Reduced Activation Ferritic Steel Research Articles

Compatibility of reduced activation ferritic steel wall with high performance plasma on JFT-2M

In JFT-2M, the applicability of reduced activation ferritic steel to a fusion demonstration reactor has been demonstrated mainly concerning the ferromagnetic effect on plasma production, control, stability and confinement. The compatibility of the close ferritic wall with high-normalized beta plasmas has been investigated mainly because the demonstration reactor will utilize the wall stabilization effect. Due to improvements in the operation scenario and the modification of the guard limiter configuration, the region showing good compatibility was extended to the closer wall position (the wall position was normalized to the plasma minor radius; rwall/a∼ 1.25) and the higher normalized beta (βN ∼ 3.5), compared with the previous works with the relatively far wall position and stronger magnetic field (Tsuzuki K. et al 2003 Nucl. Fusion. 43 1288). A reduction in the growth rate of the instability was also observed in the close wall case, which presumably corresponds to the wall stabilization effect, similar to a resistive wall without ferromagnetism.

Characteristics of Plasma Operation with the Ferritic inside Wall and Its Compatibility with High-Performance Plasmas in JFT-2M

Compatibility between plasma and reduced activation ferritic steel, which is the leading candidate for the structural material of a fusion demonstration reactor, has been investigated in the Advanced Material Tokamak EXperiment (AMTEX). Ferritic plates (FPs) were installed progressively in the JFT-2M tokamak. The effect of ferromagnetism on plasma production, control, confinement, and stability has been investigated. Impurity release behavior has also been investigated. Even when the inside vacuum vessel wall was fully covered with the FPs and the tokamak plasma was operated close to the wall, no deleterious effect was observed, and the normalized beta could be increased up to ~3.5. Thus, encouraging results are obtained for application of this material to the demonstration reactor.

Consideration on blanket structure for fusion DEMO plant at JAERI

Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5 MW/m 2 with peaking factor of 2, a neutron wall load of 3.5 MW/m 2 with peaking factor of 1.5 and a neutron fluence of about 10 MW/m 2. To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li 2TiO 3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510 °C is chosen as coolant material, with coolant pressure of 25 MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

An overview of dual coolant Pb–17Li breeder first wall and blanket concept development for the US ITER-TBM design

An attractive blanket concept for the fusion reactor is the dual coolant Pb–17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb–17Li is circulated for power conversion and for tritium breeding. A SiC f/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb–17Li and as the thermal insulator to separate the high temperature Pb–17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb–17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb–17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

Overview of design and R&D of test blankets in Japan

Japan is performing design and technology developments for the purpose of module testing and contribution to the module development and testing under the framework of Test Blanket Working Group (TBWG) with involvements of all of Japan Atomic Energy Research Institute (JAERI) and universities and National Institute for Fusion Science (NIFS). As the primary blanket option, solid breeder test blanket modules (TBMs) with reduced activation ferritic steel structure is being developed and proposed to be delivered on the first day of ITER operation, mainly by JAERI. As for the advanced blanket options, mainly universities and NIFS have been performing R&D of key technologies and design concept development of solid breeder blanket test article made of SiC composite contained inside the ferritic steel box, liquid LiPb breeder blanket TBM cooled by helium and its dual-coolant option, liquid Li self-cooled blanket TBM and molten salt self-cooled TBM. In all necessary fields of the development of the primary blanket option, the element technology development phase has been almost completed and is now stepping further to the engineering test phase. The development of advanced test blankets is also showing steady progress to overcome key issues toward module testing.

Deformation behavior of reduced activation ferritic steel during tensile test

Deformation behavior of reduced activation martensitic steel F82H during tensile tests were studied. True stress–true strain diagrams were calculated with minimum diameter determined from the specimen profile obtained by laser micro-gauge scanning the diameter along the longitudinal direction during tensile test. Cylindrical specimens of F82H were used for the measurement and test temperatures were room temperature (RT), 300, 400, 500 and 600 °C. Tensile tests were carried out with 1 × 10 −4 s −1 of strain rate. Other strain rates (1 × 10 −3 and 1 × 10 −5 s −1) were applied for the tests at RT. Although uniform elongation of F82H is relatively small at elevated temperature, true stress increases to fracture after necking starts. True stress decreases temporarily after yielding at 600 °C, but it increases again to fracture like the specimens tested at lower temperatures. Influence of strain rate to true stress–true strain relationship at room temperature was small, but unstable deformation occurred in narrower area at higher strain rate.

Neutronics design of the low aspect ratio tokamak reactor, VECTOR

A neutronic study has been carried out for VECTOR, which is a low aspect ratio tokamak with the aspect ratio of 2.3, to achieve simultaneously sufficient shielding for super-conducting toroidal field coils (TFCs) and tritium breeding ratio (TBR) larger than unity, using the three-dimensional Monte Carlo neutron transport code MCNP-4C. The tritium production is mainly made by outboard blankets consisting of the self-cooled lithium lead (LiPb) as tritium breeding and neutron multiplying material, and the reduced activation ferritic steel F82H as structural material. The inboard blanket consisting of vanadium hydrate (VH 2) and F82H as structural material plays the role of radiation shield for the TFCs. Adding Be layers to the LiPb layers effectively improves the TBR. By employing a 0.8 m thick inboard blanket with a 0.13 m thick LiPb layer, sufficient shielding for the inboard TFC and a total TBR of 1.08 are achieved.

Ferritic steel-blanket systems integration R&D—Compatibility assessment

The reduced activation ferritic steel (RAFS) has been selected as structural material for a variety of blanket systems for ITER test blanket modules (TBM). In the evaluation of integrated performance of ferritic steels as structural components of blanket systems, there are unique issues as well as common issues for each blanket system. One of the unique issues for each system is the compatibility of ferritic steels with the coolant materials. The corrosion rate of ferritic steels in hot water, super critical pressurized water (SCPW), humid air, Pb–17Li, lithium and Flibe at various temperatures is reviewed in this work. Efforts to improve corrosion resistance have been made, taking the alloy design into account. A dispersion of yttria was effective to improve corrosion resistance of a RAFS. The compatibilities of RAFSs with hot water, Pb–17Li, lithium and Flibe are considered to be good enough for the TBM applications. The liquid metal embrittlement (LME) is considered to be a critical issue for the utilization of RAFSs for the lithium systems. Several issues towards DEMO and beyond are shown from the compatibility point of view.

Exposure of reduced activation ferritic steel F82H to TEXTOR plasma

Reduced activation ferritic steel, which is a leading structural material candidate for a fusion demonstration reactor, has been installed in TEXTOR and exposed to neutral beam heated plasma. When the heat load was lower than 5 MW/m 2, the insertion of the test limiter did not influence the plasma parameters. Visible spectroscopy near the surface showed that the emission related to constituent element such as iron and chromium is comparable to that from the stainless steel. It means that the sputtering from the ferritic steel is comparable to that from the stainless steel even for plasma exposure in high heat flux condition. When the heat load was ∼10 MW/m 2, the limiter was melted slightly. The strange motion of melted layer was observed by 2D camera. It slowly moved (∼0.5 cm/s) in a spiral motion, forming a small droplet ∼1 mm in diameter which was then released. The released droplet turned toward the limiter and hit the surface. The mechanism of this motion is discussed.

Blanket–plasma interaction in tokamaks: Implication from JT-60U, JFT-2M and reactor studies

Optimization of blanket–plasma interaction is one of important subjects of tokamak reactor design. This paper summarizes key physics R&D results of blanket–plasma interactions in tokamak system based on JT-60U, JFT-2M and DEMO studies at JAERI. Conflicting requirement between wall stabilization and disruption tolerance with respect to the electromagnetic coupling between plasma-structure seems to have a solution with somewhat distant wall location while minimization of the distance between plasma and stabilizing wall is critically important for high beta operation. Effect of ferromagnetism on plasma stability is well studied in JFT-2M and theoretically giving positive prospect for use of reduced activation ferritic steel as a first wall material. Enhancement of ion/neutral 1st wall interaction with 2nd SOL and saturated wall is discussed as an important R&D element for DEMO and existence of high-energy neutrals in JT-60U was mentioned.

SPクリープ試験法による低放射化フェライト鋼のクリープ特性評価

SPクリープ試験法による低放射化フェライト鋼のクリープ特性評価

Small specimen test technique for evaluating fracture toughness of blanket structural materials

Fracture toughness of blanket structural materials is one of the important material engineering properties to design structural components and evaluate the degradation during reactor operation. The present study showed that the master curve (MC) methodology to evaluate change in fracture toughness in the cleavage regime of a reduced-activation ferritic (RAF) steel before and after a thermal embrittlement treatment was applied by using sub-sized compact tension (CT) specimens. The reference temperature of MC for the JLF-1LN steel was estimated as −52 °C by using 1/2CT specimens. The shift of the reference temperature obtained by the MC methodology was judged to be somewhat larger than that of the ductile–brittle transition temperature shift obtained by the Charpy impact tests.

Tritium breeding experiments with blanket mock-ups containing [formula omitted]-enriched lithium titanate and beryllium irradiated with DT neutrons

Experiments have been conducted with breeding blanket mock-ups consisting of layers of candidate materials for JAERI’s pebble bed blanket design: lithium titanate, beryllium and reduced activation ferritic steel F82H. The mock-ups were irradiated with 14 MeV neutrons from the DT neutron source at FNS. The lithium titanate was enriched to 40 and 95% in Li 6 . The local tritium production rate in the 12 mm thick breeding layers was measured by means of pellet detectors of lithium titanate, Li 2CO 3 and liquid scintillation counting techniques. Additional verification of the neutron spectrum was obtained from material activation such as activation foils. The results were analyzed with the Monte Carlo neutron transport code MCNP-4C and several transport cross section libraries such as FENDL/MC-2.0. The calculation/experiment ratio was found to be within the combined error estimate of calculation and experiment but in some cases, the calculations showed an overestimation tendency.

Tritium Self-Sufficiency and Neutron Shielding Performance of Self-Cooled Liquid Blanket System for Helical Reactor

Compatibility between tritium self-sufficiency and neutron shielding performance of self-cooled liquid blanket systems without solid neutron multiplier was investigated for the purpose of application to the conceptual helical reactor design of modified FFHR2 having the blanket space of 120 cm. The results of the neutronics calculation indicated that all of the Li/V-alloy, Flibe/V-alloy and Flibe/JLF-1 (Reduced Activation Ferritic steel) blankets are feasible for tritium breeding ability and neutron shielding performance. With the use of vanadium alloys, operation efficiency will be enhanced.

Activation Analysis for Two Molten Salt Dual-Coolant Blanket Concepts for the U.S. Demo Reactor

The US has considered, among other options, two blanket concepts for Demo reactor in which helium is primarily used to cool the first wall (FW) and structure whereas molten salt (MS) is used as both coolant and breeder. Conventional reduced activation ferritic steel (RAFS, F82H) is used as the structural material in both blanket concepts. The low melting point Flibe (~380°C) is used in the first option while the Flinabe (~305°C) is used in the second option. In this paper, we present the results for assessing the radioactivity and decay heat. This assessment is performed separately for the structural material, the Be multiplier and the breeder (Flibe/Flinabe). The Class C waste disposal rating (WDR) was estimated for each material. For Flibe, Flinabe and Be the WDR is much lower than unity. However, the WDR for F82H is ~0.6-1.3. They are attributed to reactions with Mo and Nb present in F82H with levels of 70 wppm and 4 wppm, respectively. To ensure that F82H qualifies for shallow land burial, it is suggested to reduce these two impurities to ~50 and ~3 wppm, respectively.

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li2BeF4 and the low melting point molten salts such as LiBeF3 and LiNaBeF4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiCf/SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

Effects of Nickel Addition on Microstructural Evolution and Mechanical Properties of Reduced Activation Martensitic Steels Irradiated in the ATR-A1

Validity of nickel isotope tailoring (NIT) method to be able to generate a large amount of transmutation helium in 9%Cr-2%W reduced activation ferritic (RAF) steel for fusion reactor structural material was investigated and discussed through mechanical properties and microstructural evolution after fission neutron irradiation. Miniature tensile and Charpy impact specimens of RAF steels and 1% nickel added RAF steel were irradiated in the ATR-A1 to evaluate irradiation hardening and shift in ductile-brittle transition temperature (� DBTT). The amount of transmutation helium in all the RAF steels with and without nickel addition was calculated as only about 0.6 appm. After irradiation at 621 K, theDBTT for the 1% nickel added steel was similar for the JLF-1. After irradiation at 543 K, however, theDBTT for the 1% nickel added steel was significantly larger than that for the JLF-1. Microstructure observations revealed that irradiation-induced dislocation loops in the 1% nickel added steel were finer and denser than in the RAF steel without nickel addition, suggesting that the nickel addition to the RAF steels directly affected nucleation and growth processes of dislocation loops and enhanced irradiation hardening and embrittlement. Therefore, there is a limit in the NIT method as a simulation method for understanding effects of helium on irradiation embrittlement of RAF steels. The RAF steels used in the present study were JLF-1, Mod.JLF-1/LSM and Mod.JLF-1/LSM/Ni. (The ''LSM'' means low concentration of Si and Mn compared with the JLF-1.) The chemical compositions and the heat treatments of these steels are shown in Table 1. Mod.JLF-1/LSM/Ni contained 1 weight percentage of natural nickel. Neutron irradiation was carried out in the two sub-capsules of the ATR-A1, AS4 and AS16, in the Advanced Test Reactor (ATR) at Idaho National Engineering and Environ- mental Laboratory for 132.9 Effective Full Power Days (EFPDs). 11) Irradiation temperature and displacement dam- age were estimated for the two sub-capsules as at 543 K to 2.2 dpa and 621 K to 3.4 dpa. These sub-capsules were equipped with gadolinium shields of 1.7 mm in thickness to reduce the thermal neutron flux in order to suppress trans- mutation effects. The amount of transmutation helium was also suppressed due to the thermal shields. Greenwood et al. calculated the concentration of transmutation helium as 0.6 appm in Mod.JLF-1/LSM/Ni and 0.5 appm in the other RAF steels without nickel addition, respectively. 12)

Thermal helium desorption behavior in advanced ferritic steels

Thermal helium desorption measurements were performed to investigate the difference in the helium trapping and accumulation behavior among a reduced activation ferritic (RAF) steel and oxide dispersion strengthening (ODS) steels after implantation of He + ions at room temperature. Thermal helium desorption spectra (THDS) were obtained during annealing to 1200 °C at a heating rate of 1 °C/s. The THDS of the ODS steels are very similar to that of the RAF steel, except for the presence of the peak in the temperature range from 800 to 1000 °C, where the α–γ transformation related helium desorption from the γ-phase is considered to occur in the 9Cr-ODS martensitic steels. The fraction of helium desorption becomes larger at higher temperatures, and this trend is increased with the amount of implanted helium. In the 9Cr-ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel. This suggests that the bubble formation was suppressed in the ODS steels. In the 12Cr-ODS steel, the fraction of helium desorption by bubble migration reached more than 90%, suggesting that the trapping capacity of martensite phase in the 9Cr-ODS steel is rather large.

Specimen size effects on fracture toughness of JLF-1 reduced-activation ferritic steel

Four different sizes of compact tension (CT) specimens of a reduced activation ferritic steel, JLF-1 steel, were fabricated and the fracture toughness was measured at room temperature by means of the unloading compliance method referring to in the ASTM E1820-99a. The J Q values obtained for the 1T–1CT and 1/2T–1CT specimens were 404 and 623 kJ/m 2, respectively, indicating that the J Q value increased with decreasing the specimen thickness. On the other hand, the J Q value decreased when specimens were miniaturized while maintaining the similar proportions. The J Q values for the 1/2CT and 1/4CT specimens were 371 and 321 kJ/m 2, respectively. Specimen size effects are interpreted in terms of an increase in the plain stress state region and plastic zone size at the crack tip in the specimen.

Evaluation Strength and Fracture Toughness of Reduced Activation Ferritic Steel (JLF-1) for Fusion Reactor

Reduced activation ferritic (JLF-1) steel is one of the leading candidates for blanket/first-wall structures in D-T fusion reactors. In fusion applications, structural materials will suffer the effects of repeated temperature changes. Therefore, a database of tensile strength and fracture toughness are very important as the temperature is 400 ° C and it's TIG welded joint. In this paper, fracture toughness JIC and KIC) and tensile tests were carried out at room temperature (RT) and at elevated temperature (400 ° C). The tensile properties of the TIG welding joint of JLF-1 steel was also investigated. It was shown that the tensile strength and fracture toughness values of the JLF-1 steel are slightly decreased with as increase in temperature. The fracture toughness values of JLF-1 steel at room temperature and at 400°C show excellent JIC values of about 530 kJ/m2 and 340 kJ/m2, respectively.