Akira Tokuhiro
PhD

Simulation-based Stochastic Optimization of a Dry, Used Nuclear Fuel Storage Array for Nuclear Criticality Safety

International Conference on Nuclear Criticality Safety

Branch-and-Bound Algorithm Applied to Dynamic PRA with Uncertainty Quantification for LWR Station Blackout

23rd International Conference on Nuclear Engineering, ICONE-23

CFD Analysis of the ¼-Scale Air Reactor Cavity Cooling System with Focus on the Inlet Plenum

23rd International Conference on Nuclear Engineering, ICONE-23

Characterizing Single- to Two-Phase Flow Behavior in a Crosscutting RCCS Design with Decay Heat Variation

23rd International Conference on Nuclear Engineering, ICONE-23

Emerging Regulatory Requirements and Industry Practices on Emergency Preparedness in the Post-Fukushima Operations Era

23rd International Conference on Nuclear Engineering, ICONE-23

Lesson Learnt from Fukushima

5th Annual Nuclear Supply Chain Conference

Supply Chain Management in the Global Market, Bridging the Divide Between Vendor, Utility and Supplier

Nuclear Exchange 2013

The Realities and Consequences of the Post-Quake and Tsunami Fukushima NPP Incidents

Nuclear Exchange 2011

Production of Energetic Light Fragments with Expanded Cascade Exciton Model (CEM)

2014 ANS Annual Meeting

Is the Food in Safe in Japan?

American Chamber of Commerce Japan Conference, Meeting on Food Safety (Post-Fukushima)

Branch-and-Bound Algorithm Applied to Uncertainty Quantification of a Boiling Water Reactor Station Blackout

Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost-effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, the large uncertainty in a particular aspect of the model may or may not have a large impact on the final results.

View more - Branch-and-Bound Algorithm Applied to Uncertainty Quantification of a Boiling Water Reactor Station Blackout

Optimization Method to Branch-and-Bound Large SBO State Spaces Under Dynamic Probabilistic Risk Assessment via Use of LENDIT Scales and S2R2 Sets

Traditional probabilistic risk assessment (PRA) methods have been developed to evaluate the risk associated with complex systems; however, PRA methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems.

View more - Optimization Method to Branch-and-Bound Large SBO State Spaces Under Dynamic Probabilistic Risk Assessment via Use of LENDIT Scales and S2R2 Sets

Comparison Between Single- and Two-Phase Heat Transfer in a Thermally Stratified Enclosure

The ratio of local Nusselt number of two-phase natural convection, Nuy, two, to local Nusselt, Nuy, single, number of single-phase natural convection was empirically determined. Gelatin particles were immersed in a custom-built rectangular enclosure. With one end maintained at a thermal high, the particles settling and thermal enhancement characteristics were analyzed.

View more - Comparison Between Single- and Two-Phase Heat Transfer in a Thermally Stratified Enclosure

Natural Convection Heat Transfer of Hydrophilic Particle Suspension: Implications on Nuclear Waste Remediation

Natural convection of gel suspension has the potential to be applied to particulate recovery systems, such as in the dynamic filtering of radioactive waste particles. In this study, experiments were conducted to assess natural convection of gradually settling gelatin particle suspension in a rectangular cell, relative to single-phase natural convection with bulk temperature stratification, and to derive empirically a transfer relationship.

View more - Natural Convection Heat Transfer of Hydrophilic Particle Suspension: Implications on Nuclear Waste Remediation

Effects of Adverse and Favorable Pressure Gradients on Entropy Generation in a Transitional Boundary Layer Region Under the Influence of Freestream Turbulence

Fundamental understanding of the entropy generation process in characteristic wall shear flows is developed. For entropy generated by fluid friction, the rates are predictable for developed turbulent flows and pure laminar flows. The main concern lies in prediction for flows undergoing so-called “bypass” transition from laminar to a turbulent state. In this study, the entropy generation process in the bypass transition scenario is investigated for a flat plate boundary layer under the effects of adverse and favourable pressure gradients.

View more - Effects of Adverse and Favorable Pressure Gradients on Entropy Generation in a Transitional Boundary Layer Region Under the Influence of Freestream Turbulence

Morphological Analysis of Zirconium Nuclear Fuel Retaining Rods Braided with SiC: Quality Assurance and Defect Identification

In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys' thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ~50,000 individual filaments of 5 – 10 µm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament.

View more - Morphological Analysis of Zirconium Nuclear Fuel Retaining Rods Braided with SiC: Quality Assurance and Defect Identification

Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

Stability of materials under extreme conditions is an important issue for safety of nuclear reactors. Presently, silicon carbide (SiC) is being studied as a cladding material candidate for fuel rods in boiling-water and pressurized water-cooled reactors (BWRs and PWRs) that would substitute or modify traditional zircaloy materials. The rate of corrosion of the SiC ceramics in a hot vapour environment (up to 2200 °C) simulating emergency conditions of light water reactor (LWR) depends on many environmental factors such as pressure, temperature, viscosity, and surface quality.

View more - Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

Oxidation and Hydrogen Uptake in Zirconium, Zircaloy-2 and Zircaloy-4: Computational Thermodynamics and Ab Initio Calculations

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for BWRs and PWRs, respectively. Over more than 60 years these materials displayed a very good combination of properties such as low neutron absorption, creep behaviour, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion, and/or oxidation, especially in the case of Zircaloy-4 [1–3]. However, over the last couple of years, energetic efforts were undertaken to improve their oxidation resistance during off-normal temperature excursions, as well as to further improve upon the already achieved levels of mechanical behaviour and reduced hydrogen uptake [1–3].

View more - Oxidation and Hydrogen Uptake in Zirconium, Zircaloy-2 and Zircaloy-4: Computational Thermodynamics and Ab Initio Calculations

Influences of Boil-Off on the Behavior of a Two-Phase Natural Circulation Loop

This experimental study investigates the thermal-hydraulic behaviour and boiling mechanisms present in a natural circulation loop experiencing boil-off. Extended duration tests with inventory loss were conducted to examine the transient behaviour of a test loop under constant heat load.

View more - Influences of Boil-Off on the Behavior of a Two-Phase Natural Circulation Loop

Computational and Experimental Prediction of Dust Production in Pebble Bed Reactors, Part II

This paper is the continuation of Part I, which describes the high temperature and high-pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment.

View more - Computational and Experimental Prediction of Dust Production in Pebble Bed Reactors, Part II

Donald Crawford Graduate Faculty Mentoring Award

Dr. Tokuhiro was honoured for his efforts to mentor graduate students.

American Nuclear Society