| Autor: |
Nobuchika Kawasaki, Kenichiro Satoh, Naoto Kasahara, Kazuyuki Tsukimori |
| Rok vydání: |
2010 |
| Předmět: |
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| Zdroj: |
ASME 2010 Pressure Vessels and Piping Conference: Volume 1. |
| DOI: |
10.1115/pvp2010-25414 |
| Popis: |
Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induce incremental deformation and creep fatigue crack at critical portions around the sodium surface, thermal stratification layer and core support structures. Those phenomena are so complex that design evaluation becomes sometimes too conservative. In order to achieve precise high temperature design for realizing compact reactor vessels of fast reactor plants, such guidelines are proposed as for thermal load modeling, structural analysis and strength evaluation. This paper gives the brief summary of these guidelines. GUIDELINES FOR THERMAL LOAD MODELING: One of main difficulties of thermal load modeling is their inducement mechanism by interaction between thermal hydraulic and structural mechanics. Design evaluation requires envelope load conditions with considering scatter of design parameters. Proposed guidelines enable precise load modeling by grasping sensitivities of thermal stress to design parameters including thermal hydraulic ones. GUIDELINES FOR INELASTIC DESIGN ANALYSIS: Guidelines are proposed to apply inelastic analysis methods for design of reactor vessels. There are so many influence parameters in inelastic analysis that conservative and unique solutions are hardly found. To overcome such difficulties, mechanism and main parameters of inelastic behaviors of reactor vessels were clarified. Guidelines give conservative results within the same mechanism as expected reactor vessels. HIGH TEMPERATURE STRENGTH EVALUATION METHOD: Incremental deformation and creep fatigue strength evaluation methods were proposed. Accumulated strain is limited within no influence of fatigue and creep-fatigue strength. Taking design conditions of reactor vessels into account, creep fatigue evaluation considers strain concentration and an intermediate stress hold effect on creep-fatigue strength. Influences of thermal aging were also confirmed.Copyright © 2010 by ASME |
| Databáze: |
OpenAIRE |
| Externí odkaz: |
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