KAMEYAMA Takanori

教授
学位：博士（工学）

基本情報

所属

Undergraduate School of Engineering / Department of Applied Chemistry
Undergraduate School of Engineering / Department of Nuclear Engineering
Graduate School of Science and Technology / Course of Science and Technology
Graduate School of Engineering / Course of Applied Science

詳細情報

研究分野

Energy Nuclear engineering nuclear reactor physics, nuclear reactor engineering, nuclear fuel engineering

受賞

Atomic energy society of Japan JNST most cited article award Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

論文

BASIC CONCEPT OF INHERENT SAFETY FREE-SCALE REACTOR "KAMADO-FSR"

New Concept of Accident Tolerant Fuel Assembly Composed of SiC Block for KAMADO-BWR

Development of Educational Code ”S-Decay” on Nuclides Generation and Depletion

Development of a Compact and Convenient Neutron Diffusion Calculation Code ”S-Dif” for Education and Training

Development of an Educational Monte Carlo Method Code ”S-Monte” for Photon and Neutron Transport Calculation

Verification and Utilization of Tokai University Reactor Simulator (TURS) for Research and Education

Distributions of 242Cm and 244Cm Nuclide Compositions as Neutron Sources and Their Neutron Emission Rates in BWR Spent Fuels based on Three-Dimensional Neutron Transport and Burnup Calculations

Analysis of atomic distribution in as-fabricated Zircaloy-2 claddings by atom probe tomography under high-energy pulsed laser

Observation of c-component dislocation structures formed in pure Zr and Zr-base alloy by self-ion accelerator irradiation (vol 422, pg 167, 2012)

Observation of c-component dislocation structures formed in pure Zr and Zr-base alloy by self-ion accelerator irradiation

Modeling of H(n,n) Recoil Proton Injection into LWR Fuel Cladding with Sequential Use of MCNP and SRIM Codes

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

A new passive safety FBR concept of "KAMADO" - Easy replacement from the existing light water reactor to FBR

Core concept of a passive-safety fast reactor "METAL-KAMADO" and reactivity coefficients

Comparison of kinetic parameters based on continuous energy Monte Carlo method and evaluated nuclear data libraries for reactivity of MOX fuel cores

Estimation of 6 groups of effective delayed neutron fraction based on continuous energy Monte Carlo method

Core performance of new concept passive-safety reactor "KAMADO" - Safety, burn-up and uranium resource problem

Proposal of direct calculation of kinetic parameters βeff and based on continuous energy monte carlo method

High Burnup Rim Project: (III) properties of rim-structured fuel

Temperature and fission rate effects on the rim structure formation in a UO2 fuel with a burnup of 7.9% FIMA

Analyses of burnup at plutonium spots in uranium-plutonium mixed oxide fuels in light water reactors by neutron transport and burnup calculations

The FLEXBURN neutron transport code developed by the Sn method with transmission probabilities in arbitrary square meshes for light water reactor fuel assemblies

Dissolution behavior of highly burnt fuel

Numerical analysis for microstructure change of a light water reactor fuel pellet at high burnup

Concerning the microstructure changes that occur at the surface of UO2 pellets on irradiation to high burnup

Development of expert system on personal computer for diagnosis of nuclear reactor malfunctions

講演・口頭発表等

Development of particle transport calculation method based on analytical collision probability
Analyses of core damage proceeding during BWR station black-out with severe accident reactor simulator SARS
Quantification of temperature calculation uncertainty induced by thermal conductivity errors in LWR fuels
An inherent safety gas-cooled fast reactor concept of KAMADO-FR2 Design of ultra-long life core
3-dimensional measurement of neutron flux distribution in graphite moderator and single neutron source with small size Eu:LiCaAlF6 scintillation detector
Interaction analyses and validations of temperature and neutronics in Gd-doped fuel
Analyses of ECCS injection patterns in case of LOCA and SBO with LWR simulator SARS
Making a few groups of reactor constants in critical light-water reactor for neutron diffusion calculation code S-Dif
Measurement of neutron flux distribution in graphite moderator and single neutron source with the Eu:LiCaAlF6 scintillation detector
BWR fuel assembly ′′KAMADO-BWR′′ using SiC block with high accident resistance Preliminary study of basic concept, temperature distribution and critical heat flux
Verification of nuclear constant set for light water critical reactor in educational neutron diffusion calculation code S-Dif
Computer experiments for quantifying Pu inventory in spent LWR fuels by gamma ray measurement
An inherent safety gas-cooled fast reactor concept of “Eco-KAMADO”: Reactivity change due to core inundation on accident
Severe accidents analyses of LWR plant behavior with simulator of RELAP/SCDAPSIM code and graphical interface. (5) Damage level analysis of BWR core at station blackout.
Development of Nuclides Generation and Depletion Code ′′S-Decay′′ for Education
Measurement and verification of neutron flux from a single neutron source in graphite moderator by using the small Eu:LiCaAlF6 scintillation detector
Neutron scattering measurement with small detector and transport analysis considering thermal neutron scattering law in neutron moderator region
Gd-doped fuel behavior analysis modifying densification model and taking account of power change during low burnup
A User-friendly radiation dose evaluation Monte Carlo code for education / on-site use (3) Evaluation of radiation source position and intensity estimation functions
Gd-doped fuel behavior analysis modifying densification model and taking account of power change during low burnup