This paper presents an integrated computational methodology for solving the inverse heat conduction problem in geometrically complex, anisotropic reactor fuel elements. The methodology couples a parallel cell-centered finite volume forward solver capable of treating fully anisotropic thermal conductivity tensors, mixed boundary conditions of all three classical kinds, and radiation effects on unstructured three-dimensional meshes with a multilayer perceptron (MLP) trained on large datasets of numerically generated forward solutions and optimized by the Broyden-Fletcher-Goldfarb-Shanno algorithm (BFGS) quasi-Newton method. The forward solver operates in parallel using MPI with recursive coordinate bisection domain decomposition and a Jacobi-preconditioned conjugate gradient linear solver. Training datasets are generated by repeated solution of the steady-state heat conduction equation over randomly sampled physical parameters, and surface temperatures at up to 20 probe points serve as network inputs for recovery of the unknown thermophysical parameters. The approach is validated on twisted-geometry fuel rod configurations discretized on an unstructured mesh of 73 008 cells. For single-parameter inverse problems-recovery of thermal conductivity, heat transfer coefficient, or volumetric heat generation rate-the trained MLP achieves maximum prediction errors below 0.034%, with a minimal two-hidden layer network (8, 8 neurons) proving entirely sufficient. For the combined simultaneous recovery of the above parameters from 20 surface measurements, fixing at least one boundary condition resolves the intrinsic ill-posedness; a three-layer network (16, 32, 16 neurons) reduces maximum errors to below 0.02%. The methodology is further extended to anisotropic materials: The diagonal conductivity tensor components are recovered with a maximum error of 0.15%, and all six independent components of the fully general symmetric tensor are reconstructed with satisfactory accuracy for the dominant diagonal entries. A practical rule established by the numerical experiments is that the number of probe points should be approximately 4 to 5 times the number of parameters to be recovered simultaneously. Once the model is trained, inference requires only a single forward pass through the network (O(P) operations), making the approach directly applicable to real-time diagnostics, digital-twin frameworks, and in-reactor monitoring scenarios.
It is necessary for the determination of the design requirements of the detritiation system for the A-FNS to estimate the residual tritium amount in the components and pipes of the lithium target system for the A-FNS, where tritium would exist in residual lithium on the surface and dissolve in the structural materials. The residual lithium amount in each of the components and pipes of the Engineering Validation and Engineering Design Activities (EVEDA) Lithium Test Loop was measured to estimate that amount for the A-FNS, and the dissolved tritium amount in each of them was calculated numerically. As a result, the maximum tritium amounts for the residual lithium and that dissolved in the structural material for one component of the A-FNS were estimated as approximately 80 TBq and 2 GBq, respectively. It was also found that a hydrogen trap is necessary to decrease the concentration of the tritium dissolved in the materials below the clearance level.
Lithium-containing ceramics, such as lithium metatitanate and lithium orthosilicate, are candidate materials for solid breeder blanket modules in future fusion reactors. An important aspect of their application in fusion reactors is the accurate registration of tritium produced under irradiation, which is required for assessing tritium behavior and release characteristics. In this work, an irradiation ampoule device was developed and fabricated for reactor experiments with highly enriched lithium metatitanate pebbles (Li2TiO3 +5 mol % TiO2, up to 96% by Li-6). The device was installed in a physical experimental channel of the IVG.1M research reactor (Kurchatov, Kazakhstan) and equipped with K-type thermocouples for monitoring the sample temperature during irradiation. Reactor experiments were performed according to an approved experimental program at reactor thermal power levels of 1, 3, and 6 MW. During irradiation, tritium-containing molecules (HT and T-2) released from the pebble bed samples were continuously registered in situ using a vacuum extraction method. Tritium measurements were carried out by integrating two mass spectrometers-HIDEN Analytical Ltd. and RGA-100 (Stanford Research Systems)-into the LIANA experimental facility, which represents the main novelty of this work and improves the reliability of tritium registration. The temperature dependence of the partial pressures of tritium-containing molecules was investigated under reactor irradiation at different power levels. The main outcomes of this work include successful testing of the tritium registration method based on dual mass spectrometry, confirmation of the obtained experimental data, and determination of the temperature conditions governing tritium generation and release from lithium metatitanate pebbles.
As the power source for high-frequency transmitters in ion cyclotron resonance heating (ICRH) systems, high-power radio-frequency (RF) tetrodes present significant challenges in circuit modeling due to their complex operational principles and inherently interdisciplinary nature. This study addresses this challenge by developing a PSpice circuit model for a high-power RF tetrode through the fitting of its output characteristics, transfer parameters, and distributed parameters. Furthermore, the circuit modeling of a 1.5-MW ICRH transmitter and its associated high-frequency Inductor-Capacitor (LC) resonant circuit was completed. The model's functionality and frequency characteristics were subsequently verified via simulation against experimental data. Simulation results demonstrate that the numerical simulations of the proposed high-power RF tetrode model, the 1.5-MW ICRH transmitter circuit, and its high-frequency LC resonant circuit align well with the actual electrical characteristics of the 1.5 MW ICRH transmitter system. Consequently, this model can be effectively utilized for parameter optimization and performance analysis of ICRH systems.
Plasma-facing materials and components in a fusion power plant are subjected to a high flux of a mixture of hydrogen isotopes, deuterium, and tritium. They penetrate the first wall, and while most of them return to the plasma, a substantial amount diffuses through the wall and can either reach the opposite side or get trapped in structure defects created by neutron irradiation.Usually, it is considered that the distribution of deuterium and tritium in the bulk of the first wall is the same as in the incoming flux, i.e. equal concentrations of each isotope. However, the mass difference between deuterium and tritium leads to differences in the concentrations. This is related to the different reflection ratios and penetration depths of incoming ions, the difference in the diffusion coefficients, and the different detrapping energies due to zero-point energy contributions.We summarize the processes influencing the difference between the isotopes and use the open-source finite element code FESTIM to obtain numerical results for the time-dependent and steady-state distribution of isotopes in the first wall, as well as the flux of these isotopes across the wall and into the coolant. We show that relatively small differences in the properties of the isotopes of hydrogen can have an observable influence on their retention and flux in plasma-facing surfaces. Deuterium flux across the first wall was found to be higher than the flux of tritium, while tritium inventory in the wall can be up to 10% higher than that of deuterium in the case of a traps-free first wall.
An optimal lower hybrid current drive antenna has been designed for the Alborz tokamak using Multiphysics simulations performed in COMSOL. The study systematically investigates the influence of key parameters-including the antenna phasing configuration and radial displacement, along with the scrape-off layer plasma density, which is modeled under electron cyclotron resonance preionization-on the accessibility, and coupling efficiency of lower hybrid waves (LHWs) in the Alborz tokamak. The unique operational characteristics of the Alborz tokamak define a minimum value of N parallel to crit = 1.59, while the antenna achieves a representative value of N parallel to = 3.56 via the optimal phasing configuration of 120 deg, satisfying the LHW accessibility condition. Furthermore, this phasing configuration directs the power spectrum of launched LHWs asymmetrically in one toroidal direction, aligned with the electron ohmic drift, resulting in significantly enhanced power coupling. This is achieved through the attainment of the minimum reflectivity below 10% and maximum directivity of 72% when the plasma density exceeds the cutoff density of 7.44 & times;1016m-3 at the source frequency of 2.45GHz , which is critical for current drive. Additionally, a radial antenna displacement of 2cm , corresponding to the plasma density of 1.2 & times;1017m-3 , is identified as optimal, as it achieves the least reflectivity and enhances electric field intensity in the toroidal direction.
