Structural Integrity Assessment of Reactor Assembly Components of a Pool-Type Sodium Fast Reactor in a Core Disruptive Accident—I: Development of Computer Code and Validations

2010 ◽  
Vol 172 (1) ◽  
pp. 1-15 ◽  
Author(s):  
P. Chellapandi ◽  
S. C. Chetal ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Structural Integrity ◽  
Computer Code ◽  
Integrity Assessment ◽  
Sodium Fast Reactor ◽  
Reactor Assembly ◽  
Pool Type

Evaluation method for structural integrity assessment in core disruptive accident of fast reactor

2004 ◽  
Vol 227 (1) ◽  
pp. 97-123 ◽  
Author(s):  
Toshio Nakamura ◽  
Hitoshi Kaguchi ◽  
Iwao Ikarimoto ◽  
Yoshio Kamishima ◽  
Kazuya Koyama ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Evaluation Method ◽  
Structural Integrity ◽  
Integrity Assessment

Structural integrity of fast reactor components

1990 ◽  
Vol 331 (1619) ◽  
pp. 419-434 ◽  
Keyword(s):  
Fast Reactor ◽  
Structural Integrity ◽  
Ambient Pressure ◽  
Failure Process ◽  
Liquid Sodium ◽  
Secondary Circuit ◽  
Deformation And Failure ◽  
Integrity Assessment ◽  
High Boiling Point ◽  
Materials Used

The paper focuses on the generic aspects of the main structural integrity issues in the liquid-sodium-cooled fast reactor. The choice of sodium as a coolant has important consequences for the deformation and failure process in the materials used for the main plant components. For example, its high boiling point means that the prim ary and secondary circuit containment operates at ambient pressure and the system loading is dominated by thermal stress. The resultant low primary stresses make leak-before-break a viable integrity criterion for all sodium boundary components. Sodium coolant operates at comparatively high temperatures and this, together with the good heat-transfer properties, means that thermal fatigue and creep are of concern, particularly in the hotter parts of the plant. A third factor concerns the steam generators, where the integrity of the sodium—water boundary is particularly important. The paper will consider the failure processes that must be addressed in relation to these conditions and the development of the integrity assessment arguments.

Preliminary Analysis of In-Vessel Corium Confinement and Cooling in a Large Sodium Fast Reactor

2012 ◽  
Author(s):  
Franco Polidoro ◽  
Flavio Parozzi
Keyword(s):  
Fast Reactor ◽  
Preliminary Analysis ◽  
Structural Integrity ◽  
Mechanical Energy ◽  
Severe Accident ◽  
Sodium Fast Reactor ◽  
Safety Margins ◽  
Core Catcher ◽  
Core Meltdown ◽  
Accident Conditions

Considering a reasonable range of core meltdown accidents that can be postulated for GenIV sodium fast reactors, good safety margins exist for corium confinement and cooling inside the reactor vessel. Coolable conditions can be reached with the adoption of an ad-hoc device in the lower plenum, i.e. core catcher, capable to intercept the downward motion of the molten material and assure its cooling. Such device has to be designed to withstand to extreme thermal-mechanical conditions that rise as consequence of the large mechanical energy release and high temperature of molten corium. As this study has been carried out in the frame of the Collaborative Project on European Sodium Fast Reactor (CP ESFR) of the 7th Framework Programme Euratom, on the basis of the postulated accident conditions assumed for a reference 1500 MWe pool-type sodium fast reactor, the present work provides a preliminary analysis of the thermal response of a possible core catcher placed within the vessel. The dynamic thermal behaviour of the corium-structure-coolant system is analyzed with the computer code CORIUM-2D, an original simulation tool developed by RSE - Ricerca Sul Sistema Energetico, with the aim to assess the thermal interaction among corium, structures and coolant under severe accident conditions in both Light Water Reactors (LWRs) and Liquid Metal Fast Breeder Reactors (LMFBRs). The results of the numerical simulations show that the steady-state coolable configuration of core debris and the structural integrity of main containment structures can be reached in a number of partial core meltdown situations.

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
Sign in / Sign up

Export Citation Format

Share Document