Neutron shielding assessment of ordinary concrete in an AmBe mixed-energy neutron field

The neutron shielding performance of a 10 × 10 × 10 cm concrete block (density 2.41 g/cm 3 ) was evaluated experimentally and computationally in the mixed-energy neutron field of a 16 Ci 241 Am-Be source. Spectrum-integrated neutron capture reaction rates were measured at the incident and transmitted faces of the block using 197 Au (n, γ) 198 Au gold-foil activation and high-purity germanium (HPGe) gamma-ray spectrometry. The measured reaction rates at the front and rear foil positions were 9.25(33) × 10 −19 and 5.42(21) × 10 −19 atom −1 s −1 , respectively. Corresponding Monte Carlo simulations using the OpenMC code with the ENDF/B-VII.1 nuclear data library yielded. 9 . 8 8 4 ⁢ ( 7 7 ) × 1 0 − 1 9 and 5 . 7 1 ⁢ ( 1 7 ) × 1 0 − 1 9 atom −1 s −1 , representing percentage differences of 6.85% and 5.39%. The ratio of the transmitted-to-incident reaction rate agreed to within 1.37% between experiment and simulation, confirming the validity of the transport model for the spectrum-weighted quantity. The neutron transmission factor through the concrete block was determined to be 0.594, corresponding to a ∼ 41% flux reduction relative to the open-beam configuration. Gamma-ray spectrometry of the irradiated concrete identified activation products qualitatively, including 56 Mn (846.3 keV, 1809.2 keV), 24 Na (1367.2 keV), and naturally occurring 40 K (1459.3 keV). Energy-dispersive X-ray spectroscopy (EDX) confirmed the presence of the principal heavy elements (Ca, Si, Fe, Al, K, Mg) in the concrete sample, consistent with the reference composition used in the transport model. The results confirm the reliability of the OpenMC with ENDF/B-VII.1 framework for predicting neutron transport and activation in standard concrete shielding and provide an initial experimental benchmark dataset for future radiation-shielding-related studies at this facility.