A Concept of Passive Safety Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal

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概要

• 論文の詳細を見る
• 1995-09-25

著者

• Okumura Keisuke

  Japan Atomic Energy Agency

• IWAMURA Takamichi

  Japan Atomic Energy Research Institute

• 奥村 啓介

  日本原子力研究開発機構 原子力基礎工学研究部門

• Okumura Keisuke

  Department Of Nuclear Energy System Japan Atomic Energy Research Institute

• Okumura Keisuke

  Japan Atomic Energy Research Institute

• MURAO Yoshio

  Japan Atomic Energy Research Institute

• ARAYA Fumimasa

  Japan Atomic Energy Research Institute

• Murao Y

  Japan Atomic Energy Res. Inst.

• Okumura Keisuke

  Nuclear Science And Engineering Directorate Japan Atomic Energy Agency

• Iwamura T

  Japan Atomic Energy Res. Inst.

• Iwamura Takamichi

  Japan Atomic Energy Agency

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• Sensitivity Analysis of Fission Product Concentrations for Light Water Reactor Burned Fuel
• JENDL-4.0 benchmarking for fission reactor applications
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• 炉物理における数学的方法および高性能計算に関する合同国際会議
• Transport Mechanism of Thermohydraulic Instability in Natural Circulation Boiling Water Reactors during Startup
• Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes
• ICONE11-36099 CRITICAL POWER CHARACTERISTICS OF TIGHT LATTICE BUNDLES
• A Concept of Passive Safety Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal
• ICONE11-36176 CONTAINMENT PRESSURE SUPPRESSION SYSTEM WITH FUNCTIONS OF WATER INJECTION AND NONCONDENSABLE GAS CONFINEMENT
• Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment
• Conceptual Design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its Recycle Characteristics
• JENDL-4.0 Benchmarking for Fission Reactor Applications
• Applicability of REFLA/TRAC Code to a Small-Break LOCA of PWR
• Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor(JPSR)
• Transient Analysis for Design of Primary Coolant Pump Adopted to JAERI Passive Safety Reactor JPSR
• Possibility of a Pressurized Water Reactor Concept with Highly Inherent Heat Removal Following Capability
• Determination of ^Se and ^Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes
• Evaluation on Driving Force of Natural Circulation in Downcomer for Passive Residual Heat Removal System in JAERI Passive Safety Reactor JPSR
• Analyses of Assay Data of LWR Spent Nuclear Fuels with a Continuous-Energy Monte Carlo Code MVP and JENDL-4.0 for Inventory Estimation of Se, Tc, Sn and Cs (Selected Papers of the Joint International Conference of Supercomputing in Nuclear Applications an
• Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data.
• Experimental study of ECC water injection rate effects on reflood phase of PWR-LOCA.
• Experimental study of initial downcomer water accumulation velocity effects on reflood phase of PWR-LOCA.
• Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA.
• Experimental study of system behavior during reflood phase of PWR-LOCA using CCTF.
• Effect of loop seal on reflood phenomena in PWR.
• CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution.
• Experimental Study on Difference in Reflood Core Heat Transfer among CCTF, FLECHT-SET and Predicted with FLECHT Correlation.
• Burnout characteristics under flow reduction condition.
• Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data.
• System pressure effect on system and core cooling behavior during reflood phase of PWR LOCA.
• Initial thermal-hydraulic behaviors under simultaneous ECC water injection into cold leg and upper plenum in a PWR-LOCA.
• Air-Water Two-Phase Cross Flow Resistance in Rod Bundle
• Experimental study of two-dimensional thermal-hydraulic behavior in core during reflood phase of PWR LOCA.
• Water accumulation phenomena in upper plenum during reflood phase of PWR-LOCA by using CCTF data.
• Experimental study on reflood behavior in PWR with upper plenum injection type ECCS by using CCTF.
• Visual study of flow behavior in upper plenum during simulated reflood phase of PWR-LOCA.
• Effect of decay heat level on reflood phenomena during PWR-LOCA.
• Assessment of core radial power profile effect model for REFLA code by using CCTF data.
• Pressure drop through broken cold leg during reflood phase of loss-of-coolant accident of pressurized water reactor.
• Applicability of Core Thermal-Hydraulic Models in REFLA Code to 17*17 Type Fuel Assembly of PWR.
• Elimination of Numerical Pressure Spikes Induced by Two-Fluid Model.

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