Test Reactor Research Articles

Multi-Scale Characterization of Porosity and Cracks in Silicon Carbide Cladding after Transient Reactor Test Facility Irradiation

Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) cladding for light water reactor applications. The morphology of fabrication defects, including the size and shape of voids, is one of the key challenges that impacts cladding performance and guarantees reactor safety. Therefore, quantification of defects’ size, location, distribution, and leak paths is critical to determining SiC CMC in-core performance. This research aims to provide quantitative insight into the defect’s distribution under multi-scale characterization at different length scales before and after different Transient Reactor Test Facility (TREAT) irradiation tests. A non-destructive multi-scale evaluation of irradiated SiC will help to assess critical microstructural defects from production and/or experimental testing to better understand and predict overall cladding performance. X-ray computed tomography (XCT), a non-destructive, data-rich characterization technique, is combined with lower length scale electronic microscopic characterization, which provides microscale morphology and structural characterization. This paper discusses a fully automatic workflow to detect and analyze SiC-SiC defects using image processing techniques on 3D X-ray images. Following the XCT data analysis, advanced characterizations from focused ion beam (FIB) and transmission electron microscopy (TEM) were conducted to verify the findings from the XCT data, especially quantitative results from local nano-scale TEM 3D tomography data, which were utilized to complement the 3D XCT results. In this work, three SiC samples (two irradiated and one unirradiated) provided by General Atomics are investigated. The irradiated samples were irradiated in a way that was expected to induce cracking, and indeed, the automated workflow developed in this work was able to successfully identify and characterize the defects formation in the irradiated samples while detecting no observed cracking in the unirradiated sample. These results demonstrate the value of automated XCT tools to better understand the damage and damage propagation in SiC-SiC structures for nuclear applications.

Simulation study of a rhodium self-powered neutron detector for irradiation of nuclear fuel and material reactor

To accurately judge the irradiation test cycle of nuclear fuel and material in-pile irradiation, and provide precise design inputs for the post-irradiation performance evaluation of fuel and material, the NIT-Rh self-powered neutron detector has been designed and developed. The sensitivity characteristics of the detector have been simulated and verified. This study introduces the basic structure and working principle of the rhodium self-powered neutron detector for irradiation of nuclear fuel and materials in the reactor. Additionally, a new computational model has been constructed by the Monte Carlo method, and the current sensitivity of different sizes of emitters and insulators has been analyzed. The model considers a more realistic neutron source and β-electron energy spectrum, accounts for the space charge effect on the signal in the insulator, and compares the simulated sensitivity with the reactor test sensitivity. The results show that the new rhodium self-powered neutron detector calculation model is more accurate, and the relative deviation between the simulated and reactor test sensitivities is only 3.06 %. It can be used for the in-heap irradiation test and provides a method for the application of the irradiation of nuclear fuel and material reactors.

Design and analyses on the recuperator of the high temperature PbLi experimental platform for CFETR

As a candidate blanket for China Fusion Engineering and Test Reactor (CFETR), the supercritical carbon dioxide (sCO2) cOoled Lithium-Lead (COOL) blanket is under development. To solve the key scientific issue, i.e. MagnetoHydroDynamics (MHD) effect, the high temperature PbLi experimental platform is planned for construction at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). As one of the key components in this platform, the recuperator acts as waste heat recovery and saving electricity power. In this paper, two types of recuperator are designed and analyzed, including the double-tube and hairpin. Firstly, the theoretical and empirical models are used to obtain different designs by changing the structure dimensions and arrangement, then the non-dimensional evaluation criteria is improved and adopted to determine the optimal design that has better comprehensive performance, i.e. higher heat transfer capacity, lower flow resistance to reduce pressure drop and smaller consume volume of liquid metal to increase economy. Finally, the thermal-hydraulic and thermal-mechanical analyses based on the optimal design of the 3D model are further carried out, which obtains the fields of temperature, pressure, velocity as well as stress, and the results can satisfy with the multiple requirements, i.e. outlet temperature under different operating conditions, flow distribution as well as structural strength. The design methodology and analysis results will provide useful technical and data support for the following engineering design and construction of the PbLi experimental platform.

Effect of metallic redox ions on the corrosion behavior of GTAW weldment of type 304L stainless steel in simulated non-radioactive PUREX and HLLW medium

Nuclear-spent fuel reprocessing and high-level liquid waste (HLLW) storage containers possess various inorganic higher valance oxidizing metallic ions and fission products, further influencing nitric acid corrosion through the PUREX route. This study evaluated the effect of nitric acid corrosion on gas-tungsten arc weldment (GTAW) of stainless steel (SS) 304 L in simulated reprocessing and HLLW medium in non-radioactive conditions. The effect of oxidizing metallic cations, which were present as the major amount in the mark-I spent fuel of fast breeder test reactors (FBTR) in India; these are Cs+1, Nd+3, Mo+6, Fe+3, and V+5 as a surrogate redox system for Pu(VI)/Pu(IV) and Np(VI)/Np(V) system was evaluated by the electrochemical corrosion test method in 2 M, 6 M, and 8 M nitric acid solutions. The resistance of polarization (Rp) value decreases with the nitric acid concentration, and the effect of oxidizing ions in different nitric acid concentrations also changes. The three-phase corrosion (boiling liquid, vapor, and condensate phase) resistance behavior in different concentrations (6 M and 8 M) of the nitric acid medium, without and with redox ions, has been evaluated. The corrosion rate is higher in the liquid phase compared to the vapor and condensate phases. On the other hand, it is relatively higher for the inorganic metallic redox ions containing solutions.

Analysis of loss of flow without scram test in the FFTF reactor – Part I: Preparation of neutronics data

This study presents a benchmark analysis of an unprotected loss of flow transient in a sodium-cooled fast reactor at the Fast Flux Test Facility (FFTF), carried out as part of an IAEA coordinated research project. Three codes, namely Serpent (Monte Carlo), DYN3D (3D nodal diffusion) and ATHLET (system thermal hydraulics), were employed in the benchmark exercise. Two distinct modeling approaches were utilized: 1) stand-alone ATHLET with point kinetics (PK) and system thermal hydraulics (TH); and 2) coupled DYN3D/ATHLET with spatial kinetics (SK) and system TH. Neutronics data essential for both approaches were generated using Serpent. The study is organized into three parts.Part I presents a summary of the preparation of neutronics data for PK or coupled SK/TH simulations and includes the outcomes of the static neutronics stage of the benchmark. The main focus lies on verifying the cross-section generation method for DYN3D by comparing its results against the reference Monte Carlo solutions obtained with Serpent.Part II will provide a detailed description of the ATHLET TH model of the system. This model will be thoroughly evaluated by comparing the results with the ANL benchmark solution and experimental data for the transient. Furthermore, a sensitivity analysis will be conducted to explore various modeling options and assess their impact on the simulation results.Part III will showcase the transient calculation results using the two modeling approaches. Additionally, an adaptive decay heat model for nodal codes will be introduced. The performance of both modeling approaches will be assessed by comparing their results to the available experimental data.

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), 252Cf is used as the external neutron source. However, the 252Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of 252Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the 252Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s−1.cm−2 even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. The ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

Numerical study of the flow and heat transfer characteristics in a scale model of the vessel cooling system for the HTTR

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models’ execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system’s size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier–Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

Load study and structural analysis for CFETR divertor OVT

China Fusion Engineering Test Reactor (CFETR) is an important device for studying fusion reactor engineering in China. The Outer Vertical Target (OVT) are high heat load components of the divertors and work in a strong electromagnetic field, so they need sufficient heat removal capability and structural strength. In addition, due to activation of neutrons on materials, the OVT also needs to meet the requirements of Remote Handling (RH) maintenance. CFETR puts forward the concept of hybrid divertor-blanket, which cancels the cassette and adopts a separate structure. The lower part of the divertor is occupied by the breeding blanket, and the divertor target is fixed on the breeding blanket, to improve the RH maintenance efficiency and Tritium Breeding Ratio (TBR). In previous work, the structural design and thermal-hydraulic analysis of the OVT were completed. In this paper, load study and structural analysis was carried out to verify the structural strength of the OVT under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (Vertical Displacement Event (VDE) or Major Disruption (MD) event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.). The results show that the OVT meets the requirements of heat removal and strength, and effectively promote the processing and manufacturing of the prototype of the OVT.

Thermal hydraulic optimization on the First Wall of COOL blanket for CFETR

The supercritical carbon dioxide (S-CO2) cOoled Lithium-Lead (COOL) blanket is under development for China Fusion Engineering and Test Reactor (CFETR). The S-CO2 is adopted to cool the structural components, in which the key component First Wall (FW) dominates the total heat power due to the large plasma facing heat flux. In this paper, referring to the response of material temperature, coolant velocity and pressure drop, the thermal hydraulic optimization is firstly performed on the structure design (i.e. cross area and pitch of channel) using the single variable control method, and the optimization principle is that the outlet temperature remains as the same when the optimization variables are changed. Finally, the optimal structure design of FW is obtained, namely cross area of 12 × 12 mm2 and channel pitch of 5 mm. On basis of this, the comparative analyses are carried out using the different cooling system, structural material, i.e. Reduced Activation Ferritic Martensitic (RAFM) steel and Oxide dispersion-strengthened alloy (ODS) ferritic steel, as well as helium as coolant. Furthermore, the satisfied designs are obtained by referring to the constraints of coolant pressure drop and maximum temperature on the structural steel. Then, the thermal hydraulic designs of FW are continuously matching with the Power Conversion System (PCS) to ensure that the heat power removed by the S-CO2 on the primary side can be transferred to the secondary side. Finally, the achievable thermodynamic design are obtained, including the parameters of high operating pressure, turbine inlet temperature, split ratio as well as the thermoelectricity conversion efficiency. The results show that the high pressure of PCS should be larger than 25 MPa when the RAFM steel is used as the structure material, due to the limitation on the outlet temperature of FW. On the contrary, the ODS can raise the outlet temperature, which is beneficial for matching with the PCS, and the thermoelectricity conversion efficiency can reach up to 47.4 % at the high pressure of 25 MPa. The analyses results will provide useful guidance on the system design both for the Primary Heat Transfer System (PHTS) and PCS of CFETR.

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235U content. Finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm2. No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

A STUDY ON THE POWER VARIATION OF FUEL SUBASSEMBLIES CLOSER TO THE SOURCE SUBASSEMBLY OF THE 40MWT FBTR CORE

Fast Breeder Test Reactor (FBTR) which is a loop type, sodium cooled fast reactor, is one of the high flux - high temperature reactor in India with design power of 40 MWt (13.2 MWe). The design target power of 40 MWt was achieved with a core configuration with 68 Mark I (MK I) fuel subassemblies. A study has done to evaluate the variation in the power of fuel subassemblies which are closer to the source subassembly. Source subassembly in FBTR is a gamma neutron source (Sb2O3-Be) located in (05, 28) location. The study found that the presence of the source subassembly increases the power of the neighbouring fuel subassemblies. The variation in power due to the source subassembly is found to be 10% for fuel subassemblies in (05, 29) location and for other fuel subassemblies in 3rd, 4th & 5th ring, it is within 3.5%. The estimated LHR and power of the fuel subassemblies closer to the source subassembly are well within the permissible limits.

Thermohydraulic and thermal fatigue characteristics of W/Cu flat-type microchannel mock-up for CFETR divertor

The divertor of the China Fusion Engineering Test Reactor (CFETR) requires a more efficient heat dissipation structure due to the high demand of heat flux dissipation of up to 20 MW/m2. Microchannel cooling technology excels in thermal management because of large heat transfer area to volume ratio and capability to achieve uniform temperature distribution. However, the heat transfer and thermal fatigue characteristics of W/Cu flat type mock-up with microchannel cooling structures at a heat flux of 20 MW/m2 remains elusive. In this work, we analyzed the thermohydraulic (numerically) and thermal fatigue (experimentally) characteristics of W/Cu flat-type microchannel mock-up. The numerical results show that the thermal loaded area (W) of the microchannel mock-up possesses excellent temperature uniformity and heat transfer performance (the W surface temperature is only ∼685 °C even at a low volumetric flow rate V̇ = 5 L/min). We also compared the heat dissipation efficiency with the hypervapotron cooling channels, and found that the maximum temperature of the W surface and channel wall of the microchannel mock-up are reduced by 18.9 % and 46.7 %, respectively. The thermal stress analysis indicates that the maximum equivalent stress is concentrated in the middle region of the W surface at 15 s, with a value of 357.8 MPa that is below the maximum allowable stress of W material (425 MPa at 650 °C). The as-manufactured W/Cu flat-type microchannel mock-up successfully withstood a thermal fatigue test consisting of 1000 cycles at 20 MW/m2 steady state heat load at V̇ = 30 L/min and Tin = 25 °C. The maximum temperature of W is only 616.2 °C during the test, and some microcracks are detected on the W surface after the test; these microcracks are perpendicular to the heat loading surface, which have negligible influence to the heat transfer of the entire mock-up. This study reveals that the W/Cu flat-type microchannel mock-up is promising in dissipating ultra-high heat flux (steady-state 20 MW/m2) on the divertor target, which might provide valuable insights for the engineering design of the CFETR divertor.

Investigation of alpha-particle transport by Alfvén eigenmodes in CFETR using a simplified diffusion model

The aim of this work is to analyze the alpha particle transport induced by Alfvén eigenmodes (AEs) based on the latest design of China Fusion Engineering Test Reactor (CFETR). Firstly, the stability of AEs is analyzed in order to understand the physical characteristics of AEs in CFETR with latest design parameters. AEs driven by alpha particles are analyzed using the gyrokinetic ion/fluid electron hybrid code GEM. The transport of alpha particles is investigated using a reduced energetic particle transport model based on the resonance broadened quasi-linear model. GEM results show that AEs with many toroidal mode numbers can be destabilized simultaneously by alpha particles in CFETR, and the toroidal mode number of the most unstable mode is n=8 . It is found that the excitation threshold of the alpha particle central beta for the n=8 mode is substantially below the expected alpha particle beta value of CFETR. All these results indicate that AEs can be driven strongly in CFETR plasmas. Furthermore, simulation results using the reduced transport model show that, due to multiple unstable AEs, the alpha particle density decreases in the core and alpha particles are transported from the core region ( r/a<0.35 ) to the outer region ( 0.35$?> r/a>0.35 ). In addition, we find that the minimum value of the safety factor qmin and the central beta value of alpha particle have significant effects on the alpha particle transport, and the AE-induced alpha particle transport level becomes lower with smaller linear growth rates. Finally, it is found that the radial range of the redistributed alpha particle density profile depend on the qmin radial location.

Phenomena Identification and Ranking Table for Thermal-Hydraulic Safety of China Fusion Engineering Test Reactor

The China Fusion Engineering Test Reactor (CFETR) is presently in the engineering design phase, and it is crucial to establish a thermal-hydraulic safety analysis method. First of all, the identification of phenomena in selected accident scenarios affecting reactor safety performance metrics should be focused on with the phenomenon identification and ranking table (PIRT), in which phenomena related to the thermal-hydraulic safety of the vacuum vessel and related systems are screened, including both design-basis accidents and beyond-design-basis accidents. The development of this PIRT exercise for CFETR thermal-hydraulic safety is addressed in this study; specifically, the importance ranking and stage of knowledge (SoK) of phenomena in the selected accidents are evaluated by the PIRT panel. The results of the PIRT analysis revealed that there exist certain safety-critical phenomena for which SoK is relatively low despite their perceived significance in safety performance metrics, which include the phase change of coolant, the migration of the multicomponent gas mixture, the migration of radionuclides within confinement, and the dispersion of gaseous radionuclides in the atmosphere. Ongoing research works and follow-up plans to improve the SoK of phenomena are presented.

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

Long plasma duration operation analyses with an international multi-machine (tokamaks and stellarators) database

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

Thermal-hydraulic simulation of the CFETR TF coil operation during a plasma scenario: Gandalf-Fluent and 4C code-to-code comparison

Within the framework of the comprehensive research facility for fusion technology (CRAFT) activities, a prototype toroidal field (TF) coil has been designed and will be manufactured at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), in Hefei. The main purpose is to develop and validate the manufacturing technologies of the large-scale superconducting coils of the China Fusion Engineering Test Reactor (CFETR). To ensure the safe operation of the CFETR device, the Gandalf-Fluent tool and the 4C code are employed to evaluate the performance of the superconducting magnets, and their thermal-hydraulic simulation results are compared with each other. The comparison shows that the thermal coupling between turns and pancakes and the transient model of the thermal coupling to the casing structures, accounted for in 4C but neglected in Gandalf-Fluent, have a significant effect on the temperature profile. Then a detailed reassessment of the minimum temperature margin (ΔTmarmin) is carried out using the 4C code. The results show that the ΔTmarmin is ∼ 1.0 K (if a realistic value for the inter-turn and inter-pancake coupling parameter is assumed), lower than the design requirement of 1.2 K, and it deserves the attention of the designers.

The dynamic factor particle swarm optimization (DFPSO) to inverse kinematics of CMOR based on pose decomposition strategy

The Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting Tokamak device, whose components would get gradually damaged by being susceptible to high temperature, strong magnetic field, radiation, and so on, and should be maintained regularly. The CFETR Multipurpose Overload Robot (CMOR) is a redundant robot designed as a macro-micro arm structure for a gigantic workspace in the blanket and diverter of CFETR's maintenance. Due to the macro-micro arm structure, the robot has a high degree of freedom, and its serial DOF is up to 16 so the traditional algorithm is low efficient for the robot's kinematics. Hence, in this paper, a kinematic decomposition strategy is proposed to allocate or decouple the position aim. Besides, a novel particle swarm optimization (PSO) with dynamic accuracy called dynamic factor particle swarm optimization (DFPSO) is designed to accelerate solving the high-dimensional nonlinear kinematic problem and to improve accuracy. The results of the research show the validation and efficiency of DFPSO in the CMOR system's kinematics. Moreover, it's also a universal kinematic algorithm for all high-DOF robots.

Study on the welding properties of modified N50 CICC jacket for future fusion applications

The modified N50 (N50H) cable-in-conduit conductor CICC jacket developed by China will be applied in China Fusion Engineering Test Reactor (CFETR). Building upon the N50H material, China has developed a corresponding welding material and conducted R&D work on N50H jacket welding. We simulated the N50H jacket welded joint to replicate the entire process of CICC preparation, which includes extrusion, bending, straightening and aging. The magnetic properties of welded joints were evaluated at various temperatures using a physical property measurement system. In addition, the microstructure of the welded joint was studied by various microscopic analysis techniques, and the welded joint showed a stable austenitic structure. Furthermore, mechanical properties of the welded joints were investigated at different temperatures with particular emphasis on their behavior at 4.2 K. The circle-in-square ReBCO conductor jackets welding joint showed a yield strength greater than 1500 MPa, a fracture toughness KIC better than 260 MPa m1/2 and the elongation at break is more than 30 % at 4.2 K. However, a decrease in strength, ductility, and fracture toughness is observed within the welded joints of Nb3Sn conductor jackets after undergoing an aging process. This study will present experimental data on welding joints and discuss the feasibility of N50H as a high-magnetic field jacket material for next-generation fusion reactors.