Recruitment for Postdoctoral Fellow
(Fixed-term researcher)

1. Physics	Chemistry	Mathematics	Geo and Environemtal Sciences	Biology	Radiation
   Mechanics	Material	Electricity and Electronics	Architectural and Civil Engineering	Applied Physics	Applied Chemistry
   Chemical Engineering	Measurements and Instruments	Computer and Information	Robotics	Other	All

2. No	Theme
   	Department	Section	Location	Contact Person	Radiation Worker/ Non-Radiation Worker	Field (for reference）
   	Summary
   J2	R&D for Laser Processing Simulation Code with Laser Processing Experiment and Deployment of Sensing Technology
   	Tsuruga Center for International and Regional Collaboration	Applied Laser Technology Institute	Tsuruga Head Office	Toshiharu Muramatsu E-mail: muramatsu.toshiharu@jaea.go.jp	Non-Radiation Worker	Mechanics Material Applied Physics Computer and Information Measurements and Instruments Material
   	We are developing laser processing technologies for industrial applications. First, the development of simulation code named "SPLICE" should be advanced to improve the numerical turbulent model and to optimize the parameters. High power laser processing experiment is needed for quantitative analysis. New laser systems and several sensing technologies are required for structural health monitoring with multiple sensors. Heat resistant optical fiber sensor has been installed on the piping system of demo-plant. Data remote processing is needed for monitoring the integrity. ※Energy base-ization of research and development Fukui-ken advances, they're the research and development concerned with advance of laser improvement process simulation code SPLICE developed aiming at contribution to a plan, and I make them promote more prefecture use of SPLICE cable through an outcome of this case. All together, I aim at preservation optics in atomic energy and innovation of status monitoring.
   J3	Study on feasibility and effectiveness evaluation for severe accident countermeasures
   	Nuclear Safety Research Center	Severe Accident Analysis Research Group	Tokai Research and Development Center Nuclear Science Research Institute	Tomoyuki Sugiyama Tel:+81-29-282-5253 E-mail: sugiyama.tomoyuki@jaea.go.jp	Non-Radiation Worker	Physics Chemistry Mechanics Material Applied Physics Computer and Information Other
   	This research aims at development of analysis models and tools to improve evaluation techniques of severe accident countermeasures. One of the following tasks or that related to the tasks is carried out. - Source term analysis of Fukushima daiichi NPS accident using the SA analysis code THALES2/KICHE. - Analysis of fluid dynamic behaviors of core melt in containment vessel using the mechanistic FCI code JASMINE. - Analysis of thermal-hydraulic and deflagration/detonation behaviors of hydrogen in containment vessel or reactor building using the open CFD code OpenFOAM.
   J4	Study on the methodology of the structural integrity assessment for nuclear reactor components
   	Nuclear Safety Research Center	Structural Integrity Research Group	Tokai Research and Development Center Nuclear Science Research Institute	Yinsheng Li Tel:+81-29-282-6457 E-mail: li.yinsheng@jaea.go.jp	Non-Radiation Worker	Mechanics Material Applied Physics Physics Architectural and Civil Engineering Computer and Information
   	Because of the ageing degradation due to long term operation for Japanese nuclear power plants, developing the methodologies of structural integrity assessments for the reactor components concerning neutron irradiation, stress corrosion cracking and so on is of great importance. In this theme, researches on the deterministic approaches such as weld residual stress evaluation, crack propagation evaluation under large scale yielding condition, fracture evaluation concerning the crack or thinning for nuclear components are conducted on the basis of numerical simulation, material testing, and fracture testing and so on. In addition, probabilistic fracture mechanics analysis codes concerning ageing degradation of nuclear components are developed on the basis of the knowledge obtained from simulation and testing.
   J5	Study on aging degradation of nuclear reactor structural materials under irradiation
   	Nuclear Safety Research Center	Materials and Water Chemistry Research Group	Tokai Research and Development Center Nuclear Science Research Institute	Yutaka Nishiyama Tel:+81-29-282-5044 E-mail: nishiyama.yutaka93@jaea.go.jp	Radiation Worker	Mechanics Material Measurements and Instruments
   	In the nuclear reactor structural materials used under irradiation, the material properties, the interface reactions with the coolant and the stress conditions etc. change simultaneously. They are important phenomena to evaluate the aging degradation of light water reactors. In this study, for ferritic steels and austenitic stainless steels used as the nuclear reactor structural materials, the changes in material and mechanical properties (such as microstructure, crack growth and fracture toughness etc.) induced by irradiation are investigated. From these results, the effects on the structural integrity of the reactor pressure vessels and the core internals are evaluated.
   J6	Experimental and analytical study on thermohydraulic safety of the light water reactor
   	Nuclear Safety Research Center	Thermohydraulic Safety Research Group	Tokai Research and Development Center Nuclear Science Research Institute	Taisuke Yonomoto Tel:+81-29-282-5263 E-mail: yonomoto.taisuke@jaea.go.jp	Non-Radiation Worker	Mechanics Measurements and Instruments Computer and Information
   	This experimental and analytical research focuses on thermo-hydraulic phenomena occurring in the reactor and the containment of the nuclear power plant during an accident before and after core damage. For the experimental study, two-phase flow and/or heat transfer are investigated using a high-pressure reactor simulation test facility or a small-scale test device that exits or will be built for this research. The development of the two-phase flow measurement technique is also an important topic for this research. By using the data obtained from the experiments, prediction models are validated and improved in order to be used in lumped parameter codes such as RELAP5 and MELCOR, or the CFD codes. A specific research topic will be selected considering the request by the applicant.
   J7	Study on high-temperature oxidation behavior of fuel cladding
   	Nuclear Safety Research Center	Fuel Safety Research Group	Tokai Research and Development Center Nuclear Science Research Institute	Masaki Amaya Tel:+81-29-282-5028 E-mail: amaya.masaki@jaea.go.jp	Non-Radiation Worker	Material Mechanics
   	It has been reported that a oxidation rate of light-water-reactor fuel cladding might rapidly increase when the fuel cladding was exposed to high-temperature steam for a long period, which is so-called "breakaway oxidation". In order to evaluate the embrittlement of fuel cladding under a loss-of-coolant accident (LOCA), it is important to consider the breakaway oxidation. However, sufficient information has not been obtained about the initiation condition of the breakaway oxidation and the effect of the breakaway oxidation on the mechanical strength of fuel cladding. In this study, the change of the high-temperature oxidation kinetics of fuel cladding, which may lead to the breakaway oxidation, will be investigated by conducting high-temperature oxidation tests of fuel cladding specimens under simulated LOCA conditions. Mechanical tests will be also conducted on the oxidized specimens in order to evaluate the effect of the change in high-temperature oxidation behavior on mechanical strength of fuel cladding.
   J25	Evaluation of correlation between changes in microstructure and mechanical properties in irradiated materials (metal, steel, and ceramics)
   	Nuclear Science and Engineering Center	Research Group for Radiation Materials Engineering, Fuels and Amterials Engineerign Division	Tokai Research and Development Center Nuclear Science Research Institute	Shinichiro Yamashita Tel:+81-29-282-5391 E-mail: yamashita.shinichiro@jaea.go.jp	Radiation Worker	Radiation Material Physics Mechanics
   	In order to increase safety and integrity in existing and future nuclear power plants, micrstructural observation and a wide variety of mechanical strength tests (tensile, hardness measurement, and toughness etc) of nuclear reactor component materials such as structural material and simulated fuel-like oxide irradiated at various environmental conditions will be performed. Based on the experimental data acquired, a correlation between changes in microstructure and mechanical properties in the materials will be evaluated. In addition to that, it is possible to conduct fundamental study on radiation damage for the materials. Through these works, it is expected that synergistic function among environmental factors (irradiation, thermal load, stress, atmosphere etc) influencing on correlation evaluation for those materials will be clarified, contributing increment of safety and integrity in the existing/future nuclear power plants.
   J28	Research and Development of Evaluation Method for Core Degradation and Release of Radioactive Materials at LWR Accident
   	Nuclear Science and Engineering Center	Develppment Group for Thermal-Hydraulics Technology, LWR Key Technology Development Division	Tokai Research and Development Center Nuclear Science Research Institute	Hiroyuki Yoshida Tel:+81-29-282-5275 E-mail: yoshida.hiroyuki@jaea.go.jp	Radiation Worker	Physics Chemistry Material Mechanics
   	A coupled analysis method of thermal-hydraulics and chemical reaction is developed to evaluate core degradation and release of radioactive materials in Light Water Reactor (LWR) accidents in this study. In detail, numerical methods and models simulating thermal-hydraulic behavior with chemical reactions are developed to solve melting of nuclear fuel and metals, release of radioactive materials and relocation of molten and released materials. In addition, new thermal-hydraulic and chemical reactions experiments are performed to validate developed methods and models. By developing this method, basic knowledge to understand multi-physics phenomena at LWR accidents is obtained, and numerical models for severe accident analysis codes are improved.
   J29	Research of fatigue evaluation method of target vessel for high-power spallation neutron source
   	Sector of Nuclear Science Research, J-PARC Center	Materials and Life Science Division Neutron Source Section	J-PARC	Eiichi Wakai Tel:+81-29-284-3745 E-mail: wakai.eiichi@jaea.go.jp	Radiation Worker	Material Mechanics Measurements and Instruments
   	In Materials and Life Science Experimental Facility (MLF) of J-PARC, it is required to develop the spallation neutron source for the stable operation at 1 MW proton beam power, and various researches of the mercury target vessel are performing by the quantitative evaluation method. In this theme, there are some research subjects as follows: (1) Study of the fatigue phenomenon and thermal stress in the mercury vessel induced by the pulsed proton beam, (2) The experimental technology R&D including the fatigue properties and the testing to improve and advance the evaluation method of life time estimation of the target vessel, (3) The related systematic studies of the spallation neutron source and the target vessel for the stable operation at 1 MW.
   J35	Study on in-vessel thermal-hydraulic behavior during the loss of forced cooling accidents in HTTR
   	Sector of Nuclear Science Research Oarai Research and Development Center Department of HTTR	HTTR Reactor Engineering Section	O-arai Research and Development Center	Daisuke Tochio Tel:029-267-1919-3730 E-mail: tochio.daisuke@jaea.go.jp	Non-Radiation Worker	Mechanics Computer and Information
   	The Loss Of Forced Cooling (LOFC) Test Plan is in progress in the High Temperature Engineering Test Reactor (HTTR) under the international collaboration. The thermal-hydraulic behavior of fuel, structure and coolant is important during LOFC. In this study,thermal-hydraulic analysis model is constructed and numerical analysis is performed using commercial CFD analysis code to reveal the in-vessel thermal-hydraulic behavior during the LOFC.
   F3	Research on severe accident progression behavior of the Fukushima Daiichi NPP accident
   	Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Sector of Fukushima Research and Development	Molten Core Behavior Analysis Group Severe Accident Propagation Behavior Evaluation Division	Tokai Research and Development Center (Nuclear Science Research Institute)	Toshio Nakagiri Tel:+81-(0)29-267-1919, Ex.5802 E-mail: nakagiri.toshio@jaea.go.jp	Non-Radiation Worker	Physics Chemistry Radiation Mechanics Material Measurements and Instruments Computer and Information
   	Analytical evaluation using SA codes (SCDAP, MELCOR, etc.) and 1F plant data are peroformed to enhance understanding on accident progression behavior in the Fukushima Daiichi NPP. In this theme, evaluation of 1F plant data with SA code analysis will be conducted and possibility of model improvement will be considered where appropriate.
   F4	Research on the degradation behavior of fuel assemblies in Fukushima Daiichi NPP accident condition
   	Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Sector of Fukushima Research and Development	Molten Core Behavior Analysis Group Severe Accident Propagation Behavior Evaluation Division	Tokai Research and Development Center (Nuclear Science Research Institute)	Toshio Nakagiri Tel:+81-(0)29-267-1919, Ex.5802 E-mail: nakagiri.toshio@jaea.go.jp	Non-Radiation Worker	Physics Chemistry Mechanics Material Measurements and Instruments
   	Present knowledge on reaction behavior of the fuel assembly materials (control blade, fuel rod, channel box) is insufficient to understand degradation behavior of fuel asseblies in the Fukushima Daiichi NPP accident. In this theme, laboratory scale experiments on reaction behavior of fuel assembly materials, large scale experiments on the degradation behavior of fuel assemblies and development of individual reaction models will be performed.
   F5	Development of Technology for Working Environment Data Collection and Accumulation by Remote Operated Robots
   	Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Sector of Fukushima Research and Development	Remote Technology and Robotics Group, Remote System and Sensing Technology Division	Tokai Research and Development Center (Nuclear Science Research Institute) or Tomioka International Collaborative Research Building (now under constructing)	Dr. Tatsuo TORII Tel:+81-(0)29-282-6329 E-mail: torii.tatsuo@jaea.go.jp	Non-Radiation Worker	Computer and Information Robotics Electricity and Electronics Mechanics
   	For long-term decommissioning process of the Fukushima Daiichi NPP, it is importatnt to gather and accumulate the data of the working environment as the reference to plan the missons and to use for the workers' training. The objective of this reseach subject is to develp the map building and working enviromental modeling method based on collected sensory data by remote operated robots and sensing systems. Design and implementation of spatio-temporal database to register the working environment data is also included in this research project.
   F15	Investigation of the characteritics of radioactive material in a severe accident condition
   	Nuclear Science and Engineering Center	Research Group for LWR Advanced Technology LWR Key Technology Development Division	Tokai Research and Development Cente (Nuclear Science Research Institute)	Masahiko Osaka Tel:+81-29-267-4141 E-mail: ohsaka.masahiko@jaea.go.jp	Radiation Worker	Physics Chemistry Material Mechanics
   	In order to acquire fundamental knowledge on fission product (FP) behavior and dose evaluation for the decommissioning of Fukushima Daiich Nuclear Power Station, we are investigating the behavior of radioactive material in the Primary Containment Vessel (PCV) under severe accident conditions. The post-doctoral fellow will investigate the characteristics of such radioactive material and aerosols (chemical form, size distribution, etc) by conducting simulation tests. These tests will simulate FP transport from core to PCV. The aerosols analyses on size distribution will be conducted in the different transport stages, and will be coupled with post characterizations of the deposits (microstructure and chemical state). The post-doctoral fellow will be also involved in the evaluation of the radioactive material behavior and aerosol formation mechanism, by performing analytical studies on the chemical reaction kinetics and by comparing the simulation tests with the analysis of environmental samples.