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The Reliability Prediction of HTR’s Graphite Component in Various Temperature and Neutron Dose Levels

[+] Author Affiliations
Xiang Fang, Haitao Wang, Xingtuan Yang, Suyuan Yu

Tsinghua University, Beijing, China

Paper No. ICONE21-15574, pp. V002T03A028; 5 pages
  • 2013 21st International Conference on Nuclear Engineering
  • Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors
  • Chengdu, China, July 29–August 2, 2013
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5579-9
  • Copyright © 2013 by ASME


In high temperature gas-cooled reactors (HTRs), graphite is used as the main structure material. The side reflecter of the reactor core is composed by a pile of graphite bricks. In real operational condition of the reactor, both high temperature and fast neutron irradiation have great effect on the behavior of graphite components. The non-uniform distribution of temperature and neutron dose cause obvious stress accumulation, which greatly affects the security and reliability of the graphite components. In addition, high temperature and neutron irradiation make the properties of graphite change in evidence, and the changes are not linear. Such changes must be considered and simulated in the calculation, in order to predict the stress concentration condition and the reliability of the graphite brick correctly. A FORTRAN code based on user subroutines of MSC.MARC is developed in INET in order to perform three-dimensional finite element analysis of irradiated behavior of the graphite components for the HTRs. In this paper, the stress level and failure probability of graphite components are calculated and obtained under different in-core temperatures and neutron dose levels of the core side of brick. 400°C, 500°C, 600°C and 700°C are selected as the core side temperature, while the range of neutron dose is 0 to 1022n cm-2 (EDN). Different constitutive laws are used in stress analysis procedure. The impact of different temperature and neutron dose levels are discussed.

Copyright © 2013 by ASME



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