
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
ResearchMapへ移動します
Inquiries about coverage or research
Inquiries about coverage
Public Affairs Division Public Affairs and Communications Department
Tel. 0463-63-4670(direct dialing)