Nb 3 Sn strands, whose properties are very sensitive to stress/strain, are utilized for ITER cable-in-conduit conductor (CICC) of the central solenoids. The Nb 3 Sn strands experience temperature range of ~1000 K from the temperature of the heat treatment with the initiation of the Nb 3 Sn reaction to the operation temperature of ~4 K. Due to this large temperature range, large thermal strain is induced in the Nb 3 Sn filaments due to the differences between the coefficients of thermal expansion and Young's moduli of the components of the strand. Therefore, it is considered that initial performance of the CICC is influenced by the thermal strain on the Nb 3 Sn, and it is important to evaluate the strain state of the Nb 3 Sn strand at low temperature. In this study, the thermal strain of the components of free Nb 3 Sn strand was measured by neutron diffraction and stress/strain state was assessed from room temperature to low temperature. As the results of diffraction measurements, it was found that 0.111% and 0.209% compressive strain were generated in Nb 3 Sn filaments at 300 and 10 K, respectively.

Synopsis: Austenite stainless steel is used for liquid natural tanks and superconducting facilities since it has a face-centered cubic lattice, which is less likely to decrease its toughness at cryogenic temperatures.The structural materials of the ITER toroidal field coil structure (TFCS) are required to have high fracture toughness at cryogenic temperature (4 K) in order to prevent unstable fracturing by the huge electromagnetic force.Yield strength at 4 K can be accurately predicted pragmatically.However, the estimation method for fracture toughness at 4 K is not yet well developed.In this study, the authors investigated the correlation between several material properties and 4 K fracture toughness of actual sized ITER TFCS materials.As a result, there is a low correlation between 4 K fracture toughness and the parameters (i.e., 4 K yield strength, nitrogen content and grain size), which were thought to be well matched for fracture toughness as reported in previous studies.In contrast, 4 K tensile strength and Md30 are in good correlation with fracture toughness because local transformation into martensite occurring at the crack tip affects fracture toughness.Md30 is used as an index of stability in the austenite phase.The authors therefore established a new method that simplifies controlling 4 K fracture toughness of austenite stainless steel using Md30.In addition, it is demonstrated that this method is effective for actual TFCS materials.The views and opinions expressed herein do not necessarily reflect those of the ITER organization.

The superconducting property of Nb3Sn strands is very sensitive to strain. The transverse electromagnetic loading has been considered as a major origin of the degradation of Nb3Sn cable-in-conduit conductors (CICCs) due to the local bending. Since the bending pitch is around 5 mm due to contacting of strands compacted by the electromagnetic transverse loading, there is a possibility of a large bending strain with small deflection of strands. The bending strain of the strands cannot be evaluated from only the small deflection obtained visually. Measuring bending strain of Nb3Sn strand in CICCs is important for evaluating the conductor performance. Neutrons, which have a large penetration depth, are a powerful tool to evaluate the internal strain of Nb3Sn in the CICC. This paper shows that the bending strain in Nb3Sn strands of CICCs can be determined by the neutron diffraction profile nondestructively and quantitatively.

The ITER Central Solenoid (CS) is composed Nb 3 Sn cable-in-conduit conductor. Short twist pitch cable design is appropriate for the CS cables to maintain the current sharing temperature under electromagnetic loading cycles. In the production process of the short twist pitch CS cable, Nb 3 Sn strands, and Cu strands are cabled tightly and compressed to get circular shape. However, these processes induced strong interstrand contact force and the strands deformation. The strand performances are degraded by the severe indentation on the strands at the contact point between the strands, although the strands are deformed before the heat-treatment. Therefore, influence of the indentation on critical current and residual resistive ratio were investigated in two types of bronze-route and an internal-tin Nb 3 Sn strands which are used for the CS cable, in order to determine the threshold of indentation depth in the strands. Also, transverse cross-sectional observation by Electron Probe Micro Analyzer was carried out on the indented strands to investigate the damage in the cross-section. Before August 2016, 37 CS cables were manufactured. On 37 CS cables manufactured before August 2016, numbers of indentations on the Nb 3 Sn strands were investigated to confirm whether there was indentation on the Nb 3 Sn strands which degrades the strand performance.

The influence of a fast electromagnetic perturbation, such as plasma disruption, on the ITER toroidal field (TF) coil conductor was studied. When a fast magnetic field change is superimposed, the TF conductor is inductively heated by internally generated eddy currents. To measure the inductive heating (IH) energy amount by calorimetry, an IH experiment using short TF conductor samples with length of 20 cm was performed in a liquid He bath. The sample components used were a TF conductor, TF jacket and a TF conductor cable. A 60-turn, single layer solenoid coil was installed around each sample as an IH coil. Also, the Joule heat of the IH coil was solely measured to subtract its thermal contribution from the other samples. A 1 kHz sinusoidal AC current was applied to each IH coil, changing the current amplitude. The heat generated in the samples, including the IH coil, was eventually absorbed into the liquid He, and the liquid He was then vaporized. Thus, the heat amount was measured by a He level sensor inside a gas collection cylinder attached above the sample. The validity of the experimental results was confirmed by comparing them with computation results of the IH energy of the samples with a computation model. Also, the consumed energy was calculated from the measured waveforms of the applied AC voltage and current to the samples. As the result, the measured and calculated IH energy were found to be in good agreement. Finally, based on the results of the experiment, the minimum magnetic field strength, which triggers quench of the TF conductor by fast dumping like plasma disruption, was evaluated using estimated TF conductor minimum quench energy.

Full-scale trials were performed to qualify and optimize the manufacturing procedure of the toroidal field (TF) coil prior to the series production. In the full-scale trials, winding, reaction heat-treatment, conductor transfer, insulation, and cover plate (CP) welding trials were performed to resolve some technical issues and to demonstrate the fabrication procedure. Major requirements are: high accuracy conductor winding of 0.01%; the evaluation of the conductor elongation due to the reaction heat-treatment; the conductor transfer in a radial plate groove without adding more than 0.1% strain to the conductor; the conductor insulation without breakage of the insulation tape; and the flatness of 2 mm of the double pancake (DP) by CP welding. Then, the first TF coil fabrication has been started from March 2014. In this paper, the progress of the series production of TF coils in Japan is reported.

The ITER toroidal field (TF) coil structure is a huge welding stainless steel structure with a height of 16 m and a width of 9 m. The TF coil structure has a function to ensure the enormous electromagnetic force, whereas in ITER operation temperature (4 K). FMYJJ1, which is full austenitic stainless steel welding material having high manganese contents, is applied for manufacturing TF coil structure. However, it was reported that some micro cracks were observed on the reheating zone in the thicker welded joint during trials. Low ductility of deposited metal at 1000 °C causes these microcracks. In addition, segregation of sulfur in grain boundary decreased ductility at high temperature. Hence, reducing sulfur content is effective to reduce microcracks. Two types of FMYJJ1 were developed by applying two reducing sulfur content methods. In the previous study, applicability of these improved FMYJJ1 was tested within welded joints having 20-mm thickness. In order to confirm the validity of applicability of them, the Japan Atomic Energy Agency manufactured welded joints with 200-mm thickness, which is the representative thickness of the TF coil structure. When ductility is increased, there is a concern that yield strength is decreased. Hence, confirmation of the mechanical properties is needed. Therefore, side bend test, tensile test, and fracture toughness test at 4 K were performed. As the result, it was confirmed that the crack sensitivity was improved, and it was also confirmed that mechanical properties at 4 K satisfied ITER design requirements.

The performance of the ITER central solenoid insert (CSI) conductor was tested in 2015. The current-sharing temperatures (Tcs) were measured over 16,000 electromagnetic cycles, including three thermal cycles between 4.2 K and room temperature. Tcs under the initial magnetization (IM) condition (13 T, 40 kA) of the CSI conductor not only increased, but also decreased between 6.71 and 6.84 K against cycling; then Tcs became almost constant at 6.74 K. Thus Tcs under the IM condition, was approximately 1.5 K higher than the specification of 5.2 K throughout the test. The slope of the hoop strain (εhoop) on the CSI conductor against the electromagnetic force was 1.55×10-4 % m/kN (in εhoop > 0) and 1.39×10-4 % m/kN (in εhoop < 0). Taking the effect of εhoop into account, the Tcs of the CSI under the SULTAN simulated condition (11.5 T, 45.1 kA) was equivalent to that of the SULTAN test after around 10,000 cycles. Before around 10,000 cycles, especially at the initial charge, the Tcs of the SULTAN test was lower than that of the CSI test. It is assumed that the hoop strain in the CSI test accelerated a strain relaxation, which increased the Tcs from the initial charge. When the strain fully relaxed and Tcs stopped increasing after around 10,000 cycles, the Tcs of the SULTAN test became equivalent to that of the CSI test. Given this perspective, the CSI test and SULTAN test were consistent. In εhoop > 0, the absolute value of the effective strain (εeff) of the CSI test decreased (i.e., Tcs increased) against the electromagnetic force (Fr) because the effect of the positive εhoop on the increase in Tcs exceeded the effect of the Fr on the decrease in Tcs. The line of εeff −εhoop of the CSI test against Fr was nearly symmetric about the y-axis (Fr=0). Comparing the εeff −εhoop of the CSI test and the εeff of the SULTAN test, the slopes of the strain against Fr were almost the same between the CSI test and SULTAN test before cycling. The εeff of the SULTAN test became close to the εeff−εhoop of the CSI test after cycling. This CSI test demonstrated that mass-produced CS conductors are highly capable of being used in the ITER.

For the stable operation of the Central Solenoid (CS) coil of the ITER without quenching, it is important to know the threshold of allowable external heating energy the CS conductor can be subjected to during operation. To evaluate the minimum quench energy of the CS conductor for the ITER, an inductive heating test was performed during the CS Insert Coil (CSIC) test campaign. A 59-turn inductive heater installed on the central turn of the CSIC was used to apply the heat energy. The heating energy from the inductive heater was calibrated by calorimetry using short conductor samples with inductive heater windings and a resistive heater. A series of inductive heating tests was performed while applying a 45.1 kA current and 12.5 T backup field on the CSIC. The alternating current (AC) applied for the inductive heater was 1,000 Hz in 40 ms, and the amplitude of the AC was varied until a quench occurred. As the result, it was obtained that the minimum quench energy for the CSIC heated by eddy current was 0.23 J/cm3 without including the joule heating energy of the heater itself.

The ITER central solenoid (CS) is a highly stressed magnet that must provide 30 000 plasma cycles under the ITER prescribed maximum operating conditions. To verify the performance of the ITER CS conductor in conditions close to those for the ITER CS, the CS insert was built under a USA-Japan collaboration. The insert was tested in the aperture of the CSMC facility in Naka, Japan, during the first half of 2015. A magnetic field of up to 13 T and a transport current of up to 60 kA provided a wide range of parameters to characterize the conductor. The CS insert has been tested under direct and reverse charges, which allowed a wide range of strain variation and provided valuable data for characterization of the CS conductor performance at different strain levels. The CS insert test program had several important goals as follows. 1) Measure the temperature margin of the CS conductor at the relevant ITER CS operational conditions. 2) Study the effects of electromagnetic forces and strain in the cable on the CS conductor performance. 3) Study the effects of the warmup and cooldown cycles on the CS conductor performance. 4) Compare the conductor performance in the CS insert with the performance of the CS conductor in a straight hairpin configuration (hoop strain free) tested in the SULTAN facility. 5) Measure the maximum temperature rise of the cable as a result of quench. The main results of the CS insert testing are presented and discussed.

The Central Solenoid (CS) insert coil consists of a 42-m-long CS conductor, of which the specifications are the same as that of the ITER CS. In order to investigate normal zone propagation and hot spot temperature, a quench test was carried out on the CS insert under End-of-Burn condition at 12.5 T and 45.1 kA of after 16,000 cycles. External heat was applied at nearly the center of the CS insert using an inductive heater, and quench was induced. A current of 45.1 kA was dumped 9.5 s (7 s) after voltage generation (Quench detection, QD). The Normal zone propagation length reached 23.4 m, and the maxim propagation velocity was 3.1 m/s just before dumping. Considering the distribution of temperature, which is calculated by GANDALF, hot spot temperature was expected to reach 227 K. As the result, it was found that the hot spot temperature exceeded the criteria of 150 K which is designed on ITER. However, heating the CS insert to 227 K did not influence conductor performance, because the current sharing temperature was maintained after the quench test. Therefore, the quench detection has a margin of approximately 9.5 s (7 s) after voltage generation (QD) in view of the conductor performance under the conditions applied in this quench test. If the hot spot temperature is kept to less than 150 K, the current should be dumped 7.5 s (5 s) before voltage generation (QD). These results are very useful for designing quench protection of the ITER CS.

A suite of advanced austenitic stainless steels are used for the ITER TF, CS and PF coil systems.These materials will be exposed to cyclic-stress at cryogenic temperature. Therefore, high manganese austenitic stainless steel JK2LB, which has high tensile strength, high ductility and high resistance to fatigue at 4K has been chosen for the CS conductor. The cryogenic temperature mechanical property data of this material are very important for the ITER magnet design. This study is focused on mechanical characteristics of JK2LB and its weld joint.

Aramid fiber-reinforced plastic (AFRP) has been applied in severe environmental conditions, such as aeronautical and space environments, etc. AFRP rod has been particularly developed as a structural material that has the advantages of light weight and high strength. Therefore, it is necessary to examine its strength in various environments. In this study, tensile tests were carried out to measure the tensile properties of AFRP rod on the market for reinforcement of concrete at room temperature, liquid nitrogen temperature (77 K) and liquid helium temperature (4.2 K). Especially at cryogenic temperatures, it is difficult to perform a tensile test of the bar because the specimen slips through the jig grip. To prevent the AFRP rod from slipping, tensile tests were carried out with some filling conditions. The applicable and appropriate tensile test conditions were established by modifying the jig grip, treating the surface of the rod and using cryogenic epoxy infill to grip the AFRP rod. Additionally, the effects of cryogenic temperatures on the tensile properties were evaluated. From the tensile tests, the tensile strength decreased about 7% at 77 K and about 15% at 4.2 K as compared to room temperature. However, they were more than 1100 MPa. Additionally, the AFRP rod included a temperature dependence in which the Young's modulus increased as the test temperature decreased. The increased was approximately 60% at 77 K and 4.2 K as compared to the room temperature measurement. From the results of dynamic viscoelastic measurement, it was confirmed that the Young's modulus increased because aramid fiber was more dominant than epoxy.

The Japan Atomic Energy Agency (JAEA) has developed the prediction method for yield stress and ultimate tensile strength at liquid helium temperature (4K) using the quadratic curve as a function of the content of carbon and nitrogen. Prediction method was formulated based on the tensile strength data of materials with shape of rectangle. In this study, tensile strength of the forged materials with round bar and complex shape were obtained so as to compare with the predicted value. The accuracy of the prediction method was 10.2% of Yield Strength (YS), 2.5% of Ultimate Tensile Strength (UTS) when the prediction method was applied to round bar forged materials. By contrast, the accuracy about prediction method was 1.8% of YS, -0.8% of UTS when prediction method was applied to complex shape forged materials. It can be presumed the tendency of tensile strength other than materials with shape of rectangle. However, it was found accuracy of round bar is larger than other materials because of difference in the forging method.“The views and opinions expressed herein do not necessarily reflect those of the ITER Organization”

Following the recent improvement of the heat treatment pattern for the ITER CS conductor from 250 hour at 650°C to 250 hour at 570°C and then 100 hour at 650°C, the mechanical properties at 4K and the material structure of the conductor jacketmade of JK2LB after heat treatment including the 570°C phasewere studied. The mechanical test results at 4K showed no noticeable difference with the jacket heat-treated only at 650°C to the one additionally heat-treatedat 570°C.Both satisfied the ITER requirement. The metallographic test results also did not show large difference. In conclusion the 570°C step does not affect noticeably the JK2LB jacket and the improved heat treatment pattern is applicable.

ITER toroidal field coil (TFC) structures are large welding structures composed of coil case and support structures made of heavy thick high strength and high toughness stainless steels. Japan Atomic Energy Agency plans to apply narrow gap Tungsten Inert Gas (TIG) welding with FMYJJ1 (0.03C-10Mn-12Cr-14Ni-5Mo-0.13N) which is full austenitic stainless filler material. In order to evaluate effect of base material thickness and combinations of base material on tensile properties, tensile tests were performed at room temperature, 77K and 4K by using tensile specimens taken from 200 mm thickness welded joints of two combinations of base materials and 40 mm thickness welded joints of four combinations of base materials. As the results, it was confirmed that there were no large distribution of yield and tensile strength along the thickness of welded joints of 200 mm thickness and yield and tensile strengths of welded joints were decreased with decreasing of C + N contents of base material. (C) 2013 Published by Elsevier B.V.

A prediction method for both yield and tensile strengths of stainless steels from room temperature to liquid helium temperature (4K) has been developed by JAEA in order to rationalize qualification tests of cryogenic structural materials used in large superconducting magnet for a fusion device. This method is to use quadratic curves which are expressed as a function of carbon and nitrogen contents and strengths at room temperature. This study shows results of tensile tests at 4K and confirmation of accuracy of prediction method for tensile strengths at 4K for large forgings and thick hot rolled plates of austenitic stainless steels, which can be used in the actual coil case of the International Thermonuclear Experimental Reactor (ITER) toroidal field (TF) coils. These products are 316LN having high nitrogen and maximum thickness is 600mm. As the results, it was confirmed that the materials which satisfied ITER design requirement can be manufactured and the tensile strengths of these products at 4K can be predicted by using appropriate quadratic curves.The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

The ITER Central Solenoid (CS) conductor is composed of 576 superconducting strands and 288 Cu strands assembled together into a multistage cable and protected by a circle-in-square jacket with the outer dimension of 49 mm × 49 mm. In R&D to prepare for the ITER CS conductor manufacturing, mechanical tests of jacket, welding tests and manufacturing of 181-m long dummy conductor have been performed. In this paper, the R&D activities are presented, showing that as a result of this R&D, the CS conductor manufacturing technologies have been preliminary defined to start the procurement of the CS conductor.

Twenty-five percent of the total toroidal field (TF) coil conductors at the ITER are supplied by the Japan Atomic Energy Agency (JAEA). The jacket section of a TF conductor is made of modified 316LN. The JAEA tested three types of tensile specimens (Japanese-Industrial-Standards-type and ASTM-type) cut from the jacket at 4.2 K. The ASTM-type specimen had a longer and wider reduced section than did the JIS-type specimen. The results of the test showed that the EL of the as-received (AR) jacket was independent of the specimen shape. However, after cold working and aging, the EL of the specimens deteriorated because of sensitization, and the EL distribution in these specimens was larger than that in the case of the AR specimens. It could be inferred that the shape of a test specimen having low ductility is the key determinant of the specimen's susceptibility to fracture.