Structural Integrity Assessment of Reactor Assembly Components of a Pool-Type Sodium Fast Reactor in a Core Disruptive Accident—II: Analysis for a 500-MW(electric) Prototype Fast Breeder Reactor

2010 ◽  
Vol 172 (1) ◽  
pp. 16-28 ◽  
Author(s):  
P. Chellapandi ◽  
S. C. Chetal ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Structural Integrity ◽  
Fast Breeder Reactor ◽  
Integrity Assessment ◽  
Sodium Fast Reactor ◽  
Reactor Assembly ◽  
Pool Type

Sodium-Free Level Fluctuations: Concern on Structural Integrity for the Pool-Type Sodium-Cooled Fast Breeder Reactor

2017 ◽  
pp. 407-417
Author(s):  
V. R. Chandan Reddy ◽  
R. Suresh Kumar ◽  
Anil Kumar Sharma ◽  
K. Velusamy ◽  
P. Selvaraj
Keyword(s):  
Structural Integrity ◽  
Fast Breeder Reactor ◽  
Free Level ◽  
Pool Type

Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

2006 ◽  
Author(s):  
P. M. Rao ◽  
N. Kasinathan ◽  
S. E. Kannan
Keyword(s):  
Structural Integrity ◽  
Peak Pressure ◽  
Safety Issue ◽  
Pressure Rise ◽  
Pool Fire ◽  
Mode Analysis ◽  
Fast Breeder Reactor ◽  
Reactor Assembly ◽  
Computer Codes ◽  
Breeder Reactors

The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE–II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m3 RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa.

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