Creep–Fatigue Assessment for Interaction Between IHX Seal Holder and Inner Vessel Stand Pipe in a Pool-Type Fast Reactor as Per RCC-MR

2017 ◽  
pp. 299-306
Author(s):  
Sanjay Kumar Pandey ◽  
S. Jalaldeen ◽  
K. Velusamy ◽  
P. Puthiyavinayagam
Keyword(s):  
Fast Reactor ◽  
Fatigue Assessment ◽  
Creep Fatigue ◽  
Inner Vessel ◽  
Pool Type

Development of high-chromium steel for the sodium-cooled fast reactor in Japan and creep–fatigue assessment of the steel

2008 ◽  
Vol 85 (6) ◽  
pp. 352-359 ◽  
Author(s):  
Takashi Wakai ◽  
Masayuki Sukekawa ◽  
Shingo Date ◽  
Tai Asayama ◽  
Kazumi Aoto ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Chromium Steel ◽  
Fatigue Assessment ◽  
High Chromium ◽  
Creep Fatigue

Assessment of Life of Inner Vessel for a Pool-Type Fast Breeder Reactor Based on Inelastic and Viscoplastic Deformations

1993 ◽  
Vol 115 (2) ◽  
pp. 185-192 ◽  
Author(s):  
R. S. Alwar ◽  
P. Chellapandi ◽  
S. B. Bhoje
Keyword(s):  
Fatigue Damage ◽  
Kinematic Hardening ◽  
Material Model ◽  
Design Stage ◽  
Benchmark Problems ◽  
Analysis Tool ◽  
Inelastic Analysis ◽  
Creep Fatigue ◽  
Inner Vessel ◽  
Pool Type

Towards assessing the creep-fatigue damage for the inner vessel of a pool-type fast reactor, 3-D elastic and axisymmetric inelastic analysis with classical kinematic hardening model, ORNL material model and more realistic CHABOCHE visco-plastic model have been performed. The material properties of stainless steel type 316 LN at various temperatures are used in the analysis. Using the results of this analysis, the creep-fatigue damage and ratcheting are assessed as per the rules of RCC-MR and results have been compared for the various constitutive models used. Inelastic analysis has been performed only at highly stressed local regions. Computations with the ORNL material model is found to be satisfactory in terms of conservatism and computer cost involved, and hence recommended for use in the design stage. Further, conservatism in estimating the creep fatigue-damage of the component associated with the elastic route of RCC-MR has been quantified. The inelastic and viscoplastic analysis tool has been validated using experimental benchmark problems.

Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 232-236 ◽  
Author(s):  
D. L. Zhang ◽  
P. Song ◽  
S. Wang ◽  
X. Wang ◽  
J. Chen ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Pool Type ◽  
Code Development

scholarly journals Manufacturing Technology Development of Single Torus Inner Vessel for Sodium Cooled Fast Reactor

2014 ◽  
Vol 86 ◽  
pp. 367-374 ◽  
Author(s):  
Gagan Gupta ◽  
Rosy Sarkar ◽  
R. Sritharan ◽  
V. Rajan Babu ◽  
V. Balasubramaniyan ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Technology Development ◽  
Manufacturing Technology ◽  
Inner Vessel

A practical methodology devoted to pool‐type phenomena simulation in safety analysis for sodium‐cooled fast reactor

2020 ◽  
Author(s):  
Shibao Wang ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Wenxi Tian
Keyword(s):  
Fast Reactor ◽  
Safety Analysis ◽  
Pool Type

A CONCEPTUAL DESIGN STUDY OF POOL-TYPE SODIUM-COOLED FAST REACTOR WITH ENHANCED ANTI-SEISMIC CAPABILITY

2019 ◽  
Vol 2019.27 (0) ◽  
pp. 1651
Author(s):  
Shigenobu Kubo ◽  
Yoshitaka Chikazawa ◽  
Hiroyuki Ohshima ◽  
Masato Uchita ◽  
Takayuki Miyagawa ◽  
...  
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Design Study ◽  
Pool Type

Inertial impaction behavior of fuel particles under severe accident scenario in pool type sodium cooled fast reactor

2013 ◽  
Vol 256 ◽  
pp. 285-290 ◽  
Author(s):  
B. Thilak ◽  
P. Mangarjuna Rao ◽  
B.K. Nashine ◽  
P. Chellapandi
Keyword(s):  
Fast Reactor ◽  
Severe Accident ◽  
Inertial Impaction ◽  
Accident Scenario ◽  
Fuel Particles ◽  
Pool Type
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