## 4 Computational Technique  
### Computer Codes and Nuclear Data  
The assessment of the isotope production in the irradiation unit of the DEMO was conducted in two steps: (1) the assessment of the spatial distribution of the neutron spectra in the irradiation rods; (2) the activation analyses involving the neutron spectra to quantify the generation of the radioactive isotopes.  
The neutron transport simulation (1) implies the calculation of the neutron spectra in all geometry cells of the irradiation rods where activation and depletion analyses are carried out. To enable this, each rod was subdivided in the radial direction into 50 mm long parts (the typical size of the irradiation capsule) to assess the radial distribution of the neutron spectra. This was achieved by applying the variance reduction technique, resulting in a low statistical uncertainty of the results, typically below 0.5%. The neutron flux intensities and the spectra necessary for the activation analysis were mapped over the irradiation unit geometry, making use of MCNP6.2 [44], CCFE 709 neutron energy group structure [45] to obtain accurate collapsed reaction rates and JEFF-3.3 nuclear cross-section data [46]. An explicit plasma neutron source description was introduced in the MCNP input deck involving the model developed in [47]. In the present study, only the accumulation and the time evolution of the generated radioactive isotopes inventories were analyzed. To this end, the neutron spectra were tallied only in the geometry cells containing precursor materials for the isotope production. Special attention was paid to the accuracy of the neutron flux calculations, and the achieved statistical uncertainty of the total neutron flux was less than 0.5%. For some low-energy bins, for instance, 1 \(\times\) 10\({}^{-7}\)-1 \(\times\) 10\({}^{-4}\) MeV, the relative error of the MCNP calculations was below 10%. The DEMO fusion nuclear power of 1998 MW [1] was taken for the normalization of the nuclear responses, which results in plasma neutron source intensity of 7.094 \(\times\) 10\({}^{20}\) s\({}^{-1}\).  
The activation analyses, step 2, assume numerous inventory calculations. To support the study, a specially developed computer interface was applied to produce the final results [43]. The interface goes through the MCNP input file for the neutron transport calculations and picks up actual material compositions used in the calculations. For the execution of numerous FISPACT II [45] inventory calculations, the interface prepares specific input files using fine material composition and a proper irradiation scenario for the production of each radioisotope and submits them in parallel mode, making use of the MPI technique. The latest version, 2021, of the TENDL activation library [48] is used in the FISPACT II calculations. The results of the FISPACT inventory calculations are collected and processed by a post-processing code. Finally, the code provides the time evolution of the activity and decay heat in a separate cell and groups of cells, results for dominant nuclides, specific data such as masses, differential activity, and decay heat of the nuclides accumulated in the geometry cells.