スポンサーリンク
Nuclear Safety Research Center Japan Atomic Energy Agency | 論文
- Internal-Shear Mode Instabilities on High-Speed Liquid Jet, (II) : Experimental Analysis of Curved Target Flow
- Internal-Shear Mode Instabilities on High-Speed Liquid Jet, (I) : Characteristics of Linear Solutions
- Criticality Safety Benchmark Experiment on 10% Enriched Uranyl Nitrate Solution Using a 28-cm-Thickness Slab Core
- Effect of Cooling History on Cladding Ductility under LOCA Conditions
- ICONE11-36288 OPTICAL MEASUREMENT OF WAVES ON HIGH-SPEED LITHIUM JET SIMULATING IFMIF TARGET FLOW
- In-pile Experiment in JMTR on the Radiation Induced Surface Activation (RISA) Effect on Flow-boiling Heat Transfer
- Interfacial Friction Factor for High-Pressure Steam/Water Stratified-Wavy Flow in Horizontal Pipe
- Effect of Correlations of Component Failures and Cross-Connections of EDGs on Seismically Induced Core Damages of a Multi-Unit Site
- Core Heat Transfer Coefficients Immediately Downstream of the Rewetting Front during Anticipated Operational Occurrences for BWRs
- Benchmark Critical Experiments of a Heterogeneous System of Uranium Fuel Rods and Uranium Solution Poisoned with Gadolinium, and Application of Their Results to JACS Validation
- Fluctuation of the Neutron Multiplication Factor Induced by an Oscillation of the Fuel Solution System
- Development of a Statistical Method for Evaluation of Estimated Criticality Lower-Limit Multiplication Factor Depending on Uranium Enrichment and H/Uranium-235 Atomic Ratio
- Ring Compression Ductility of High-Burnup Fuel Cladding after Exposure to Simulated LOCA Conditions
- A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel (2) Feasibility study for the Fukushima Daiichi Nuclear Power Plant (Fukushima NPP Accident Related)
- Verification of FEMAXI-7 code by using irradiation test in Halden reactor for He-pressurization effect on FGR of BWR fuels under power transient
- A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel : II : Feasibility study for the Fukushima Daiichi Nuclear Power Plant
- A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel : I : Basic concepts and model descriptions for the Fukushima Daiichi Nuclear Power Plant
- RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg
- Effect of Higher-Harmonic Flux in Exponential Experiment for Subcriticality Measurement
- A Water Radiolysis Code for the Irradiation Loop System