In-Vessel Melt Retention Mechanisms in Nuclear Reactor Systems

In-vessel melt retention (IVR) constitutes a cornerstone of severe accident management in water-cooled nuclear reactors, aiming to prevent radioactive release by stabilising molten core material—known as corium—within the pressure vessel. The approach relies on maintaining the integrity of the reactor pressure vessel (RPV) lower head through external cooling and on understanding the complex interplay between thermal loads imposed by the molten pool, natural convection within that pool, and the heat transfer across the vessel wall. Core degradation and relocation determine the initial distribution and composition of corium, while melt pool convection and stratification dictate local heat flux hot spots that challenge vessel material limits. Successful IVR demands optimised water flow paths for external cooling, accurate prediction of critical heat flux (CHF) on downward-facing surfaces, and thorough assessment of thermo-mechanical stresses and creep in vessel steel. Experimental facilities and advanced simulation codes have been developed to characterise high-temperature properties of corium, to evaluate stress–strain responses of ablated vessel walls, and to predict natural circulation patterns. Together, these efforts underpin a global strategy to enhance safety margins in advanced Generation III and III+ designs, facilitate retrofits in existing plants, and inform regulations for future reactors.

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In-Vessel Melt Retention Mechanisms in Nuclear Reactor Systems publication trend

The graph below shows the total number of articles in in-vessel melt retention mechanisms in nuclear reactor systems across all publications each year (not limited to Nature Index journals).

Technical terms

In-Vessel Melt Retention (IVR): A severe accident management strategy that seeks to stabilise and cool molten core materials within the reactor pressure vessel by external water cooling.

Corium: The molten mixture of nuclear fuel, cladding, structural materials and control-rod alloys formed during a severe reactor accident.

Critical Heat Flux (CHF): The heat-transfer limit at which a boiling surface transitions from efficient nucleate boiling to film boiling, significantly reducing cooling capability.

Reactor Pressure Vessel (RPV): The primary containment structure housing the nuclear core and coolant, designed to withstand high pressures and temperatures.

Melt Pool Convection: Natural circulation within the corium pool driven by temperature gradients, affecting heat flux distribution on the vessel lower head.

References

1. Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal‐Hydraulic System Analysis Code. International Journal of Energy Research (2023).
2. Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention. Nuclear Engineering and Design (2021).
3. High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention. Frontiers in Energy Research (2022).