Fusion Reactor Safety Research Articles

GaN-based W-band receiver chip development for fusion plasma diagnostics

Millimeter-wave diagnostics have proven effective on various magnetic fusion devices worldwide, yet the formidable challenges posed by the harsh environments of future burning plasma devices, characterized by extreme temperatures, pressures, and radiation levels, remain a significant hurdle. To address these challenges, the utilization of wide bandgap Gallium Nitride (GaN)-based millimeter-wave diagnostics is a most promising solution for fusion reactor safety monitoring and control. A noteworthy W-band GaN-based system-on-chip receiver has been the demonstrated by employing HRL T3 40 nm GaN technology. This receiver chip, compactly designed with dimensions of 3 × 5 mm2, incorporates essential components such as the 75–110 GHz RF Low-Noise Amplifier (LNA), mixer, Intermediate Frequency (IF) amplifier, and Local Oscillator (LO) chain. This receiver chip will be packaged as a millimeter-wave receiver module and applied on the DIII-D National Fusion Facility, for fusion plasma edge shape monitoring for operational safety and dangerous disruption prediction. The laboratory measurement results have demonstrated suitable performance. This advancement is pivotal for accurate analysis of plasma behavior in the extreme conditions of burning plasma devices, driving progress in fusion research and technology.

Mechanism of hydrogen isotope exchange for tritium removal in plasma-facing materials: a multi-scale investigation

The safety of future fusion reactors is critically dependent on the tritium (T) retention in plasma-facing materials. Hydrogen isotope (HI) exchange offers a method to redistribute HIs within solid materials, presenting a feasible approach for removing T from bulk materials and trapped by strong trapping sites. Nonetheless, unraveling the intricate mechanism behind HI exchange remains an urgent yet formidable challenge. This study undertakes a comprehensive investigation into the mechanism of HI exchange in tungsten materials across multiple scales. First, we developed a multi-component hydrogen isotope transport and exchange model (HIDTX) based on classical rate theory. The model validation was further carried out, demonstrating good consistency with the well-controlled laboratory experiments. From the results of different comparative models in HIDTX, it is found that the reduction in deuterium retention due to HI exchange was primarily driven by three synergistic effects: competitive re-trapping, collision, and swapping effects. Through molecular dynamics (MD) and first-principles calculations, the microscopic mechanism of HI exchange was revealed to be that the presence of hydrogen atoms in the interstitial sites surrounding a vacancy in tungsten decreased the binding energy between the vacancy and hydrogen. Meanwhile, we discovered that the combination of thermal desorption and HI exchange can significantly lower the temperature required for the hydrogen removal and enhance the removal rate. Particularly, the hydrogen removal time can be shortened by approximately 95% with simultaneous HI exchange compared to that with only thermal desorption. This work provides a practical guideline for comprehending and subsequently designing for efficient T removal in future nuclear fusion materials.

Estimation of different calculation models for evaluating heavy ion-induced damage in plasma facing materials

In a fusion reactor, plasma-facing materials are bombarded by various energetic particles (such as neutrons) from fusion reactions, leading to inevitable irradiation damages. These damages can significantly affect the material servicing life time and threaten the safety of fusion reactors. Accurate and consistent evaluation of irradiation damage is quite important to investigate the evolution mechanism of irradiation defect. In this study, the irradiation damage of tungsten caused by high-energy particles were evaluated using the SRIM code and molecular dynamics simulations. The accuracy of three atomic displacements calculation models, i.e., VAC model, NRT model, and arc-dpa model, under two computational modes of SRIM (the K-P mode and F-C mode) were discussed. It was found that with the increase of the incident particle atomic number, the peak depth of displacement moved to the shallow layer and increased in peak height, while the trend was opposite with the increase of the incident particle energy. As to the different calculation modes, the F-C and K-P modes produced a ratio of vacancies of about 1.8 times in the VAC model. What's more, the VAC model and NRT model predicted results consistent with molecular dynamics simulation results at lower energy ranges (approximately <100 eV), but deviated from them as the energy increased. In contrast, the arc-dpa model exhibited good consistency with the molecular dynamics simulations and had higher accuracy. Therefore, it is recommended to use the arc-dpa model under the K-P mode for the calculation of atomic displacement. This model can accurately and quickly calculate the irradiation damage level caused by high-energy particle irradiation in fusion reactors.

An approach for multi-objective optimization based on preliminary PSA and its applications for fusion reactors

The fusion reactors, which adopt Deuterium-Tritium (D-T) nuclear reaction to produce energy, are promising in the future. The nuclear safety of fusion reactors is an important research topic on the road of fusion reactor engineering. The Probabilistic Safety Assessment (PSA) is one of the important tools to evaluate the safety concerns of fusion reactors, and the importance obtained from the PSA of fusion reactors can provide references for the items safety classification and reliability allocation. Reasonable safety classification and reliability allocation are of great significance for improving the safety and economy of fusion reactors. The PSA, items safety classification and reliability allocation for fusion reactors influence each other, form a whole, and can be considered as a multi-objective optimization problem. In this contribution, preliminary PSA and its applications for fusion reactors, which include the preliminary PSA, items safety classification and reliability allocation based on PSA for fusion reactors, were performed. Then, the multi-objective optimization among the preliminary PSA and its applications for fusion reactors were preliminarily discussed. The results show that the proposed systematic approach can provide important references for the optimization and improvement of the safety and economy for fusion reactors.

Steady-state tritium inventory in plasma-facing tungsten for fusion reactors: An effective calculation method and implications of calculation results

The tritium inventory in plasma-facing components needs to be evaluated with sufficient accuracy since it is relevant to the safety and fuel-cycle sustainability of fusion reactors. Here, we propose an effective method to calculate the steady-state tritium inventory in plasma-facing tungsten. In this method, the surface and diffusion processes of hydrogen isotopes (HI) are first modelled by rate theory to obtain a steady-state concentration profile of HI dissolved in the lattice. Subsequently, trap effects are modelled by equilibrium theory. By decoupling the trapping/detrapping processes from the surface/diffusion processes and focusing only on the steady state, the calculation cost is largely reduced, and atomic resolution is realized in the rate model to describe surface processes in detail. The applicability and performance of the proposed method are demonstrated with a one-dimensional test case simulating the ITER tungsten mono-block. The calculation results show that tritium retention in the deep region becomes the main source of tritium inventory at a steady state due to the large temperature gradient in plasma-facing tungsten. Several implications for DEMO and beyond are obtained regarding tritium inventory and HI effects on mechanical integrity. The proposed method can be complementary to the macroscopic rate model for tritium inventory calculations.

Configuration and characters of novel Li2TiO3 tritium breeder cubic units with tuned channel structure via lithography-based 3D printing

Li2TiO3 is regarded as a vital tritium breeder candidate due to its favorable properties in Helium-Cooled Pebble Bed. However, the point contact stress breakage for traditional pebble structure can cause the blocking of tritium recovery system, and seriously reduce the safety and tritium self-sufficiency ability of fusion reactor. For addressing above problems, in this paper, complex-shaped Li2TiO3 cubic units were manufactured firstly via lithography-based 3D printing. According to the principle of tritium transport, the cubic unit structures with ordered channels were built and evaluated by finite element analysis. The photocuring parameters were optimized to improve printing accuracy by establishing an overgrowth model. After optimizing the sintering schedule, the Li2TiO3 cubic units with excellent relative density (92.3%TD) and packing factor (80.99%) were yielded, and it also had an outstanding compressive strength (84.4 MPa). The results of thermal cycling test shown that the Li2TiO3 cubic units possessed a prominent thermal cyclic durability.

Preliminary Development and Verification of Divertor Module for CFETR System Safety Analysis Code

China Fusion Engineering Test Reactor (CFETR), the next-generation fusion device of China, is designed to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and the commercial fusion power plant by solving the essential physical, engineering, and technical problems, including tritium self-sufficiency and high duty cycle of 0.3–0.5. Before the completion of the design and construction of CFETR, the thermal-hydraulic safety assessment is needed in order to ensure the safety of nuclear facilities and to get the national nuclear safety license. Up to now, there is no specialized code for the fusion reactor safety assessment. And the safety assessment work that had been done for the fusion device is completed by the system safety analysis code for pressurized water reactors, such as RELAP5, CHATHARE, and so on. For developing a simulation tool that can meet the safety assessment needs of the fusion reactor, some special models need to be established compared to the fission reactor. In this article, the development of a module for the divertor is presented. The thermal-hydraulic safety of the divertor with design basis accidents, and in-vessel loss of coolant (LOCA), is analyzed based on the developed code. Comparisons are performed to verify the reliability and feasibility of the developed code.

A New Reliability Allocation Method Based on PSA and AHP for Fusion Reactors

Reasonable and feasible reliability index allocation is significant for improving safety and reducing costs for fusion reactors. Although the reliability index is allocated when designing some key systems, reliability allocation is not implemented from the overall layer to the system and component layers for fusion reactors. In this contribution, fusion reactors are divided into different layers by applying the Analytic Hierarchy Process (AHP); these layers include the overall layer, system layer and component layer, failure mode layer, etc. Combining the Probabilistic Safety Assessment (PSA) model for fusion reactors and AHP, the reliability index is allocated from the overall layer to the system and component layer. Firstly, a new reliability index allocation method based on PSA and AHP for fusion reactors is proposed. Secondly, the PSA model for an in-vessel LOCA accident of fusion reactors is selected as a case study to demonstrate the applications of the proposed method. Lastly, the relationship between PSA, items safety classification and reliability allocation are discussed. The allocation results indicate that the proposed method provides a more systematic reliability allocation scheme and contributes to improving the safety and economy of fusion reactors.

The role of grain boundaries and denuded zones for tritium retention in high-dose neutron irradiated beryllium

Neutron irradiation of beryllium leads to a lasting material damage. Besides the displacement damage, large amounts of helium and tritium are produced by neutron induced transmutation. In future fusion reactors the accumulation of large amounts of tritium in beryllium must be prevented due to safety concerns and potential waste disposal problems. It is therefore essential to understand where tritium is trapped and how it is released from the sample during irradiation. Using a simple self-developed theoretical model we show that all tritium which is produced and already released during irradiation originates from the area of later bubble denuded zones (DZ) around grain boundaries (GB). In these areas no or only very few bubbles can be observed since grain boundaries are strong sinks for point defects. The formation of precipitation-bubble-pairs in the immediate vicinity of these regions presumably prevents the release of larger amounts at temperatures ≤ 500 °C, as they act as additional tritium traps. The presented model can be used in the future to estimate the tritium retention in irradiated samples using a small number of easily determined experimental parameters, providing important insights for the blanket safety of future fusion reactors.

Numerical simulation on the eddy current electromagnetic forces and the mechanical reliability of ITER in-wall shielding blocks induced by plasma disruption events

In-wall-shielding (IWS) blocks located in ITER vacuum vessels (VVs) are used for neutron shielding. They are subjected to complex eddy current electromagnetic forces during plasma disruption events, especially for IWSs of irregular sectors. Reliable evaluation of this kind of electromagnetic force is of great significance for the structural design of IWSs as well as the safety of fusion reactors. In this paper, the eddy current electromagnetic forces on 121 IWS blocks in the irregular sector and their associated conductive parts were numerically calculated based on one of the most important disruption events (the MD UP LIN36ms disruption event). To promote the accuracy of the simulations, the excitation loads during the plasma disruption current were processed from the database simulated by DINA code into the FE (finite element) model of plasma with a special transformation procedure. The results show that the method of applying plasma current from DINA to the finite element model is reasonable and accurate. The electromagnetic forces of the blocks are demonstrated to be reliable. The block with the maximum electromagnetic force is selected for mechanical analysis, and the reliability of its components is verified.

Onset of flow instability with one-side heated swirl tube for fusion reactor safety

In this study, the onset of flow instability (OFI) heat flux of a one-side heated swirl tube is experimentally investigated. The OFI heat flux means the minimum heat flux that can cause flow instability by the vapor generated in the flow path. An analysis of the effect of system parameters on the OFI heat flux indicates that as the pressure increases, the bubble size decreases. Therefore, the void fraction decreases and, consequently, the OFI heat flux tends to increase. Similarly, the higher the flow rate and degree of subcooling, the faster the vapor can be removed; thus, the OFI heat flux increases. In addition, the prediction performances of the existing OFI correlations developed under the subcooled flow-boiling condition are evaluated. Therefore, although the Wang correlation indicates the lowest error rate, it yields a high mean absolute error rate of 87.75%. Thus, it is difficult to predict the OFI heat flux of a one-side heated swirl tube using the existing OFI correlations. Therefore, in this study, a new correlation is developed using a Python code created by employing an artificial intelligence regression method. The developed correlation incorporates the impact of one-side heating, swirl tape, mass flow rate, subcooling, and pressure (mean absolute error = 12.17%, root mean square error = 14.99%).

Development of a Novel Two-Phase Pressure Drop Multiplier Correlation of Screw Tube for Fusion Reactor Safety

In this study, the two-phase pressure drop in a one-sided high-heat-loaded screw tube was analyzed under subcooled flow boiling conditions. When the loaded heat flux is gradually increased under one-sided heating conditions, although the pressure drop changes slightly in the single phase, it starts to increase when loading above the onset of significant void (OSV) heat flux. The effect of system parameters on the two-phase pressure drop in relation to the change of heat transfer mechanism or fluid properties was analyzed according to each variable change. In addition, this study evaluated the prediction performance of existing two-phase pressure drop multiplier correlations developed under subcooled flow boiling conditions. Because most existing correlations are based on a smooth channel with a small diameter, they tend to overpredict the experimental values. The correlation proposed by Ramesh showed the lowest average error rate (31.94%); however, the results were underpredicted for high-heat-flux conditions. To obtain more accurate results, we developed a Python code using an artificial intelligence regression method to proceed with the optimization process based on the experimental values. Using this code, we developed a novel two-phase pressure drop multiplier correlation for a one-side-heated screw tube that can be used in fusion divertors under high-heat-load conditions.

Deuterium transport and retention properties of representative fusion blanket structural materials

Reduced activation ferritic-martensitic (RAFM) steels have been developed for decades for use as fusion blanket structural materials, and have advantages in both mechanical properties and irradiation resistance following careful engineering of the microstructure. However, the hydrogen isotope behavior in these proposed fusion structural materials is not well understood, but is important to assess since it impacts the fusion reactor safety and self-sufficient tritium fuel cycle. In this work, we investigated deuterium transport and retention in representative advanced RAFM steels, including castable nanostructured alloys (CNAs), and oxide-dispersion-strengthened (ODS) steels. A gas-driven permeation (GDP) system was used to measure the permeability, diffusivity and solubility of the studied materials, covering the temperature range from 623 K to 873 K, and the loading pressures from 1.8×104 to 1.0×105 Pa. The results indicated that the deuterium permeability has little material dependence. In contrast, the deuterium diffusivity of the studied materials showed significant variation. The deuterium diffusivity in ODS steels is one order of magnitude lower than that in RAFM steels and CNAs, and correspondingly, have an effective solubility that is 2–10 times larger than RAFM steels and CNAs. In addition, thermal desorption spectroscopy (TDS) measurements were performed to assess the deuterium retention and desorption of these materials following a static thermal deuterium charging at 723 Kfor 1 hour under the deuterium pressure of 1.0×105Pa. It was found that ODS steels exhibit the highest deuterium retention and have broader desorption peaks. Microstructural features contributing to deuterium retention and impacting deuterium transport are discussed to rationalize the observed deuterium behavior in the studied RAFM steels.

Investigation on low cycle fatigue properties of CLAM steel electron beam welded joint at 550 °C

The low cycle fatigue (LCF) properties of structural materials at high temperatures are important for operational safety of fusion reactors. In this work, the investigation on LCF properties of CLAM steel joints welded by electron beam was conducted at strain amplitudes varying from 0.2 % to 0.7 % at 550 °C. It is found that the fatigue life and cyclic stress response of welded joints are lower than those of base metal. Continuous cyclic softening is observed in welded joints and base metal during LCF test. A majority of the fatigue cracks initiated at the surface of the the welded joint specimen with several initiation sites and propagated in a transgranular mode. The area of the crack propagation region decreased and the fatigue striation spacing increased with increasing strain amplitude. The cyclic softening is attributed to the coarsening of martensite laths and the reduction in dislocation density.

Quantitative safety goals for fusion power plants: Rationales and suggestions

“How safe is safe enough?” is a key and fundamental question for any risky engineering and safety goals has been extensively studied in many fields, such as chemical industry, nuclear industry, etc. Fusion reactors are advanced nuclear reactors which are expected to be more attractive in terms of safety and environmental impact, and remarkable progress has been made toward the realization of fusion energy in this regard. However, fusion safety is not guaranteed automatically: accidents could happen with any possible future fusion reactor, potentially resulting in releases of radioactive materials. Meanwhile, there are also safety issues about radioactive wastes and occupational exposure, due to high neutron radiation produced in the reactor. To ensure the safety of fusion reactors from the origin, defining safety goals is very necessary to guide reactor design. In this paper, the evolution of safety goals for nuclear reactors is reviewed and a hierarchical framework for fusion quantitative safety goals is established in the four aspects of occupational radiation exposure, routine releases of radioactive materials, accidents and radwastes, based on the characteristics of fusion reactors from a designer's perspective as well as the experiences from the pressurized water reactor, Generation-IV, and International Thermonuclear Experimental Reactor. These safety goals will provide reference for designers of fusion reactors These safety goals will be helpful to guide the designer of fusion reactors to seek and develop appropriate means to enhance safety.

Neutronic analyses of port impact on blankets and superconducting coils of CFETR

• A simplified model of CFETR toroidal field coil is developed. • Scenarios of maximum opened ports are obtained. • Port impact on blanket TBR, radiation damage of blanket first walls and nuclear heat deposition of TF coils are calculated. Chinese Fusion Engineering Test Reactor (CFETR) is an ITER-like test superconducting TOKAMAK fusion reactor. In CFETR, blankets face core plasma directly and are in charge of tritium breeding and neutron shielding, while the condition of superconducting coils needs to be maintained in a low temperature for providing a steady magnetic field to torus. Thus, doing neutronic analyses for blankets and superconducting coils is important for the safety and steady operation of fusion reactor. In addition, ports are used for remote handing, plasma diagnose and other necessary operations, which will cause a different neutron transport behavior in the reactor. In order to evaluate the neutronic performance of CFETR in port opening operations, port impact on blankets and superconducting coils is necessary to be analyzed. In this paper, global TBR and neutron radiation damage for blanket first walls and nuclear heat deposition for toroidal field coils are calculated in different port schemes. Different port opening scenarios are also given in order to meet the requirement of tritium self-sufficiency. According to the result, nuclear heat on the toroidal field coils near the top port exceeds the limited value. Thus, neutronic protection to decrease the leakage of neutron from top port cannot be ignored. Neutronic shielding of superconducting is necessary to be strengthen as well.

Understanding hydrogen retention in damaged tungsten using experimentally-guided models of complex multispecies evolution

Fuel retention in plasma facing tungsten components is a critical phenomenon affecting the mechanical integrity and radiological safety of fusion reactors. It is known that hydrogen can become trapped in small defect clusters, internal surfaces, dislocations, and/or impurities, and so it is common practice to seed W subsurfaces with irradiation defects in an attempt to precondition the system to absorb hydrogen. The amount of H can later be tallied by performing careful thermal desorption tests where released temperature peaks are mapped to specific binding energies of hydrogen to defect clusters and/or microstructural features of the material. While this provides useful information about the potential trapping processes, modeling can play an important role in elucidating the detailed microscopic mechanisms that lead to hydrogen retention in damaged tungsten. In this paper, we develop a detailed kinetic model of hydrogen penetration and trapping inspired by recent experiments combining ion irradiation, hydrogen plasma exposure, and thermal desorption. We use the stochastic cluster dynamics method to solve the system of coupled partial differential equations representing the mean field description of the multispecies system. The model resolves the spatial distribution of defects and hydrogen clusters during the three processes carried out experimentally and is parameterized with information from atomistic calculations. We find that the calculated thermal desorption spectra are broadly characterized by three H emission regions: (i) a low temperature one where dislocations are the main contributors to the release peaks; (ii) an intermediate one governed by hydrogen release from small overpressurized clusters with multiple overlapping peaks, and (iii) a high temperature one defined by clean isolated emission peaks from large underpressurized bubbles. These three temperature intervals are seen to largely correlate with the depth at which the clusters are found. The relevance of the ‘super abundant’ vacancy mechanism is assessed, finding that its main role is to transfer more clusters from the intermediate to the high temperature regions as its relevance increases. We find this picture to be in very good agreement with the experiments, adding confidence to the predictive potential of the models and their useto understand irradiation damage and plasma exposure effects in plasma facing components.

Creep of stainless steel under heat flux cyclic loading (500-1000C) with different mechanical preloads in a vacuum environment using 3D-DIC

In nuclear fusion reactors, the key structural component (i.e., the plasma-facing component) undergoes high heat flux cyclic loading. To ensure the safety of fusion reactors, an experimental study on the temperature-induced creep of stainless steel under heat flux cyclic loading was performed in the present work. The strains were measured using a stereo digital image correlation technique (3D-DIC). The influence of the heat haze was eliminated, owing to the use of a vacuum environment. The specimen underwent heat flux cycles (500 C-1000 C) with different mechanical preloads (0 kN, 10 kN, 30 kN, and 50 kN). The results revealed that, for a relatively large preload (for example, 50 kN), a single temperature cycle can induce a residual strain of up to 15000.

Caution Is Needed in Operating and Managing the Waste of New Pebble-Bed Nuclear Reactors

Rainer Moormann studied physical chemistry in Braunschweig, Germany, obtaining his PhD in 1976. From 1976 to 2013 he worked at Research Center Juelich (FZJ) on the safety of pebble bed reactors, fusion reactors, and spallation sources. In 2011 he received the German Whistleblower Prize for publications on widely unknown problems and accidents in German pebble bed reactors. R. Scott Kemp is the MIT Class of ’43 Associate Professor of Nuclear Science and Engineering and Director of the MIT Laboratory for Nuclear Security and Policy ( http://lnsp.mit.edu ). His research combines engineering, politics, and history in the search for sustainable energy and international security. He obtained a PhD in Public and International Affairs from Princeton University and a Bachelor’s in Physics from the University of California, Santa Barbara. He is the recipient of the Sloan Research Fellowship in Physics and was elected a Fellow of the American Physical Society in 2017. Ju Li is the BEA Professor of Nuclear Science and Engineering and a Professor of Materials Science and Engineering at MIT. He also holds adjunct professorship at Xi'an Jiaotong University, China. His group ( http://Li.mit.edu ) performs computational and experimental research on mechanical properties of materials, and energy storage and conversion. Ju obtained a PhD degree in nuclear engineering from MIT in 2000, and a Bachelor's degree in Physics from University of Science and Technology of China in 1994. In 2016 Ju Li co-founded one of the MIT Energy Initiative Low-Carbon Energy Centers, the Center for Materials in Energy and Extreme Environments. Ju was elected a Fellow of the American Physical Society in 2014 and a Fellow of the Materials Research Society in 2017.

Thermal hydraulic analysis of water cooled blanket under transient condition

Nuclear fusion is known as the most effective way in the future to solve energy problems. And blanket is the key component to achieve the function of energy transfer. In a fusion reactor, the fusion reaction produces high energy neutrons, and the blanket is designed to slow and absorb neutrons in a reliable and efficient manner. Energy absorbed from the fast neutrons is extracted and passed into the primary coolant in the blanket. Hence, the design of the blanket will strongly affect the operating power and the safety of fusion reactors. It is significant to analyze the hydraulic properties and heat transfer properties of the blanket.The object of this study is a water cooled blanket within conceptual design. In the first place, the original complicated design of the blanket is simplified reasonably. Then, RELAP5 is adopted to analyze the thermal hydraulic properties in steady state of the blanket under 100% power level. The system pressure drop along the flow and the temperatures of water, channel wall and blanket structure were discussed. After that, this study analyzed the transient response characteristics of the blanket under transient condition.The results of this study turn out that, under transient condition, the blanket of current design cannot keep safety. To ensure the normal operation of fusion reactors in the future, the current design should be improved.