Integrity Assessment of Reactor Pressure Vessels Under Thermal Shock

Reactor pressure vessels (RPVs) form the primary barrier against radioactive release in water-cooled nuclear reactors and must withstand extreme thermal transients without loss of structural integrity. Pressurised thermal shock (PTS) events arise when cold coolant is introduced into a hot vessel, creating steep temperature gradients that interact with internal pressure to drive stress concentrations. Such conditions accelerate irradiation-induced embrittlement of ferritic steels, heightening the risk of brittle fracture. Contemporary integrity assessments combine detailed thermomechanical simulations with fracture mechanics to predict crack initiation and growth. Three-dimensional finite element models capture coupled temperature–stress fields, while sub-modelling techniques refine stress intensity factor distributions in critical regions such as nozzle intersections. Probabilistic and deterministic frameworks both inform safety margins and service-life extension strategies. Advances in adaptive meshing, master-curve fracture toughness evaluation and multi-step coupling schemes have improved the fidelity of predictions, supporting regulatory decision-making and guiding operational limits. The global significance of this research underpins continuous operation of ageing nuclear fleets and informs design of new reactors with enhanced thermal shock resilience.

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Integrity Assessment of Reactor Pressure Vessels Under Thermal Shock publication trend

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Technical terms

Pressurised thermal shock (PTS): Rapid introduction of cold fluid into a hot, pressurised vessel causing sharp thermal gradients and elevated stress.
Stress intensity factor (SIF): Parameter that characterises the magnitude of the stress field near a crack tip, fundamental to fracture mechanics.
Finite element model (FEM): Numerical technique dividing a structure into discrete elements to approximate its response under load.
Fracture toughness: Material property denoting resistance to crack propagation, typically assessed via the master-curve approach.

References

1. Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System. Science and Technology of Nuclear Installations (2022).
2. Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus‐FRANC3D Simulation Method. Science and Technology of Nuclear Installations (2021).
3. Effects of angles and shapes of a corner crack on the driving force at a set-in nozzle-cylinder in a PWR pressure vessel. Annals of Nuclear Energy (2023).