A comprehensive depletion code MACT for molten salt reactors

In the realm of reactor design and analysis, the computation of nuclide depletion and the quantification of radioactive source terms are indispensable. These processes are underpinned by the rigorous solution of the depletion equation, which are essential for obtaining accurate determinations of nuclide inventories. To address the characteristics of online reprocessing, continuous fuel feeding, and nuclide migration in liquid fuel molten salt reactors (MSRs), a radioactive source term calculation code named MACT has been developed specifically for MSRs. The depletion solution algorithms in MACT employ the widely used Transmutation Trajectory Analysis (TTA) method and the Chebyshev Rational Approximation Method (CRAM). Pseudo-decay constant, pseudo nuclide method, and augmented matrix method are utilized to handle the reprocessing and feeding issues of MSRs. Additionally, a depletion model under multi-node isotopic migration conditions has been established, along with the development of corresponding calculation module. Finally, MACT has been validated through various tests, including the decay of Ra-228, activation of Cr-50, testing of extremely short-lived fission products, decay heat testing of U-235 thermal neutron fission, fuel salt irradiation and feeding, and multi-node nuclide migration tests. Numerical results demonstrate that MACT possesses high precision in nuclide inventory calculations and can easily handle the characteristics of MSR reprocessing, feeding, and nuclide migration, making it a potent engineering calculation tool for depletion calculations and source term analysis in MSRs.