Effect of MOX fuel and the ENDF/B-VIII on the AP1000 neutronic parameters calculations by using MCNP6

The present work studies the effect of introducing MOX fuel on Westinghouse AP1000 neutronic parameters. The neutronic calculations were performed by using the MCNP6 code with the ENDF/B-VII.1 library and the new release of the ENDF/B-VIII, the AP1000 core with three 235U enrichment zones (2.35 %, 3.40 %, and 4.45 %). The obtained results showed that the simulated model for the AP1000 core satisfies the optimization criteria as a Westing- house reference. The results which included: effective multiplication factor, keff, delayed neutron fraction, beff, excess reactivity, rex, shutdown margin, temperature reactivity coefficients, whole core depletion, neutron flux, power peaking factor and core power density, were calculated and compared with the available published results. The keff in the cold zero power was found to be 1.20495 and 1.20247 with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which matches the value of 1.205 presented in the AP1000 Design Control Document for the UO2 fuel core. On the other hand, keff in the cold zero power was found to be 1.19988 and 1.19860 for MOX core with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which show good reception and confirm the safety of the design and efficient modeling of AP1000 reactor core.

Study on Variation of HPGE Detector Dead Layer Thickness and its Effect on the Measurements of the Detector Response and Samples Characterization Using Monte Carlo Simulation

The main objective of this work is to produce an optimal modeling for our aged Planar-HPGe detector using Monte Carlo method (MC). That optimization included the analysis of the germanium dead (inactive) layer thickness for our old detection system (planar-HPGe detector). DL is one of the important parameters needed in order to obtain the smallest discrepancy between simulated and experimental measurements of detector efficiency. Also, precise determination of 235U enrichment for UO2 samples which is necessary for purposes of nuclear materials verification in the field of nuclear safeguards. The thickness of Germanium dead layer (DL) can be vary by time as it is not well known due to the existence of a transition zone where photons are strongly attenuated and absorbed, that cannot contribute to the total photon energy absorption which causes a significant decrease in efficiency. Therefore, using data provided by manufacturers since long years (manufacture date) in the detector simulation model is not convenient. As a result, some strong discrepancies appear between measured and simulated efficiency, in addition to that non-accurate results for 235U enrichment determination. The Monte Carlo method applied to overcome this difficulty was to vary the thickness of dead layer step by step in simulation, a good agreement (minimum deviation) between estimated and experimental efficiency was reached when a suitable germanium dead layer thickness was chosen. Calculations and measurements were performed for radioactive nuclear material samples in the form of UO2 powder with different sizes and enrichments at different locations, under different gamma-lines emitted after a-decay of the 235U nuclei. Results indicated that a good agreement between simulated and measured efficiencies is obtained using a value for the germanium dead layer thickness approximately (2.45 mm) six in comparison with (0.389 mm) provided by the detector manufacturer.

Uranium Enrichment Measurement Using Enrichment-Meter Approach

The 235U enrichment is one of the most important characteristics of nuclear materials for nuclear safeguards purposes. The multi-group γ-ray analysis method for uranium (MGAU) is an important non-destructive gamma spectroscopy method for 235U enrichment determination. Using that method, the typical measurement bias is below 3% for uranium material with abundance from 0.3 to 93 %. However, it is not applicable for the samples with thick container or without isotopic decay equilibrium. In this work, the enrichment meter method was studied with two uranium dioxide samples (235U abundance 0.71 % and 3.167 %). The nuclear materials spectra were measured using a planar high-purity germanium detector. Based on the specific gamma peak (185.71 keV) of relative high intensity, this traditional enrichment meter approach gives measurement bias more than 10 %. Thus, this work represents two objects: (1) an alternative approach which was investigated, where the calibration is performed through Monte Carlo simulation (MCNP5) instead of experiment in advance, as the measurement bias was reduced to be around 5 %. Thus, to use this approach, one should have the sample details, such as dimensions, chemical composition and container. (2) The influence of the container wall thickness on the measurement accuracy by Monte Carlo simulation. So, if the container wall thickness is not modeled correctly the measurement accuracy is influenced, which is investigated by simulation.