Development of a versatile depletion code AMAC
The reactor fuel consumption calculation is an important part of reactor design and analysis, which usually needs to be completed iteratively by transport calculations and point-depletion calculations. AMAC is a newly versatile point-depletion and radioactive-decay code for use in simulation nuclear fuel cycles and calculating the nuclide compositions and radiation characteristics of materials. The code implements two depletion algorithms to treat the stiff depletion systems, including Transmutation trajectory analysis (TTA) method and Chebyshev rational approximation method (CRAM). With different higher-order approximations, the IPF form of CRAM is implemented to meet the high-precision requirements for the decay calculation of long-term nuclide system. And the accuracy analysis of higher-order CRAM is also discussed in the paper. In addition, the calculation of material radiation characteristics can be provided in AMAC to extend the application in different computational and analytical requirements. Then based on the complex nuclide system and a series of efficient programming methods, the high precision and efficiency of point-depletion calculation can be guaranteed. Three different calculation modes are supported by AMAC to execute the decay, constant flux and constant power calculations. By comparing with ORIGEN and FISPACT, the performance of AMAC in nuclide compositions and varieties of material radiation characteristics is proved and discussed. Furthermore, by coupling OpenMC and AMAC, the OECD/NEA burnup credit criticality benchmark and IAEA-ADS benchmark are used to validate the code performance of efficiency and precision in coupling calculations. All the results have shown good agreements with reference values and other codes, which demonstrates AMAC is a potential tool for accurate depletion calculation of reactors.