Radiation damage on low activation materials used in a hybrid reactor

Materials & Design Materials and Design 28 (2007) 1453–1460 www.elsevier.com/locate/matdes

Radiation damage on low activation materials used in a hybrid reactor ¨ beyli Mustafa U

*

¨ niversitesi, Mu¨hendislik Faku¨ltesi Makina Mu¨hendislig˘i, So¨g˘u¨to¨zu¨-Ankara, Turkey TOBB Ekonomi ve Teknoloji U Received 21 July 2005; accepted 17 March 2006 Available online 11 May 2006

Abstract Structural material selection in design of fusion–fission (hybrid) reactors is very crucial to enhance the neutronic performance. Low activation materials have important potential to be used in fusion or hybrid reactors. This study presents radiation damage behavior of the primary low activation materials namely, ferritic steel (9Cr–2VWTa), vanadium alloy (V–4Cr–4Ti) and SiCf/SiC composite used as first wall and fuel clad structural material in a (DT) fusion driven hybrid reactor under a NWL of 2 MW/m2 for a plant operation of 1 year. Calculations were performed by using the code, SCALE4.3 solving the Boltzmann neutron transport equation. Among the investigated materials V–4Cr–4Ti showed the best performance with respect to radiation damage. Ó 2006 Elsevier Ltd. All rights reserved. Keywords: Ceramic matrix composite; Ferrous alloy; Nonferrous alloy; Radiation damage

1. Introduction Nuclear fusion is a safe and clean energy source which serves unlimited energy to mankind. One of the most important advantages of a fusion energy system is the abundant fusion fuel availability in the nature, contrary to relatively scarce fission fuel resources. Furthermore, a fusion energy system has aspects of an attractive product with respect to safety and environmental advantages compared to other energy sources. Hence, market penetration of fusion energy reactors may create a revolution in energy generation in near future. However, the market penetration of commercial pure fusion reactors have not been expected before the year 2050. On the other hand, the combination of a fusion and fission reactor may have realistic chances for a relatively earlier introduction of fusion power plants for electricity production. A fusion–fission (hybrid) reactor is a multi-functional type reactor that is a combination of the fusion and fission processes. The main idea in this type reactor is to convert the fertile materials (U238 or Th232), surrounding the fusion plasma, into fissile materials *

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(Pu239 or U233) by transmutation through the capture of the high yield fusion neutrons [1–20]. Furthermore, they may also undergo a substantial amount of fission under the irradiation of very energetic 14 MeV- (D,T) neutrons to increase fusion energy. Moreover, a hybrid reactor can burn and/or transmute the nuclear wastes effectively by using very energetic fusion neutrons [13,21–23]. In design of a hybrid reactor, one of the most important parameters is the selection of the suitable structural material to improve its neutronic performance. In hybrid reactors, first wall structural material of the fusion chamber adjacent to the plasma will be subjected to high heat loads, neutron and energetic particle fluxes together with alternating mechanical stress patterns. Therefore, radiation damage will limit lifetime of the first wall material. In order to decrease operational costs of either a fusion or hybrid reactor, replacement of the first wall structural material should be eliminated during the reactor lifetime. Different workers have suggested the use of a protective, flowing liquid zone between plasma and solid first wall to protect the solid first wall of a fusion reactor from direct exposure to the fusion reaction products [24–27]. This could extend the lifetime of the first wall to the expected lifetime of the fusion reactor, namely, to 30 FPYs

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[24–29]. Furthermore, it would allow a wider choice in selection studies for first wall materials in relaxing the material requirements and reduce the development costs for the first wall significantly. In addition, using flowing liquid wall allows for handling high neutron wall load (NWL) of P10 MW/m2 in a fusion reactor to get high neutronic and thermal performance. This would have a direct impact in the reduction of the cost of the electricity [24–26]. Despite liquid walls offer a significant advantage with respect to elimination of first wall material replacement every year, the high neutron flux will strongly be attenuated in the protective liquid wall before reaching the first wall so that it will not be available for incineration of nuclear waste and fissile fuel breeding in the blanket region of a hybrid reactor. This study presents radiation damage calculations on the low activation materials, namely, ferritic steel (9Cr– 2WVTa), vanadium alloy (V–4Cr–4Ti) and SiCf/SiC composite used as first wall and fuel clad structural material in a (DT) fusion driven hybrid reactor under a NWL of 2 MW/m2. Calculations were based on the plant operation of 1 year.

 Waste management: Activated waste is defined as low level waste (LLW), if the dose rate after a 50-years decay does not exceed 2 mSv/h. Furthermore, the decay heat must be below any ‘‘significant amount’’, e.g. less than 1 W/m3. If this limit is not fulfilled, the waste may release significant amounts of decay heat, i.e., more than 10 W/m3 after 50 years of cooling. This value, and/or a dose rate higher than 20 mSv/h, rates the waste as HLW. If the material has a dose rate under 20 mSv/h and a decay heat less than 10 W/m3, after 50 years of cooling, it can be classified as medium level waste (MLW). With this classification, all LLW and most MLW may be recycled, while HLW has to be permanently disposed of.  Accidental release: The radiological effect on the public due to the release of a conservative quantity of activated material (1–10 kg) must not exceed 50 mSv.  Maintenance: The dose rate produced by a material placed around the plasma chamber (first wall, divertors) should not exceed 104 Gy/h, after one day of cooling. The primary low activation materials being developed by the international fusion materials community are:

2. Material selection criteria Structural materials at the first wall of the fusion chamber adjacent to the plasma will be subjected to high heat loads, neutron and energetic particle fluxes. This decreases the lifetime of the first wall material drastically and increases the cost of energy production remarkably. The general requirements for materials in the power-generating component can be given as below [30,31]:  Attractive high temperature physical and mechanical properties, which include tensile strength, creep strength, impact toughness, and fatigue.  Broad compatibility with cooling fluids and gases.  Easy fabrication with multiple processes.  Resistant to 14 MeV neutrons induced displacement damage.  Resistant to helium and hydrogen produced by nuclear reactions.  High heat conductivity independent of radiation damage level.  Low swelling or void formation. Dimensional stability is essential for permanent components.  Low activation from 14 MeV neutrons.  Low neutron absorption cross sections.

3. Primary low activation materials Low activation property is a very important criterion for selecting structural materials for fusion applications. Some ‘evaluation parameters’ to decide whether the material is a ‘low activation’ one, or not, can be summarized as follows [32]:

(1) ferritic/martensitic steels, (2) vanadium alloys and (3) SiCf/SiC composites [32–35]. Ferritic/martensitic steels have been considered as candidate structural materials for first wall and blanket structures of fusion power plants since the late 1970s, when they showed more swelling resistant than austenitic stainless steels in fast reactor irradiation. Alloying elements of reduced-activation or low activation steels producing long-lived radioactive isotopes during neutron irradiation must be eliminated or replaced. Elements that must be eliminated or minimized include Mo, Ni, Nb, Cu and N. Besides the elements listed above, there are various other elements that must be restricted to extremely low levels. Such elements could appear in the materials, as tramp impurities and include Ag, Ho, Bi, Co, Sm, Lu, Dy, Gd, and Cd [36]. Therefore, in the low activation steels, tungsten replaced Mo to produce Cr–W steels as alternatives to the Cr–Mo steel and small amounts of Ta replaced the Nb. Ferritic steels provides improved resistance to thermal stresses compared to austenitic stainless steels for a fusion power plant operating in a pulsed mode due to having higher thermal conductivity and lower thermal expansion coefficient than austenitic steels [37]. The ferritic steel 9Cr–2WVTa has been chosen as structural material for ARIES-ST fusion reactor [38–42] and HT-9 for KOYO [43] and 10X9VFA for Russian Demo design [44]. Ferritic steel has a very low NWL limit of 1.5 MW/m2 if the interface temperature is 500 °C. If the interface temperature is dropped to 450 °C, the NWL increases to 2.9 MW/m2 but thermal efficiency decreases [45]. The selection of the appropriate matrix and alloying elements must have

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enough fast decay of the induced radioactivity to the suitable level. And also eliminating trace elements which give rise to the harmful long-lived radionuclide is very important in producing low activation ferritic steels [46]. Vanadium alloys have high-performance mechanical and physical properties, good fabricability, high neutron fluence support capability, accommodation ability of high heat loads, low degradation under neutron irradiation, low decay heat, low waste disposal rating and potential for high performance and long operating lifetime in a fusion environment [47–50]. A great number of studies have been done to find an optimum vanadium alloy composition for use in fusion reactors. Vanadium alloys have been selected and evaluated in major design studies involving the blanket selection study (BCSS) [51], the Tokamak Power Systems Study (TPSS) [52], the TITAN Reversed Field Pinch Design [53], the ARIES-II Tokamak Design [54] and the ARIES-RS Tokamak Design [55]. These studies are focused on the vanadium–chromium–titanium alloy system with (3–9 wt%) Cr and (3–10 wt%) Ti. The V–4Cr– 4Ti alloy became the reference composition. These three elements are mutually soluble in each other at elevated temperatures and have low activation characteristics [48]. The minimum operating temperature limit for V–Cr–Ti alloys is 400 °C since marked irradiation hardening occurs below this temperature while the maximum operating temperature limit for these alloys is between 700 and 750 °C due to thermal creep [56]. Vanadium alloys offer higher NWL capabilities of 3.2 and 4.7 MW/m2 for interface temperatures of 600 and 550 °C, respectively [45]. SiC fiber-reinforced SiC matrix composites (SiCf/SiC) are primary composites to be considered as structural material for fusion reactors because of their low-induced reactivity by 14-MeV neutron irradiation, their excellent high-temperature strength, and corrosion resistance [57– 62]. In particular, the achievability of a high thermal efficiency of the power plant system is one of the advantages of a SiCf/SiC material in a fusion reactor. Furthermore, a very great advantage of SiCf/SiC composites with respect to environmental problems is their low activation behavior [32,34]. There are three major fusion reactor design concepts using SiCf/SiC composites. These are ARIES-I, ARIES IV and ARIES-AT [63], designed by USA, DREAM [64–68], designed in Japan and TAURO [69], designed in the EU. SiCf/SiC composites are also considered as first wall structure in PROMETHEUS IFE reactor design [70]. The maximum operating temperature for SiCf/ SiC due to void swelling concerns is taken to be 990 ± 40 °C [56]. SiCf/SiC composites are limited to a low NWL of 2.5 MW/m2 at interface temperature of 700 °C and this makes their performance poor [45].

presence of impurities and alloying elements. Micro structural defects on atomic level will lead to macro structural defects. Nuclear radiation tends to destroy crystal order and alter properties by generating several types of defects. The most important material damage types leading to failure are atomic displacement and gas production, which will be addressed below.

4. Material damage under neutron irradiation

In fusion blankets, another very serious damage mechanism for structural materials will be gas production in the metallic lattice resulting from diverse nuclear reactions, mainly through (n, p) and (n, a) and to some extent through (n, d) and (n, t) reactions above a certain threshold energy.

All metals have a crystalline microstructure and crystalline materials have their atoms in long-range order. Imperfections in the atomic arrangement always exist due to the

4.1. Atomic displacement under neutron irradiation The displacement of an atom from its lattice position results from transferring threshold energy, typically of the order of few dozens of electron volts, to the target. The displacement per atom (DPA) cross section is the integral effect of displacements induced directly, by the neutron– nuclei interaction and, indirectly, via the interaction between high energy knocks on atoms and the target atoms in a cascade type process [71]. Vacancy-interstitial pairs (Frenkel defects) are formed when energetic particles collide with atoms ejecting them from stable lattice sites. There may be a cascading effect because of a knocked-on atom having received sufficient energy to eject more atoms. These displaced atoms may finally lose their energy and occupy positions other than normal lattice sites, thus becoming interstitials. The presence of interstitials and vacancies makes it more difficult for dislocations to move through the lattice, usually increasing the strength and reducing the ductility of a material. In fact, the production of Frenkel defects is the dominating cause for property changes (swelling, creep, embrittlement, sputtering, etc.) in metallic and most other materials in fission and fusion reactors. Atomic displacements are the fundamental process of radiation damage in metals. Hence, this topic has been investigated in a great number of scientific works [71–74]. In nuclear reactors, the atomic displacements are caused due primarily to the scattering of fast neutrons. The neutrons >1 MeV energy carry practically the total neutron energy and are responsible for the major amount of lattice disruption. Thermal neutrons themselves do not cause atomic displacement. However, (n, c) reactions produce high energy gamma-rays in the range 5–10 MeV. The recoil energy given to the nucleus emitting the gamma is often high enough to cause several Frenkel defects with resultant property changes. In fusion or fast hybrid blankets, displacements of the atoms from their lattice sites as a result of collisions with highly energetic fusion neutrons will be a damage mechanism for structural materials at much higher levels than in conventional fission reactors. 4.2. Gas production

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Materials suffer from embrittlement due to gas bubble formation even for fission applications, which is in general at lower MeV range. As the energy of the fusion neutrons with 14 MeV is significantly higher than the energy of the fission neutrons (2 MeV), gas production in fusion reactors might build up at levels several orders of magnitude higher than in fission reactors. The hydrogen isotopes will diffuse out of the metallic lattice under high operation temperatures, but a-particle’s will remain in metal and generate helium gas bubbles. These reactions will limit the lifetime of the first wall to few years. The highest material damage will occur in the first wall as it will be exposed to the highest neutron, gamma ray and charged particle currents, which are produced in the fusion chamber.

5. Hybrid blanket structure The neutronic calculations were conducted on a previously introduced experimental hybrid blanket geometry [8,9]. Side view of (DT) fusion driven IFE type hybrid blanket is shown in Fig. 1. In this study, the low activation materials namely, 9Cr2VWTa, V–4Cr–4Ti or SiCf/SiC composite were considered as first wall and fuel cladding material used in a (DT) fusion (IFE type) driven hybrid reactor. The first wall separates the fuel zone from the fusion plasma chamber. The latter had a zone thickness of 12.5 cm which was made up of UO2 fuel as nuclear fuel. Vmoderator/Vfuel was taken as 2 and the fuel rods were arranged in hexagonal geometry as 10 rows having pitch

Fig. 1. Cross sectional view of the investigated blanket (dimensions are given in cm).

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length = 1.25 cm radially. Natural lithium was used as coolants to breed tritium for fusion driver and supply heat transfer. Three Li2O zones and three carbon zones forming sandwich structure were used to enhance the tritium breeding in the blanket and to reduce neutron leakage out the blanket. Among solid tritium breeders Li2O was selected due to its high breeding potential, low vapor pressure even at high temperature, low activation and high lithium density. Carbon, a very common reflector material, was chosen to use in the blanket because of its good reflectivity property.

In order to determine neutron flux through the blanket, calculations were conducted with the help of SCALE4.3 System by solving the Boltzmann transport equation with code XSDRNPM [75] in S8-P3 approximation with Gaussian quadratures [76] using the 238 groups library, derived from ENDF/B-V [77] and CLAW-IV [78]. The resonance calculations in the fissionable fuel element cell were performed with

40 9Cr-2WVTa SiC f /SiC V-4Cr-4Ti

35

30

25

DPA/FPY

6. Calculation procedure

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20

15

10

5

 BONAMI [79] for unresolved resonances and  NITAWL-II [80] for resolved resonances. CSAS control module [81] was used to produce the resonance self-shielded weighted cross sections for XSDRNPM. Calculations were performed by using a NWL of 2 MW/m2 that is suitable for the investigated materials, 9Cr2VWTa, V–4Cr–4Ti and SiCf/SiC. 7. Numerical results and discussion 7.1. Displacement per atom Material damage criteria must satisfy both the DPA and helium production limits in design of reactor structures. At present there is no consensus on a DPA limit backed up by experimental data due to the lack of an intense fusion neutron source. In an earlier work [82] higher limits for fusion reactors, namely DPA = 300–1000, were proposed whereas, more recent studies [83,84] assumed a lower damage limit as DPA = 165. Damage limit for DPA can be significantly higher than those values practically. In the present work, a very conservative limit of DPA = 100 is chosen. DPA values for the first wall and clad materials were computed in the investigated hybrid reactor. DPA per (DT) neutron can be defined as follows: Z Z X DPA ¼ U  dE  dt ð1Þ ðn;dpaÞ

P where t is the irradiation time, ðn;dpaÞ is DPA macroscopic cross section, U is neutron flux, E is the neutron energy. Fig. 2 shows the DPA change with respect to radial distance from the first wall to the outer fuel zone under a

0 300

302

304

306

308

310

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Radial distance [cm] Fig. 2. DPA profile through the first wall and fuel zone for the investigated structural materials under a NWL of 2 MW/m2.

NWL of 2 MW/m2 for the investigated materials after a plant operation of 1 full power year (FPY). One can see that DPA/FPY decreases exponentially in radial direction due to the exponential decreasing nature of the neutron flux. Neutron spectrum softening takes place along with the deeper penetration of neutrons in the fuel zone. After 1 year operation period, DPA/FPY value is around 35, 38 and 37 at the first wall structure for the materials 9Cr–2WVTa, V–4Cr–4Ti and SiCf/SiC composite, respectively. For this reason, the replacement of first wall structure would be needed every 3 years for the investigated low activation materials because of DPA damage. On the other hand, DPA/FPY values are 33, 36 and 35 for the fuel clad next to the first wall they drop down to 14, 18 and 19 at the outermost fuel rods for the materials 9Cr– 2WVTa, V–4Cr–4Ti and SiCf/SiC composite, respectively. Fuel clads also will require replacement every 3–7 years depending on the position of the fuel rod in the fuel zone for the investigated materials. 7.2. Helium gas production Blink et al. [83] and Perlado et al. [84] suggested a helium limit of 500 atomic parts per million (appm). And

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also in this study this limit is taken as reference value for helium production. Helium gas production per (DT) fusion neutron was calculated in a similar manner given below: Z Z X He generation ¼ U  dE  dt ð2Þ ðn;aÞ

P

where ðn;aÞ is the macroscopic cross section for helium generation. Fig. 3 depicts the helium production versus radial distance for investigated materials in the hybrid blanket after an operation period of 1 year. As in DPA, helium production decreases through the blanket from first wall to outermost fuel rod. The lowest helium production value is found for V–4Cr–4Ti first wall that is 127 appm/FPY whereas the highest one is found for SiCf/SiC composite which is 5638 appm/FPY. In addition, the helium production value is computed as 267 appm/FPY for ferritic steel at first wall. First wall replacement every 2 years, 4 years and 1 month will be needed for the materials, 9Cr–2WVTa, V–4Cr–4Ti

10

and SiCf/SiC composite, respectively. On the other hand, for fuel clads of these materials, helium production values range 236–67 appm/FPY, 127–30 appm/FPY and 5100– 1608 appm/FPY, through the fuel zone, respectively. Again, when considering the helium generation limit, the lifetime will be in the range of 0.5–7.5 years, 4–17 years, 1– 4 months depending on the fuel rod position in the fuel zone for the materials, 9Cr–2WVTa, V–4Cr–4Ti and SiCf/ SiC composite, respectively. Although the DPA values for three candidate low activation materials are very similar, helium generation values are quite different for each other. A very huge amount of helium generation in SiCf/SiC composite is directly related to its very high (n, a) cross section under energetic fusion neutrons. Therefore, helium generation in this material decreases its lifetime considerably in the reactor structure. Radiation damage criteria for a structural material must comply with both the DPA and helium production limits. Therefore, the replacement of first wall would be needed every 2 years, 3 years and 1 month for 9Cr–2WVTa, V– 4Cr–4Ti and SiCf/SiC composite, respectively. 8. Conclusions

4

Helium production [appm/FPY]

Under the lightness of the numerical results, the main conclusions for this study can be cited as below: 10

3

10

2

10

1

10

0

 Vanadium alloy, V–4Cr–4Ti showed the best performance with respect to radiation damage criteria whereas SiCf/SiC composite exhibited the lowest performance due to its very high (n, a) cross sections.  All the investigated materials would be needed a replacement during the lifetime of the reactor (30 years) that increases the cost of electricity generation considerably. Determining materials behavior in fusion environment is not available now due to the lack of an intense fusion neutron source at 14 MeV. Development of an intense source of 14 MeV neutrons comes as an important subject in order to get realistic data about the irradiation behavior of structural materials [85,86]. References

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Radial distance [cm] Fig. 3. Helium generation profile through the first wall and fuel zone for the investigated structural materials under a NWL of 2 MW/m2 (DPA profile through the first wall and fuel zone for the investigated structural materials under a NWL of 2 MW/m2).

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