Masatsugu Akiyama

Two-Parameter Thermal Lattice BGK Model with a Controllable Prandtl Number

In this paper, two time relaxation parameters are introduce to a thermal lattice BGK, model to make its Prandtl number controllable. The dependency of the Prandtl number on the two parameters is derived. Numerical measurement of the transport coefficients is used to demonstrate the validity of the method. Furthermore, two examples of convective heat transfer are calculated, with one to show the effectiveness, and the other to show the breakdown of the two-parameter formulation under different conditions.

Heat Transfer in Lattice BGK Modeled Fluid

The thermal lattice BGK model is a recently suggested numerical tool aiming at solving problems of thermohydrodynamics. The quality of the lattice BGK simulation is checked in this paper by calculating temperature profiles in the Couette flow under different Eckert and Mach numbers. A revised lower order model is proposed to improve the accuracy and the higher order model is proved to be advantageous in this respect, especially in the flow regime with a higher Mach number.

Prandtl number of lattice Bhatnagar–Gross–Krook fluid

The lattice Bhatnagar–Gross–Krook modeled fluid has an unchangeable unit Prandtl number. A simple method is introduced in this report to formulate a flexible Prandtl number for the modeled fluid. The effectiveness of this method was demonstrated by numerical simulation of the Couette flow.

A conceptual design of light ion beam fusion reactor—UTLIF(1)

Abstract UTLIF(1) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of pin-bundle blanket. The study includes nuclear and structural analyses of the blanket, consideration on materials, tritium handling system and power conversion system designs, pellet and beam driver designs, and economic analysis of the plant. The pin-bundle blanket has been shown to be attractive for light ion beam fusion reactors. Some subjects to be developed have been pointed out from reactor engineering aspects.

Subchannel analysis of a critical power test, using simulated BWR S × S fuel assembly

Abstract Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8 × 8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F ‘multi-fluid’ computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, through comparisons, between the prediction and the obtained test data.

Neutron streaming through a slit and duct in concrete shields and comparison with a Monte Carlo analysis

A series of measurements of about14-MeV deuterium-tritium neutrons streaming through a slit and a duct in concrete shields has been carried out using a Cockcroft-Walton-type neutron generator. Measured neutron energy spectra are compared with calculations in six configurations of the shields. The configurations are the simplified geometries of streaming paths of tokamak reactors, such as a divertor throat and a neutral beam injection port. The measured data were obtained with an NE-213 liquid scintillator using pulse shape discrimination methods to resolve neutron and gamma-ray pulse height data and using a spectral unfolding code to convert these data to energy spectra. The experiments were analyzed by a Monte Carlo code. The calculated neutron energy spectra slightly underestimate the measured data, especially in the range of 6 to 8 MeV. The agreement between the calculated and measured integral flux above 2.2 MeV ranges from 87.5 to 72.% depending on the configurations.

Study of epithermal neutron columns for boron neutron capture therapy

Abstract Optimal neutron energy for boron neutron capture therapy for treating brain tumors was studied. Epithermal neutrons gave lower dose at the brain surface and pentrated deeper than thermal neutrons. An epithermal neutron column and a reactor facility for boron neutron capture therapy were conceptually studied. Layers of alminum and heavy water whose volume ratio 85 15 were effective in decreasing fast neutrons while maintaining a high epithermal neutron level. An epithermal neutron column was experimentally developed at YAYOI reactor. It has been used for fundamental study of BNCT and radiation biology, although the beam intensity was not high enough to treat the patients due to the low reactor power.

Measurement and analysis of reaction rates in a large spherical depleted uranium pile surrounding a 14-MeV neutron source

Abstract Measurement was made of the reaction rate distributions for 238U(n, f), 235U(n, f), 238U(n, γ), 27Al(n, α) and 58Ni(n, p) in a large depleted uranium (DU) pile. The pile consisted of DU blocks forming a spherical shell of 45.72 cm radius and 40.64 cm thick. 14-MeV neutrons were generated at the center. Fast neutron leakage spectrum was also measured by an NE-213 spectrometer. In order to assess the 238U neutron cross sections of JENDL-2, the experiment was analyzed using the Monte-Carlo transport code MCNP with continuous energy cross sections. The agreement between the calculation and the experiment is generally unsatisfactory. The ratios of calculation to experiment of low energy reactions decreased with the thickness of the DU layer. The analysis of the Livermore pulsed sphere experiment for the small DU sphere revealed underestimation of leakage neutron spectrum around 10 MeV. The 238U cross sections of JENDL-2 should be improved for 14-MeV neutronics calculation.

A conceptual design of light ion beam fusion reactor — UTLIF(2)

Abstract UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear, thermal and structural analyses have been made and the performance of the reactor has been shown to be satisfactory. Safety and reliability of the plant have been discussed and the challenge before us has been revealed. This paper also includes a fireball analysis, an outline of target design and design of a tritium handling system.

Simulation of shock-interface interaction using a lattice Boltzmann model

Abstract For simulation of fundamental processes of vapor explosion, a novel fluid dynamic simulation method is developed using the lattice Boltzmann model. We incorporate into the model capabilities to deal with compressible fluid flow and two-phase flow. By numerical simulation, we demonstrate fundamental characteristics of this model and apply it to simulation of interaction between shock wave and two-phase interfaces.