The paper aims at evaluating the Control Rod (CR) reactivity loss due to its depletion under flux, in the specific case of Pressurized Water - Small Modular Reactors operating without soluble boron. The impact of the CR depletion at the lattice level was investigated with the APOLLO2.8 code, showing a potential CR worth loss of several thousand of pcm for Ag-In-Cd and natural Hf 24-pins clusters. Then, the core level assessment was conducted on the PRATIC core, a boron-free SMR. A CR depletion model was implemented within the COCAGNE code, by modifying some specific parts at the user interface level. The CR worth loss was estimated at ~750 pcm after one irradiation cycle with a consequent increase of the CR insertion to compensate for the absorber depletion, leading to an impact on the core axial power distribution that cannot be considered negligeable for detailed safety analysis, or code comparison to the reactor experimental data.
Reactivity Impact of Control Rod Depletion in Boron-free Reactors / Girardi, E., Quarta, M., Dulla, S.. - ELETTRONICO. - (2025). (ICAPP 2025 Antibes 17-19 Settembre 2025).
Reactivity Impact of Control Rod Depletion in Boron-free Reactors
Quarta, Massimo;Dulla, Sandra
2025
Abstract
The paper aims at evaluating the Control Rod (CR) reactivity loss due to its depletion under flux, in the specific case of Pressurized Water - Small Modular Reactors operating without soluble boron. The impact of the CR depletion at the lattice level was investigated with the APOLLO2.8 code, showing a potential CR worth loss of several thousand of pcm for Ag-In-Cd and natural Hf 24-pins clusters. Then, the core level assessment was conducted on the PRATIC core, a boron-free SMR. A CR depletion model was implemented within the COCAGNE code, by modifying some specific parts at the user interface level. The CR worth loss was estimated at ~750 pcm after one irradiation cycle with a consequent increase of the CR insertion to compensate for the absorber depletion, leading to an impact on the core axial power distribution that cannot be considered negligeable for detailed safety analysis, or code comparison to the reactor experimental data.| File | Dimensione | Formato | |
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https://hdl.handle.net/11583/3013131
