Development of benchmark reduced activation ferritic/martensitic steels for fusion energy applications

RAFM steels contain elements selected so as to minimize fusion neutron production of highly radioactive and volatile nuclides during short time periods (relevant for public safety during design basis accident scenarios) and longer-lived radioactive isotopes that decay to low levels (e.g. suitable for recycling or shallow land burial) in the timeframes required by the waste management scenario (typically within 100 years following reactor shutdown). Neutronics and safety analyses have highlighted the importance of the removal of undesirable elements such as Co, Cu, Ni, Mo and Nb that are commonly found in conventional steels as impurities. The concentrations of elements essential for the mechanical properties of some RAFM steels, such as Al and N, required a compromise between the waste disposal scenario and performance demand.

The specific limits and target activation level after service are dependent on DEMO operation and waste management scenario. In the case of F82H, Japanese RAFM steel, achieving the shallow land burial limits 100 years after a reactor shutdown is the current target for reduced activation level. It was found that Ni is the most significant solute that must be removed, and Mo as well in the case of the highest dose region such as first wall, based on numerical analyses of the induced radioactivity. Limiting N to levels below the maximum level relevant for reduced activation criteria will be a potential issue, especially for a large scale melt, but concentrations of Al up to the maximum allowable amount (300 wt. ppm) has been demonstrated on large scale heats [5]. It is apparent from figure 1 of [6] that the European RAFM steels EUROFER-97 already meet the requirement of being Low Level Waste (surface gamma-dose rate below 2  ×  10⁻³ Sv h⁻¹) within 80–100 years after shutdown of DEMO reactor following 5 full power years of operation (12.5 MW year m⁻²). Up to one order of magnitude lower calculated surface gamma-dose of F82H in comparison to EUORFER97 1000 years after reactor shutdown is mainly due to the lower Nb (1 ppm) content of F82H compared to 10 ppm for EUROFER-97. Reduction of the surface dose rate down to Hands-on-Level (being nearly two orders of magnitude lower than the Low Level Waste rate) will require further reduction of undesirable impurity content which seems to be technically feasible at the expense of production cost [6].

Physical properties, such as elastic properties, thermal properties and thermal expansion are the key properties to conduct structural design activities. Magnetic properties and electrical resistivity are also essential for designing magnetic confinement fusion reactor in order to evaluate the impact of ferromagnetic property to plasma control, and to evaluate the impact of plasma disruption on in-vessel components. These physical properties of F82H and EUROFER-97 are summarized in [7, 8].

It was indicated that the deviations of elastic properties of F82H from that of Grade91 [9] was very small (figure 1 of [7]), and thermal properties such as thermal conductivity, diffusivity, or specific heat of F82H and EUROFER-97 do not differ from those of Grade91 in general. The thermal conductivity of these RAFM steels shows a higher value than that of Grade91 for lower temperature range, and the thermal conductivity data from an early heat of F82H [10] show somewhat higher values (figure 4 of [7]). The coefficient of thermal expansion (CTE) of F82H at each temperature below 600 °C shows very similar value to that of Grade91. Magnetic properties obtained from the hysteresis loops suggested that there is very small difference in its saturation magnetization of F82H and EUROFER-97.

Mechanical properties of RAFM steels are based on its specific microstructure, i.e. a high dislocation density, fine martensite lath structure, and high number density of precipitates (mainly M₂₃C₆ and MX) on various boundaries and in the steel matrix. The precipitates are expected to be fine and stable to provide good toughness and high temperature resistance, and these fine structures are regarded as the main factor of RAFM's high resistance to various irradiation effects. Since there have been no target value of RAFM properties given from design requirements, RAFM steels have been developed to have the basic mechanical properties, such as tensile, creep and Charpy impact properties, expected to be equivalent to or better than those of Grade91. The international round robin property tests were performed under IEA implementing agreement for a programme of research and development on radiation damage in fusion material in the late 1980s to 1990s on F82H as the reference IEA RAFM steel and the results have been overviewed in [11, 12].

Figure 1 shows the unirradiated tensile properties at room temperature and Charpy impact properties of various heat and thickness of F82H plates. Here the Cr equivalent concentration is described as follows [13].

$$\begin{eqnarray}&&\text{Cr}\,\text{eq}~.=\left[\%\text{Cr}\right]+{{\text{2}}^{\ast}}\left[\%\text{Si}\right]+\text{1}.{{\text{5}}^{\ast}}\left[\%\text{Mo}\right]+{{\text{5}}^{\ast}}\left[\%\text{V}\right]+\text{1}.\text{7}{{\text{5}}^{\ast}}\left[\%\text{O}\right]\\ &&+0.\text{7}{{\text{5}}^{\ast}}\left[\%\text{W}\right]+\text{1}.{{\text{5}}^{\ast}}\left[\%\text{Ti}\right]+\text{5}.{{\text{5}}^{\ast}}\left[\%\text{Al}\right]+\text{1}.{{\text{2}}^{\ast}}\left[\%\text{Ta}\right].\end{eqnarray}$$

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It was clearly shown that those values of 0.2% proof stress and ultimate tensile strength are distributed within the American Society of Mechanical engineers (ASME) defined values for Grade91. There is no requirement defined for Charpy impact properties of Grade91 in ASME, but those data of F82H show quite acceptable for French order concerning nuclear pressure equipment (ESPN) requirement defined for martensitic steels.

The investigation of the mechanical properties of the EUROFER-97 has been the subject of numerous publications [14–22]. The tensile properties of EUROFER-97 were thoroughly characterized by different European laboratories [16–18]. The tensile properties of EUROFER-97 are comparable to those of F82H steels. The influence of the laboratory scale heat treatment on the tensile strength for different product forms has been studied in [16]. The room temperature tensile strength results for rod and plate materials are shown in figure 2. It is found that the application of the laboratory scale heat treatment at a higher austenization temperature leads to the lower room temperature ultimate tensile strength and lower 0.2% proof strength values. The difference in the tensile strength properties, however, diminishes with increasing the test temperature [16]. It has to be noted that in the case of 14 mm plate the reduction of the tensile strength in comparison to the as delivered condition was also observed in lower austenization temperature of 950 °C. This observation indicates that laboratory scale heat treatment with better controllable and more homogeneous temperature profile due to much lower furnace volume leads to a further optimization of the mechanical properties. The Charpy impact properties of different heats of EUROFER-97 have been investigated by using the both standardized ISO-V (International Organization for Standardization V-notch) specimens and miniaturized KLST (Kleinstprobe in German Standard DIN 50 115) specimens [16, 17, 23]. The ductile to brittle transition temperature (DBTT) obtained from ISO-V specimens varied between  −72 and  −53 °C and ISO-V upper shelf energy (USE) varied between 237 and 300 J. The KLST DBTT values were somewhat higher and varied between  −70 and  −90 °C. The KLST USE varied between 9.1 and 9.8 J. The application of the laboratory scale heat treatment at a higher austenization temperature of 1040 °C to 25 mm EUROFER-97 plate resulted in the reduction of the KLST DBTT by  −9 °C from its as-delivered value of  −82 °C for an austenization temperature of 980 °C [19].

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It should be noted that these tensile and Charpy impact properties are controllable with the tempering condition. It is clearly shown that the yield stress decreases as the amount of tempering is increased, but for the Charpy impact properties the effect is similar but not as straight forward in case of F82H (figure 3) [24].

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Fracture toughness is an important parameter for mechanical designs based on a defect-tolerant approach. The J-integral is the key required engineering value in order to evaluate tolerance in thin-walled structures which is the expected usual operating condition for in-vessel components of fusion reactors. Conversely, most fracture toughness data obtained to data have measured the brittle fracture toughness, K value, as ductile brittle transition is very important design issue for ferritic/martensitic steels. The master-curve (MC) method (ASTM standard E1921) is a widely used method to estimate the fracture toughness reference temperature T₀, which defines the DBTT, with appropriate statistical accuracy with a minimum number of specimens. This method has been used to obtain the toughness of RAFM steels since it is beneficial for defining the DBTT shift of irradiated RAFM steels with a limited number of the specimens (highly valuable due to limited volumes in currently available fission neutron irradiation facilities). Thus, the fracture toughness has been investigated and T₀ has been defined by several laboratories for F82H [25–29] and EUROFER-97 [23, 30, 31]. It was shown in [26] that the application of the MC methods for single temperature datasets might strongly underestimate the T₀. A modified MC with a significantly lower athermal part (12 MPa m^1/2) than the recommended value (30 MPa m^1/2) has been proposed which yielded a test temperature independent reference temperature of  −78 °C for the 25 mm EUROFER plate. For comparison, the fracture toughness reference temperature determined by application of the standard MC was found to be  −121 °C for the rod material in [23].

Thermal aging effects are another key factor which helps to define the design limit of RAFM steel as they are a signature of the phase stability limit. As for the F82H-IEA heat, aging tests performed up to 100 000 h demonstrate only slight softening up to 550 °C, but produce a substantial DBTT shift to room temperature for aging at 550 °C and greater shifts for aging at 600 °C or higher [32]. This is one of the reasons that the upper limit operating temperature of RAFM has been set at 550 °C. Aging test results for EUROFER have been reported up to 12 000 h [17] and shows better aging resistance than F82H. For example, the DBTT of EUROFER does increase due to high temperature aging, but the DBTT shift was limited to 23 °C after aging at 600 °C for 10 000 h [17].

Thermal creep properties are the most important properties to define heat resistant steel. The creep rupture property of F82H [24] and EUROFER-97 [33] turned out to be comparable to those of Grade91 plates (figure 4 of [24]), and this gives a certain level of assurance to the quality of these RAFM steels as the structural material of pressure component. On the other hand, it is important to define temperature, stress and time limits for negligible creep and minimum creep rate for fusion application. The current creep database is not enough to determine the regime of negligible creep, and only limited data are available on the minimum creep rate up to now (figure 4).

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The effects of the pulsed operation expected in ITER and early DEMO operation will cause damage due to fatigue and creep-fatigue interaction. Therefore, fatigue and creep-fatigue properties are another important issue for fusion applications. Various fatigue studies have been reported by EU researchers on EUROFER-97 [14, 34] and Japanese researchers [35, 36] on F82H (figure 5). The creep-fatigue interaction for EUROFER steel has been investigated at 450 and 550 °C in [34, 37]. In [34] the influence of the hold time duration in tension, compression and tension and compression has been investigated at 550 °C for a total strain range of 1.0%. Hold time in tension had no influence on the lifetime up to investigated hold time lengths of 10 min. Hold time in compression, however, led to a progressive reduction of the lifetime with increasing hold time duration. It should be noted that environmental effects, such as oxidation, might have additional significant effects on the creep-fatigue property depending on the purge and/or coolant gas condition of the fusion blanket system. The application of the multi-step low cycle fatigue (LCF) loading by varying the total strain range amplitude on the EUROFER-97 resulted in non-linear damage accumulation in [37]. In comparison to single-step tests some multi-step experiments yielded significantly shorter lifetimes.

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In the fusion blanket system, the volume of the structural material is expected to be minimized to ensure achievement of sufficient tritium breeding ratio to enable self-sufficiency in DEMO. Therefore, it is very important to evaluate the impact of environmental effects, such as the corrosion rate under various coolant conditions, in order to evaluate minimum allowable wall thicknesses for structural components.

Aqueous corrosion is a key issue in a water cooled blanket. The 9Cr ferritic/martensitic steels generally exhibit better corrosion resistance in the steam generators compared to that of low-alloy steels, and as good as that of austenitic steels. Corrosion behavior of F82H has been studied under BWR condition water (220–290 °C, 10 MPa) for 500 h in a recirculating high temperature water environment containing 1 ppm dissolved oxygen [10], and it was revealed that weight gain of F82H was observed at temperatures over 260 °C but the weight loss was observed below 250 °C.

The impact of flow on corrosion rate is also a key factor to estimate corrosion allowance, as the flowing rate of coolant water is expected to be 2 to 5 m s⁻¹ in fusion blanket systems. The flow assisted corrosion rate has been preliminary assessed using a rotating disk apparatus in high temperature pressurized water of 15 MPa at 300 °C with  <20 ppb dissolved oxygen, and the significant weight-loss in rotating condition was reported even though the weight-gain was reported in static condition in this corrosion test condition [38], but the effect becomes insignificant with 8 ppm dissolved oxygen [39].

Corrosion resistance in super critical pressurized water (SCPW) condition was also evaluated to examine the feasibility of SCPW cooled blanket system for higher thermal efficiency. The tests conducted at 290–583 °C under 23.5 MPa pressures suggested that the weight gain due to high temperature oxidation was significant, especially over 550 °C where the corrosion rate was estimated to be 0.04 mm per year [40].

The corrosion behavior of RAFM steels in direct contact with Li–Pb is critical for the advanced blanket system using Li–Pb as the breeding material. Studies were carried out on the compatibility of RAFM steels with flowing Pb–Li as summarized in table 1 [41–47]. The corrosion rate of RAFM steels is less than 100 μm yr⁻¹ in Pb–Li at 480 °C, regardless of the testing periods and the corrosion rate significantly increases with increasing test temperature: 700 μm yr⁻¹ at 550 °C [43]. The high flow rates resulted in a high corrosion rate, indicating that the flow rate needs to be suppressed.

Fission reactor irradiation data on F82H and EUROFER has been accumulated under various programs, including domestic, EU/Russia and Japan/US collaboration programs. Now 30–70 dpa irradiation data at ~330 °C are available from a BOR-60 irradiation program [20, 58], and 20 dpa data at 300 °C from a HFIR irradiation program [59]. As indicated in figure 6 of [24] there appears to be an irradiation hardening saturation around 30 dpa for irradiation at temperatures below 350 °C, and irradiation softening was observed for higher irradiation temperatures (above 400 °C) in the FFTF/MOTA irradiation program [4]. It should be noted that the softening is much greater than that observed in thermal aging where no significant softening was observed up to 500 °C/100000 h. Comparatively, minor irradiation induced softening was observed for EUROFER-97 after irradiation to 16.3 dpa at 450 °C in HFR reactor [22]. Another significant irradiation effect on tensile properties is the loss of elongation. Figure 6 shows the dose dependence of 0.2% proof stress and total elongation of F82H at various irradiation temperature ranges. It was clearly shown that the total elongation decreases regardless whether irradiation hardening or softening was observed. It should be noted that there is an impact of size and/or shape of small sized flat tensile specimen shape on quantitative evaluation of total elongation [59]. Elongation properties remained nearly unchanged in EUROFER-97 after irradiation to 16.3 dpa at 450 °C in HFR in comparison with the unirradiated state [22].

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The increase in hardness of RAFM steels irradiated at low temperatures produces a corresponding increase in DBTT. Numerous studies have found a linear correlation between the DBTT shift and the increase in tensile strength. The DBTT shift of F82H and EUROFER-97 obtained in post irradiation Charpy impact tests are shown in figure 7. Whereas a rapid increase in DBTT occurs at low doses, only a slight DBTT shift increase is observed as dose is further increased from 30 to 70 dpa at ~330 °C [20, 58]. It should be noted that embrittlement without hardening was also observed for 400 °C irradiated F82H [60]. This tendency quite agrees with the observed irradiation effects on plasticity, as is shown in figure 6, where the loss of total elongation was observed not only at 250–350 °C but also at higher temperatures, even though no or slight hardening was observed at those temperatures, in some cases. These results observed for irradiation above 400 °C could be interpreted as irradiation accelerated aging effects [60], but it should be noted that the embrittlement level is much larger than expected as an aging effect, as the DBTT shifts in the 400 °C irradiated specimens are much larger than that of 650 °C aged F82H [32].

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The initial heat treatment can affect the irradiation hardening and embrittlement response, analogous to the heat treatment effects on the mechanical properties of as-fabricated RAFM previously shown in figure 3. It was reported that an optimized heat treatment could reduce irradiation hardening and embrittlement in F82H [20, 61–63], and this indicates that PWHT conditions could affect irradiation response. It was also reported that hardening and DBTT shift observed in EUROFER and F82H irradiated to 15–70 dpa at 330 °C can be recovered by post-irradiation heat treatment [58, 64, 65].

Irradiation creep has been recognized as one of the critical issues to be considered for design activity as it will be the key factor for relaxation of thermal stress expected at the first wall. The irradiation creep data of RAFM steels has been accumulated by the irradiation of pressurized creep tubes in FFTF [66], HFIR [67] and BOR60 [68]. In case of F82H, irradiation accelerated creep is observed at higher temperatures (>400 °C), while irradiation induced creep is observed at 300 °C where the thermal creep is negligible (figure 8). It should be noted that high dose irradiation data could include some portion of stress relaxation depending on the magnitude of creep deformation as it was obtained from creep tube specimens. Irradiation creep of EUROFER-97 was investigated in [68]. Pressurized tubes have been also utilized to determine deformation under applied stresses of 150 and 220 MPa at damage doses of 19, 42 and 63 dpa at the irradiation temperature of 330 °C. The irradiation-creep-modulus A has been determined by using the diametral strain and hoop stress. EUROFER-97 exhibited the irradiation creep modulus of A  =  (0.7  ±  0.1)  ×  10⁻⁶ (dpa MPa)⁻¹ which is very close to the creep-modulus for similar class of materials, e.g. 9Cr1Mo, 9Cr2WTaV steels.

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Irradiation effects on fatigue properties are recognized as one of the potentially most important issues to assess the lifetime of the system. The difficulty in the accurate assessment of fatigue life of the irradiated fusion blanket system is that nether in-beam fatigue tests nor post irradiation fatigue tests represent the real fatigue mode expected for the fusion blanket system. Firstly the anticipated fatigue mode is irradiation creep fatigue, and secondly thermal loading and relaxing occurs under neutron flux whereas unloading occurs without neutron iradiation. Currently, there are insufficient experimental data at those fusion-relevant creep-fatigue conditions to conclusively determine whether there will be a significant irradiation effect on the creep-fatigue behavior of RAFM steels. Nevertheless, numerous fatigue property evaluations have been conducted. Post irradiation LCF behaviour of EUROFER-97 and F82H-mod steels has been investigated [22, 65, 69, 70]. In [70] the miniaturized LCF specimens have been tested after neutron irradiation in BOR-60 reactor at 330 °C to damage doses up to 70 dpa for EUROFER-97 and up to 47 dpa for F82H-mod. The strain controlled push–pull (LCF) loading was performed at a constant temperature of 330 °C for different total strain ranges (Δε_tot) between 0.8% and 1.2% and at a strain rate of 3  ×  10⁻³ s⁻¹. For low total strain ranges, the irradiated EUROFER-97 exhibited increase of the lifetime in comparison with the unirradiated state. This observation was mainly related to the increase in the elastic part of the cyclic deformation and the related reduction of the inelastic strain amplitude due to low temperature irradiation-induced hardening. At high total strain ranges, however, some of the irradiated EUROFER-97 samples showed reduced lifetime compared to reference unirradiated state, indicating that hardening induced increase in the stress level might lead to enhanced damage evolution. 47 dpa 330 °C irradiated F82H-mod showed no lifetime reduction in comparison with the unirradiated state. Comparatively, EUROFER-97 after irradiation to 16.3 dpa at 250 °C in HFR reactor also did not show reduced lifetime [22]. In contrast, a clear tendency for an increased lifetime compared to the unirradiated state was observed with deceasing total strain range. At higher irradiation temperatures of 350 and 450 °C, however, the lifetime remained in the range of unirradiated material [22]. In-beam (proton beam) fatigue tests have reported a longer fatigue life as the general trend of irradiation effects [71, 72]. However shorter fatigue life is observed in some cases. An additional effect associated with implanted hydrogen may be anticipated in the case of proton beam in-beam fatigue tests [73]. The in-beam fatigue lifetime was between that of unirradiated and post irradiation fatigue test by the direct comparison of in-beam fatigue to post irradiation fatigue test with standard size specimen at 420 °C [74].

Investigation of the irradiation response of joints of RAFM steels is essential for ensuring structural integrity of the blanket, since the blanket has a geometrically complex structure with various joints and some portion of the joints are located in regions where high irradiation dose is expected. The joints of F82H, which were prepared using TIG and EB welding and diffusion bonding (HIP), were irradiated up to 4.5 dpa at 300 °C [49]. The post-irradiation mechanical tests revealed that, in TIG welding, the hardening in WM and BM are greater than 300 MPa. On the other hand, the HAZ radiation hardening was about half that of the WM and BM. The irradiation hardening observed in the WM region of EB welds irradiated at 400 °C was larger than that in TIG WM regions. Therefore neutron irradiation significantly enhances the weakness of the HAZ. On the contrary, the mechanical properties of HIP joints were very similar to that of BM even after the irradiation.

Evaluation of transmutation-formed He effects is arguably the most important issue in evaluating 14 MeV fusion neutron irradiation effects for RAFM steels. Since an intense fusion neutron source is not available, effects of He on microstructure evolution, especially on void swelling, have been investigated by neutron irradiation on boron doped RAFM steel or by dual/triple ion irradiation (figure 9) [75–84]. The ion irradiation experiments indicate that void swelling in RAFM is maximized near 470 °C [82] and swelling can be particularly enhanced under triple ion irradiation conditions, suggesting synergistic effects of H and He on dramatically enhancing void swelling [83]. However, the swelling observed in these ion irradiation experiments is affected not only by dose rate (and its depth profile) but also by various undesirable effects due to the specimen surface, boundaries and injected ions [82, 84]. Therefore, the quantitative level of swelling measured in the ion irradiation experiments may be an underestimate of what would occur in a fusion reactor.

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A variety of techniques have been used to investigate helium co-production during neutron irradiation, including use of B- or Ni-doped samples. In general, these studies have observed enhanced cavity swelling in samples with fusion-relevant He levels compared to the low swelling observed in standard (low-He) fission neutron irradiation. A new He injection technique during neutron irradiation was reported utilizing Ni(n,α) reactions by plating NiAl on specimen surfaces, and this could solve some of the issues associated with ion irradiation, such as dose rate effect [85]. The key issues with this technique are associated with the limitation of the He injected zone which is just a few μm from the specimen surface, irradiation temperature ambiguity due to irradiation capsule design, and a He-injected neutron irradiation volume that is not sufficient for the quantitative determination of bulk mechanical properties.

Irradiation of B or Ni doped RAFM steels has been conducted to evaluate He (and H) effects on mechanical properties [86–89], and those results generally indicate the presence of helium effects on hardening and embrittlement. However, the application of doping technique has some difficulty in interpretation of the results. In the case of B doped RAFM steel, the transmutation of B, i.e. ¹⁰B(n,α)⁷Li, will complete in the very early stage of irradiation and He/dpa will become excessively high in low dose, if thermal neutrons are not shielded, and the product Li could have an impact on mechanical property changes as Li has very low solubility in Fe, even though some portion of the product Li was observed in (or adjacent to) He bubbles [90]. In the case of Ni doped RAFM steel, the experimental validity was discussed [91, 92] and the formation of a Ni intermetallic phase [93], enhancement of neutron induced amorphization [94] and the impact of initial heat treatment conditions [95] were reported. These indicate the potential difficulty in interpreting results on neutron irradiated Ni doped RAFM as only due to a helium effect.

High energy proton irradiation utilizing a spallation neutron source is the only currently available method that can produce high concentrations of He (and H) during medium-dose (>1–10 dpa) irradiation of bulk specimens. Lots of work has been done at the SINQ Target irradiation at PSI, Switzerland, showing that He will cause extra hardening and DBTT shift even at low dpa [96, 97]. But it should be noted that the helium production ratio in those experiments is 10 times larger than in a fusion environment, the formation of hydrogen is also large, and production and segregation of transmutation elements such as Ca, Ti and Sc were observed [98]. All of these processes could have an impact on the observed mechanical properties. Furthermore, as the specimens are heated mainly by protons, the irradiation temperature of a specimen is more or less proportional to the instantaneous proton flux leading to a large variation in the specimen temperature during the prolonged accelerator irradiation experiment.

Helium implantation using a cyclotron is another alternative method to investigate He effects, even though the implantation depth is limited to about 1 mm from the incident surface. The results of this type of experiment suggest a variety of different responses. In particular, high temperature helium implantation up to 1000 appm was found to cause intergranular embrittlement during Charpy impact tests at low temperature [99], no high temperature He embrittlement effect was observed in post implantation creep tests [100], and significant reduction of the fatigue life and an increase in the micro-crack propagation rate was reported on 350 appm He implanted F82H [29].