## 1 Introduction  
To provide an exact estimate of the flows of fuel components in ITER, DEMO, and other fusion facilities under construction, it is necessary to conduct detailed integrated simulations of the plasma, outside the scope of the simple studies of particle balance and plasma transport in the active zone that are usually used for this purpose [1]. This is necessary because the conditions necessary for ensuring parameters and fusion plasma confinement can have a substantial effect on plasma replenishment by the recycled particles and on the transfer of edge particles [2, 3]. It was noted in [1] that the results obtained from integrated simulations can differ by more than one order of magnitude from the estimates obtained by simpler models (in particular, [4, 5]). Earlier, an approach was developed and discussed to simulations of tritium FC systems and calculation of the amount of tritium at the facility, consistently with the core and divertor plasma [6, 7, 8]. In this approach, similarly to [9], the state of the core (inside the separatrix) and edge (outside the separatrix) plasma is simulated by a combination of the ASTRA [10] and SOLPS4.3 [11] codes and the particle flows in the FC systems are simulated by the FC-FNS [12] code. The integration between these three independent codes is carried out by an indirect scheme, where the output data from one code is parametrized and further used as input data in the other code [13]. This allowed us to create an effective calculation workflow that describes the interaction of different components of the model with strongly different calculation times. In the developed approach, SOL modeling was carried out with parameterization of the results [13], then these data were used as boundary conditions in the core modeling [6; 8]. Based on the obtained simulations, the characteristic particle retention times were calculated (see below) and zero-dimensional modeling of the particle balance inside the vacuum chamber was carried out:  
\[ \begin{split} N_{core}=N_{sep}+S_{NB}\tau_{NB}+(S_{pd(LFS)}+S_{pel (HFS)}-\alpha_{ELM}P_{SOL}/3\times T_{pd})\tau_{pol}+\\ +S_{sep}\tau_{sep}-S_{fus}\tau_{tot}\end{split}  \tag{1}\]  
by which the flows in the pumping, processing and injection systems were modeled. \[ \begin{split} S_{sep}^{T}=S_{sep}(S_{puff}^{T}+S_{out}^{T})/(S_{ pump}+S_{sep})\end{split}  \tag{2}\]  
\[ \begin{split} S_{puff}^{T}=S_{pel(HFS)}^{T}(1/k_{eff\_HFS}-1)+S_{GIS }^{T}\\ S_{puff}^{D}=S_{NB}^{D}(1/k_{eff\_NBI}-1)+S_{pd(HFS)}^{D}(1/k_{ eff\_HFS}-1)+\\ +S_{pd(LFS)}^{D}(1/k_{eff\_LFS}-1)+S_{GIS}^{D}\end{split}  \tag{3}\]  
et al. (see [8]). The earlier used model assumes that the core plasma is replenished by particles from three different sources (neutral beam injection, pellet injection, and gas supply to edge/SOL) that have different confinement times. These \"partial\" confinement times were determined after some sources were artificially decreased by \(\sim\)10% [7]. Since, in earlier simulations by the ASTRA and SOLPS codes the hydrogen isotopes were not distinguished, in order to estimate that deuterium and tritium flows by the FC-FNS code, we used an additional assumption that the transport coefficients of the D and T particles were equal in the core plasma. This assumption is rather rough and may carry a potential inaccuracy for the result (so the flows may be in opposite directions for D and T--see [3; 14]). Theoretically, it is possible to introduce different transport coefficients for different isotopes, yet there is no detailed data on the particle transport of different isotopes, and in particular on the mixing of isotopes. The existing theoretical estimates and experimental data (see, e.g., [15]) indicate that the dependence of particle confinement on its mass is relatively weak, \(\sim\)\(m^{1/2}\). Therefore, taking into account the ratio of T/D masses that is equal to 1.5, the above assumption appears reasonable.