In-Vessel Retention (IVR), which arrests relocated molten core materials in the vessel during severe accident, has been singled out as an appealing accident management approach to many reactors. The heat transfer imposed by in-vessel corium is a vital part for IVR success considering the difficulty of significantly altering ex-vessel CHF. For a given decay power, corium pool configuration determines the heat flux profile along the vessel wall, which may produce uncertainties associated with IVR strategy. In this paper, a thermodynamic tool is employed to study the corium pool configurations by analyzing the possible interaction among relocated corium, zircalloy cladding and core internals. The results reveal the immiscibility gap phenomena under high temperature which separates molten materials into oxidic and metal phase in the lower head. The oxidic phase is quite stable and its density is only slightly changed by various accident scenarios. The metal phase is relatively unstable and its density is susceptible to the condition of cladding oxidation degree and crust integrity. The corium pool configurations in the lower head are determined based on the results of thermodynamic analysis and phase density comparison. Both two-layer and three-layer corium pools are likely to be formed under different accident scenarios. CAP1400 has intentionally increased the mass of lower core support plate, which is a beneficial design change to prevent possible focusing effect if material infiltration through crust is assumed to be impossible.
Skip Nav Destination
2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5785-4
PROCEEDINGS PAPER
Research on IVR-Relevant Phenomena of Material Thermodynamic Interaction and Corium Pool Configuration Available to Purchase
Peiwen Gu,
Peiwen Gu
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Guobao Shi,
Guobao Shi
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Kemei Cao,
Kemei Cao
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Jiayun Wang
Jiayun Wang
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Peiwen Gu
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Guobao Shi
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Kemei Cao
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Jiayun Wang
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Paper No:
ICONE25-66969, V007T11A011; 6 pages
Published Online:
October 17, 2017
Citation
Gu, P, Shi, G, Cao, K, & Wang, J. "Research on IVR-Relevant Phenomena of Material Thermodynamic Interaction and Corium Pool Configuration." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 7: Fuel Cycle, Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Mitigation Strategies for Beyond Design Basis Events. Shanghai, China. July 2–6, 2017. V007T11A011. ASME. https://doi.org/10.1115/ICONE25-66969
Download citation file:
21
Views
Related Proceedings Papers
Related Articles
COBRA-TF Simulation of DNB Response During Reactivity-Initiated Accidents Using the NSRR Pulse Irradiation Experiments
ASME J of Nuclear Rad Sci (July,2016)
Heat Transfer Behavior of Silica Nanoparticles in Pool Boiling Experiment
J. Heat Transfer (April,2008)
Gas Microflows in the Slip Flow Regime: A Critical Review on Convective Heat Transfer
J. Heat Transfer (February,2012)
Related Chapters
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Effect of Chromium Content on the On-Cooling Phase Transformations and Induced Prior-β Zr Mechanical Hardening and Failure Mode (in Relation to Enhanced Accident-Tolerant Fuel Chromium-Coated Zirconium-Based Cladding Behavior upon and after High-Temperature Transients)
Zirconium in the Nuclear Industry: 20th International Symposium
E110opt Fuel Cladding Corrosion under PWR Conditions
Zirconium in the Nuclear Industry: 20th International Symposium