Many integral neutronic parameters such as the effective multiplication factors (keff) are based on neutron reactions with matter through cross sections. However, these cross sections present uncertainties, of origin multiple, which reduce the safety margin of nuclear installations. In order to minimize these risks, a sensitivity analysis is necessary to indicate the rate of change of a reactor performance parameter compared to variations in cross sections. Thus, several critical benchmarks were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE), and their sensitivities and covariance matrix of the desired cross section were processed by MCNP6 and NJOY codes, respectively, in ENDF/B-VII.1 and JENDL-4.0 evaluations. The results obtained show that the 44 energy groups give the most varied sensitivity profiles than those given by others (15 and 33). In addition, we observed large uncertainties on the keff due to the H-1 and O-16 cross-sectional uncertainties (∼200–1000 pcm) in ENDF/B -VII.1 and the U-235 cross section in JENDL-4.0; however, keff’s uncertainties due to the cross-sectional uncertainties of the U-238 are very small.
Part of the book: Nuclear Power Plants
The different codes based on the Monte Carlo method, allows to make simulations in the field of medical physics, so the determination of all the magnitudes of radiation protection namely the absorbed dose, the kerma, the equivalent dose, and effective, what guarantees the good planning of the experiment in order to minimize the degrees of exposure to ionizing radiation, and to strengthen the radiation protection of patients and workers in clinical environment as well as to respect the 3 principles of radiation protection ALARA (As Low As Reasonably Achievable) and which are based on: -Justification of the practice -Optimization of radiation protection -Limitation of exposure.
Part of the book: The Monte Carlo Methods
In this chapter we present our MCNP modeling, concerning fast critical experimental benchmarks, about qualifying our libraries of cross-sections deduced from the evaluations ENDF/B-VII, JEFF-3.1, JENDL-3.3, JENDL-4 processed by the code NJOY. The benchmarks analyzed are characterized by simple geometries which help to have a precise calculation. In our neutron calculation, we used the MCNP code (version 5), the reference code for the neutron transport calculation with the Monte Carlo method. It is also very efficient for criticality calculation. The cross-section data for all the isotopes that make up the material of the studied benchmarks are processed in ACE format at 300 K temperature using the NJOY 99.9 modular system. A detailed comparison of the criticality results of our simulation was carried out to highlight the influence of these evaluations on the keff calculations.
Part of the book: Qualitative and Computational Aspects of Dynamical Systems