Fumitoshi Watanabe

Maximum likelihood estimators for generalized Cauchy processes

Maximum likelihood estimator (MLE) for a generalized Cauchy process (GCP) is studied with the aid of the method of information geometry in statistics. Our GCP is described by the Langevin equation with multiplicative and additive noises. The exact expressions of MLEs are given for the two cases that the two types of noises are uncorrelated and mutually correlated. It is shown that the MLEs for these two GCPs are free from divergence even in the parameter region wherein the ordinary moments diverge. The MLE relations can be regarded as a generalized fluctuation-dissipation theorem for the present Langevin equation. Availability of them and of some other higher order statistics is demonstrated theoretically and numerically.

Space-dependency analysis of amplitude and decay ratio based on Forsmark noise data: new approach to contraction of space-dependent information on reactor stability

Abstract We have proposed a new method of contraction of information on spatial-dependency of feasible reactor stability indexes between the amplitude of local power oscillation and the local decay ratio based on nonlinear stochastic model in the complex normal form. The method is applied to neutron noise data form Forsmark. Demonstrated is the importance of quantitative contraction of local information on spatial-dependency to have a better understanding of the state of reactors global stability.

Experiments and Analytical Simulation on Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.Copyright

Visualization Study on Complicated Flow Through Lower Plenum of BWR

The increase of power outputs enables us to decrease a generation cost such as over thirty nuclear power plants which have adapted an up-rating method in the United States. To success it, it is necessary to evaluate detail thermal hydraulics behavior with high accuracies due to the severe use of internal core structures. The evaluation of coolant flow at a lower plenum of an advanced boiling water reactor (ABWR) is very important because there are a lot of structures such as control rod guide tubes (CRGT) and the core support beams on the fuel assemblies. The coolant flow direction changes from downward to upward with three-dimensional complicated flow in the lower plenum. The simulation results by a CFD (Computational Fluid Dynamics) code can predict such complicated flow in the lower plenum. It is necessary to compare the simulation results with the actual flow in wide range of high Reynolds numbers. And it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of CFD code. In the constructed model of the lower plenum of ABWR, we measured velocity profiles by LDV (Laser Doppler Velocimetry) and PIV (Particle Image Velocimetry) techniques with a high speed video camera. The turbulent flow structure of lower plenum of ABWR was evaluated experimentally. In the range of Reynolds number from 103 to 104 , the velocity at the center of the test section was faster than the velocity near the wall. The intensity of turbulent increased when the Reynolds number was higher. The velocity profiles in downstream showed the tendency to be flat in the core support beam.Copyright

Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.Copyright

Role of Complicated Flow Fields in Lower Plenum on Coolant Flow Distribution to Core of ABWR

In order to achieve increase of power outputs of an ABWR, it is extremely important to evaluate coolant flow in a lower plenum. Numerical simulation is helpful to predict the coolant flow, and thus verification experiments are needed. Hence, the present study is focusing on the construction of benchmark of CFD code. The objective of the study is to clarify relation between flow distribution to the core and the complicated flow structures in the lower plenum.We constructed a 1/10 model of a lower plenum to conduct flow experiment for validation of CFD analysis. In the experiment, it turned out that coolant flow distribution becomes to be uniform at core support beam, and there are complicated flows like vortices around side entry orifices. Regarding differential pressure distribution, it was revealed that profile of the region including side entry orifices is very dominant. This pressure loss might be caused by contraction flow at the orifices. Such flow structures were also described by CFD analysis. As good performance of analysis, we investigated flow distribution to the core particularly by using CFD results. And relation between the flow distribution and the complicated flow structures were also discussed.Copyright

Effects of Lower Plenum Flow Structure on Core Inlet Flow of ABWR

One of the strategies of cost reduction of nuclear power generation is increase of power outputs. In order to achieve increase of power outputs of a Boiling Water Reactor (BWR), it is extremely important to evaluate coolant flow in a lower plenum of a BWR. Although the simulation by a CFD code is helpful to predict the coolant flow in a lower plenum, experimental data to verify the simulation results are not enough, and the simulation results considerably depends on models. Hence, the present study is focusing on the establishment of the benchmark of CFD code by using the visualization method in a lower plenum. The objective of the present study is to clarify the flow structure of a lower plenum in detail, and to investigate effects of complicated flow structure of lower plenum on core inlet flow. We constructed a 1/10 model of a lower plenum to measure velocity profiles with LDV and PIV. And differential pressure of the lower plenum was measured with differential pressure instrument. In the experiment, it turned out that flow structure of the lower plenum vary depending on experimental condition.

ICONE19-43420 INVESTIGATION OF FLOW STRUCTURE TRANSITION IN LOWER PLENUM OF ABWR

In the lower plenum, the velocity at the center section is faster than that near the shroud. The velocity profiles show the tendency to be flat in the core support beam and core inlet region. These experimental and analytical profiles are coincident each other. ・In the PIV results, it is confirmed that vortices arise around the orifices at the corner and near the beam. However, there are no vortices around the orifices at the center of the grid.

Study on Pressure Loss Induced by Complicated Flow Through Lower Plenum of BWR

One of the strategies of cost reduction of nuclear power generation is the increase of power outputs. Especially, in order to achieve performance upgrade of Advanced Boiling Water Reactor (ABWR), it is extremely important to evaluate coolant flow in the lower plenum of ABWR. With the plenty construction in the lower plenum, it is thought that the flow structure is complicated. Moreover, according to the previous studies, there is strong evidence that vortexes arise around side entry orifice when coolant flows in there. Such complicated flow may affect the pressure loss (differential pressure in the lower plenum) and the coolant flow distribution to each core fuel assemblies, and consequently it would influence advancement of fuel economics. Although the simulation results by a CFD code can predict such complicated flow in the lower plenum, the accuracy of simulation data are not enough. Hence, the present study is focusing on the establishment of the benchmark of CFD code by using the visualization method in the lower plenum of ABWR. The objective of the present study is to investigate correlation between the structure of vortexes and complicated flow in upstream of core support beam, and the effect of such fluid behavior to the differential pressure. In the constructed model of the lower plenum of ABWR, velocity profiles were measured by LDV (Laser Doppler Velocimetry) and PIV (Particle Image Velocimetry) techniques. And differential pressure of constructed model is measured by differential pressure instrument. Each measurement was worked out in the range of Reynolds number from 103 to 104 . It was found from the LDV measurement that the velocity at the center of the test section was faster than that near the wall in upstream. In downstream, the velocity profiles showed the tendency to be flat in the core support beam. Vortexes were observed around side entry orifice by PIV measurement. Concerning differential pressure, it is necessary to examine correlation between complicated flow structure and differential pressure. Thus in the present study, the differential pressure distribution of constructed model is experimentally investigated.Copyright