To minimize the potential risk of design extension conditions (DEC) with core meltdown, some advanced reactors employ ex-vessel core catchers which stabilize and cool the corium for prolonged period by strategically flooding it. This paper describes the coolability of the melt pool and ablation process in a scaled down ex-vessel core catcher employing sacrificial material which reduces the specific volumetric heat, temperature, and density of the melt pool. To understand these phenomena, a simulated experiment was carried out. The experiment was performed by melting about 500 kg of corium simulant using thermite reaction at about 2500 °C. The bricks of oxidic sacrificial material were arranged in the core catcher vessel which was surrounded by a tank filled with water up to a certain level. After the time required for melt inversion, water was introduced to flood the test section from the top. The melt pool temperatures were monitored at various locations using “K” and “C” type thermocouples to obtain ablation depth at different elevations with time. The results show that the coolability of the molten pool in the presence of water for the present geometry is achievable with outside vessel temperatures not exceeding 100 °C. A ceramic stable crust was observed at the top surface of the melt pool, which prevented water ingression into the molten corium. The ablation rate was found to be maximum at the lower corners of the brick arrangement with the maximum value being 0.75 mm/s. An average rate of about 0.18 mm/s was obtained in the brick matrix.
Skip Nav Destination
Article navigation
October 2019
Research-Article
Experimental Investigation of Melt Coolability and Ablation Behavior of Oxidic Sacrificial Material at Prototypic Conditions in Scaled Down Core Catcher
Samyak S. Munot,
Samyak S. Munot
Engineering Sciences,
Homi Bhabha National Institute,
Anushakti Nagar,
Mumbai, Maharashtra 400094, India
e-mails: samyakmunot@hotmail.com;
samyakm@barc.gov.in
Homi Bhabha National Institute,
Anushakti Nagar,
Mumbai, Maharashtra 400094, India
e-mails: samyakmunot@hotmail.com;
samyakm@barc.gov.in
Search for other works by this author on:
Ganesh V,
Ganesh V
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: ganeshv@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: ganeshv@barc.gov.in
Search for other works by this author on:
Parimal P. Kulkarni,
Parimal P. Kulkarni
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: parimalk@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: parimalk@barc.gov.in
Search for other works by this author on:
Arun K. Nayak
Arun K. Nayak
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India;
Engineering Sciences,
Homi Bhabha National Institute,
Mumbai, Maharashtra 400094, India
e-mail: arunths@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India;
Engineering Sciences,
Homi Bhabha National Institute,
Anushakti Nagar
,Mumbai, Maharashtra 400094, India
e-mail: arunths@barc.gov.in
1Corresponding author.
Search for other works by this author on:
Samyak S. Munot
Engineering Sciences,
Homi Bhabha National Institute,
Anushakti Nagar,
Mumbai, Maharashtra 400094, India
e-mails: samyakmunot@hotmail.com;
samyakm@barc.gov.in
Homi Bhabha National Institute,
Anushakti Nagar,
Mumbai, Maharashtra 400094, India
e-mails: samyakmunot@hotmail.com;
samyakm@barc.gov.in
Ganesh V
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: ganeshv@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: ganeshv@barc.gov.in
Parimal P. Kulkarni
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: parimalk@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India
e-mail: parimalk@barc.gov.in
Arun K. Nayak
Reactor Engineering Division,
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India;
Engineering Sciences,
Homi Bhabha National Institute,
Mumbai, Maharashtra 400094, India
e-mail: arunths@barc.gov.in
Bhabha Atomic Research Centre,
Mumbai, Maharashtra 400085, India;
Engineering Sciences,
Homi Bhabha National Institute,
Anushakti Nagar
,Mumbai, Maharashtra 400094, India
e-mail: arunths@barc.gov.in
1Corresponding author.
Manuscript received November 14, 2018; final manuscript received February 26, 2019; published online July 19, 2019. Assoc. Editor: Tomio Okawa. This work was prepared while under employment by the Government of India as part of the official duties of the author(s) indicated above, as such copyright is owned by that Government, which reserves its own copyright under national law.
ASME J of Nuclear Rad Sci. Oct 2019, 5(4): 041206 (7 pages)
Published Online: July 19, 2019
Article history
Received:
November 14, 2018
Revised:
February 26, 2019
Citation
Munot, S. S., V, G., Kulkarni, P. P., and Nayak, A. K. (July 19, 2019). "Experimental Investigation of Melt Coolability and Ablation Behavior of Oxidic Sacrificial Material at Prototypic Conditions in Scaled Down Core Catcher." ASME. ASME J of Nuclear Rad Sci. October 2019; 5(4): 041206. https://doi.org/10.1115/1.4043106
Download citation file:
Get Email Alerts
Cited By
Operation Optimization Framework for Advanced Reactors Using a Data-Driven Digital Twin
ASME J of Nuclear Rad Sci (April 2025)
Numerical Analysis of Gas Generation and Migration in a Radioactive Waste Disposal Cell of a Deep Geological Repository
ASME J of Nuclear Rad Sci (April 2025)
Related Articles
Adaptive Collocation Method for Simultaneous Heat and Mass Diffusion With Phase Change
J. Heat Transfer (August,1984)
High-Resolution Thermal Profiling of a Combustor in a Non-Dedicated Test Using Thermal History Coatings
J. Turbomach (November,2022)
Ablation of a Solid Material by High-Temperature Liquid Jet Impingement: An Application to Corium Jet Impingement on a Sodium Fast Reactor Core-Catcher
ASME J of Nuclear Rad Sci (January,2022)
The Development of Candling Module Code in Module In-vessel Degraded Analysis Code MIDAC and the Relevant Calculation for CPR1000 During Large-Break LOCA
ASME J of Nuclear Rad Sci (April,2016)
Related Proceedings Papers
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Control and Operational Performance
Closed-Cycle Gas Turbines: Operating Experience and Future Potential
Nuclear Fuel Materials and Basic Properties
Fundamentals of Nuclear Fuel