The water temperature at the condensation tube outlet is lowered to less than the saturated temperature at the partial steam pressure of the maximum PCV design pressure to prevent the PCV failure. The condenser is located at a lower level, e.g., underground, for easier access and for supplying cooling water to a condenser pool without electricity during an event. The lower level condenser pool has an advantage that it can be seismically designed.To evaluate our concept of the water-cooling system, heat transfer tests were conducted using full-scale U-shaped single tubes with three diameter sizes under a wide range of pressure and inlet steam velocity conditions. The heat transfer data were obtained at system pressures of 0.2 to 3.0 MPa (absolute) and inlet steam velocities of 5 to 56 m/s. The heat transfer data with this wide range of pressure and inlet velocity conditions include thermal hydraulics conditions for a passive containment cooling system (PCCS) and some of the data can be extrapolated to isolation condenser (IC) conditions. We also confirmed thermal hydraulics conditions to determine the practicality of our new concept.Copyright

Development of Inherently Safe Technologies for Large Scale BWRs: (3) Infinite-Time Air-Cooling System

To address long-term station black outs, which occurred at the Fukushima Nuclear Power Station, we have been developing the infinite-time air-cooling system which operates without electricity by a natural circulation loop. The air-cooling heat exchanger, which is located outside the primary containment vessel of a reactor, transfers the decay heat to the atmosphere by natural circulation resulting from the density difference of the air. Improvement in the heat-transfer performance of air-cooling is a key technology in the development of the infinite-time air-cooling system. In this paper, we developed the air-cooling enhancing technology for the infinite-time air-cooling system by using a micro-fabrication surface, turbulence-enhancing structures, and heat-transfer fins. To evaluate the performance of this air-cooling enhancing technology, we conducted a heat exchange test using an element test apparatus. A single tube of the air-cooling heat exchanger, which includes a sheath heater and thermo-couples, was used. The air flow outside the tube and the heat quantity were respectively controlled using an air-compressor and the sheath heater. The heat-transfer performance was calculated from the heat-quantity and temperature difference measured using thermo-couples. The developed air-cooling enhancing technology demonstrated superior heat-transfer performance in this test. The heat-transfer performance increased approximately 100 % with this technology compared with a bare pipe. From these experimental results, we confirmed good feasibility for implementing the infinite-time air-cooling system.Copyright

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

Numerical study by large-eddy simulation on effects and mechanism of air-cooling enhancing technologies

ABSTRACT Learning from the lessons of the Fukushima Daiichi Nuclear Power Station incident in which a long-term station blackout occurred, we have been developing an air-cooling system that can operate without electricity for a virtually indefinite time. We developed air-cooling enhancing technologies by using heat transfer fins, turbulence-enhancing ribs and a micro-fabrication surface. To achieve further improvement of the heat transfer performance, it is important to understand the mechanism of the air-cooling enhancing technologies. In this study, we used numerical analysis to investigate the effects and the mechanism of the developed air-cooling enhancing technologies. We confirmed that the Nusselt number was increased 75% by the heat transfer fins. In the heat transfer enhancement by the turbulence- enhancing ribs, the Nusselt number was increased 43% by the turbulence-enhancing ribs. The enhancement ratio of the Nusselt number by the micro-fabrication surface can be explained by the apparent thermal conductivity. The Nusselt number was increased 4%–8% by adding the micro-fabrication to the surface of the pipe with the turbulence-enhancing ribs. For the combination of the micro-fabrication surface and the turbulence-enhancing ribs, the interaction between the better heat transport in the thermally conductive layer and the mixing effect by the large-scale vortex is the heat transfer enhancement mechanism.

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

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