Abstract
The iron cross-section in thermal regions influences the thermal neutron flux prediction in steel structural components of reactors and also in regions adjoining them. The thermal neutron flux level is proportional to pin power density in fuel. This quantity is an important criterion reflected in limits and conditions of reactor operation. The new power density evaluation shows notable, well distinguishable discrepancy between calculations realized using the CENDL-3.1 nuclear data library and experimentally determined pin power density in boundary rows of pins. All experiments were carried out in a water–water energetic reactor (VVER-1000) transport mock-up placed in the LR-0 reactor.
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