Research and Development Status of Reduced Activation Ferritic/Martensitic Steels Corresponding to DEMO Design Requirement

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1 Research and Development Status of Reduced Activation Ferritic/Martensitic Steels Corresponding to DEMO Design Requirement Hiroyasu Tanigawa 1, Hisashi Tanigawa 1, M. Ando 1, S. Nogami 2, T. Hirose 1, D. Hamaguchi 1, T. Nakata 1, H. Sakasegawa 1, M. Enoeda 1, Y. Someya 1, H. Utoh 1, K. Tobita 1, K. Ochiai 1, C. Konno 1, R. Kasada 3, A.Möslang 4, E. Diegele 5, M.A. Sokolov 6, L.L. Snead 6, Y. Katoh 6, R.E. Stoller 6, S.J. Zinkle 6 1. Japan Atomic Energy Agency, Japan 2. Tohoku University, Sendai, Miyagi Japan 3. Institute of Advanced Energy, Kyoto University, Uji, Kyoto, Japan 4. Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany 5. Fusion for Energy, Barcelona, Spain 6. Oak Ridge National Laboratory, Oak Ridge, TN U.S.A. tanigawa.hiroyasu@jaea.go.jp Abstract. A strategy and issues for reduced activation ferritic/martensitic steels R&D were studied with respect to DEMO design requirements, since fusion structural material R&D must proceed to a new stage involving ITER test blanket module (TBM) preparation and DEMO construction. The design limit and technological challenges were identified in consideration of the impacts of irradiation induced loss of plastic deformability and predicted 14 MeV neutron irradiation effects. 1. Introduction Crucial issues on the path to fusion power are the development of plasma facing and breeding blanket materials which are capable of withstanding high neutron fluences of 25 to 50 dpa/year and high heat fluxes of 0.4 to 0.5 MW/m 2. Reduced activation ferritic/martensitic (RAFM) steels, such as F82H in Japan and EUROFER in the EU, are now considered to be the candidates for structural applications in the fusion demonstration reactor, DEMO [1], because they have a sound engineering basis, such as fabrication technologies and material property database, including irradiation effects. But it is also well recognized that the severe DEMO operating conditions, especially 14MeV fusion neutron irradiation, could cause extra degradation of mechanical properties, such as the loss of plastic deformability which is not covered by current design codes. Thus, estimation methods of materials behavior under 14MeV fusion neutron irradiation and a design methodology for highly irradiated structure have become indispensable elements in DEMO developments. The objective of this paper is to identify the design requirements, the design limits and technological challenges based on the current status of the R&D of RAFM steels. 2. Loading condition in DEMO blanket The blanket system is stressed under different operation modes by various loading conditions. A preliminary analysis of failure mode and event-related material properties corresponding to an assigned DEMO operation status was conducted. DEMO operation mode was assigned based on ITER operation mode (Table.1). In here, the DEMO blanket is expected to be able to sustain and use continuously in normal operation and likely events. Thermal stress and electromagnetic load are calculated based on currently proposed Japanese

2 DEMO design, SlimCS [2], assuming that F82H is used as its structural material. In normal operation, 3 MW/m 2 average neutron wall load and 0.5 MW/m 2 surface heat flux on first wall is assumed to load under steady state or pulse operation. Thermal stress loaded on the blanket in normal operation mode was calculated in 2D plane of the middle cross section of an outer board blanket at equatorial plane. 15MPa pressurized water is assumed as a coolant water which temperature is 290 C C, and breeding zone with 16% Li 2 TiO 3, 64% Be 12 Ti and 20% He was considered to calculate nuclear heat only (Fig. 1). The calculated thermal stress distribution suggests that the uniaxial stress up to 400MPa in toroidal direction (z direction) is dominant around the cooling channel, where the temperature is about 300 C (Fig.2). As a likely event, plasma current disruption in a normal worst case fast disruption is assumed where plasma current (16.7MA) was assumed to linearly decrease in 30ms, and the Eddy current and electromagnetic forces loaded to an inner board blanket at equatorial plane were calculated. Four 100x100mm key structures were attached to fixate the blanket. No cooling channel but a solid plate with different thickness was considered in this model (Fig.3). It was clearly indicated that the uniaxial stress up to 600MPa in toroidal direction (z direction) is also dominant at the corner of the middle plane cross section. Stress concentration over 600MPa was observed around the key structure and edge of the blanket box structure (Fig.4). 3. Key issues of RAFM R&D for DEMO application In case of normal operation, thermal stress level is about 400MPa as shown in Fig.2, and this is well below the yield stress level at 300 C (fig.5). On the other hand, irradiation creep data (fig. 6) indicate that 0.05%/dpa of irradiation creep is anticipated at the stress level of 400 MPa, and the induced thermal strain (0.20%) will be relaxed in 4 dpa irradiation. This suggest that thermal stress will be not an issue in case of steady state operation, but the creep-fatigue in combination with irradiation effects will be the key properties for pulse operating conditions. In case of plasma disruption, the calculated electromagnetic force is almost reached to the yield stress level, and fast fracture toughness resistance will be the key in this case, since the load will be applied in 30ms. There are no reported post irradiation fast fracture toughness data but the ductile brittle transition temperature shift, and the future irradiation works should consider this requirement. The impact of electromagnetic force also indicates that the fatigue accompanied with plastic deformation should be anticipated. And the structural discontinuity by cooling channel and first wall side wall weld should be considered, as it will induced peak stress above the calculated stress level. It should be also noted that the poloidal direction (y) stress was also observed at where the maximum toroidal direction (z) stress was observed. This indicates that the presence of biaxial stress loading. The biaxial fatigue will lower the fatigue life in an order, thus the impact of this including irradiation effects should be investigated. A critical issue is the materials loss of ductility under irradiation. It was found that a significant hardening and loss of uniform elongation were observed after irradiation to a few dpa at 250~350 C, and both tensile strength and uniform elongation decrease at higher temperature irradiation (Fig. 5). This loss of plasticity could finally result in unstable crack propagation, when fracture or fatigue crack propagation is the failure mode 14MeV fusion neutron irradiation effects, especially helium effects, are anticipated to enhance this irradiation induced loss of plasticity. Reports of helium effects such as extra hardening [5], increment of fatigue crack propagation rate [6], or increased ductile-brittle

3 transient temperature shift [7] indicate that a certain amount of helium could increase the possibility of unstable fracture. These results are based on simulation irradiation experiments, thus a mechanistic understanding of these property changes based on microstructural analyses is essential. 4. Needs of 14MeV neutron source As of today, there are indications from many studies and helium effects can be modeled or estimated, however, it is mandatory to verify fusion neutron irradiation effects at an early date with 14MeV neutron irradiation. Above all else, the estimation and early experimental verification of the critical condition is found to be essential, up to the condition where no significant 14 MeV fusion neutron irradiation effects are expected and fission irradiation data can be used for design activity, regarding to the fact that an intense 14MeV neutron source IFMIF - will not be available before 2020, and only the fission irradiation database and complementary modeling is available for design activity until that time. Reference [1] H. Tanigawa, K. Shiba, et. al., J. Nucl. Mater. 417 (2011) 9 [2] K. Tobita, S. Nishio, et. Al., Nucl. Fusion 49 (2009) [3] M. Ando et al.,, J. Nucl. Mater (2007) 122 [4] A. Kohyama et al., J. Nucl. Mater (1994) 751 [5] M. Ando, E. Wakai, et. al., J. Nucl. Mater (2004) 1137 [6] S. Nogami, M. Fukuda, et. al., presented in ICFRM15 (Oct. 2011) [7] A. Hasegawa, M. Ejiri, et al., J. Nucl. Mater (2009) 241 Load category Assigned DEMO operation status to a blanket system I Normal operation Steady state operation Pulse operation (Planned ON/OFF operation ) II Likely events Plasma current disruption (Thermal/current quench) A normal worst case fast disruption Earthquake III Unlikely events Plasma current disruption an upset/emergency worst case fast disruption extremely fast plasma current quench Plasma vertical displacement events (VDEs) Runaway Electron (RE) events IV Extremely unlikely events. Coolant leakage Loss of coolant accident (LOCA) Loss of forced flow accident (LOFA) Loss of off site electric power (LOOP) Table 1 Assigned DEMO operation status to a blanket system.

4 Fig.1 A calculated temperature distribution of middle plane cross section of an outer board blanket. Breeding zone is considered just as the nuclear heat source. Fig. 2 Left: Von Mises stress distribution, Right: thermal stress /strain distribution ploatted against the distance form plasma facing surface (PFS) at the dotted line indicated in left figure. Irradiation dose and the amount of transmutation formed He was also plotted (right).

5 Fig. 3 An assumed calculation model for the evaluation of electromagnetic force. Fig. 4 Von Mises stress (left) x, y, and z direction stress (right) distribution in case of a plasma current disruption. Fig. 6 Stress dependence of irradiation creep rate of F82H at each irradiation temperatures. [3,4] Fig. 5 Temperature dependence of yield stress and uniform elongation of irradiated/ unirradiated F82H tested at each irradiation temperatures.