Selection of fuel cladding material for nuclear fission reactors

Engineering Failure Analysis 18 (2011) 1943–1962

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Review

Selection of fuel cladding material for nuclear fission reactors C.R.F. Azevedo ⇑ Universidade de São Paulo, Dep. Engenharia Metalúrgica e de Materiais, Escola Politécnica, Brazil

a r t i c l e

i n f o

Article history: Received 27 July 2010 Accepted 8 June 2011 Available online 21 June 2011 Keywords: Nuclear fission reactor Materials selection Fuel cladding Failure modes

a b s t r a c t The growing understanding of the link between carbon emissions and global warming has been promoting a discussion on the environmental and safety viability of nuclear power generation. Current open fuel cycle reactors, however, result in low energy efficiency and produce large volumes of nuclear waste. More advanced nuclear reactors, which are currently under investigation, are expected to allow more efficient and safer use of nuclear energy. In all these cases, the fuel cladding is the most important safety barrier in fission nuclear reactors, as it restrains most of the radioactive fission products within its volume. The selection of fuel cladding material is based on many design constraints, such as neutron absorption cross section, service temperature, mechanical strength, toughness, neutron radiation resistance, thermal expansion, thermal conductivity, and chemical compatibility. The present paper reviews the selection of nuclear fuel cladding materials since the early reactors, illustrating some of the main failure modes and briefly discussing the challenges facing the development of fuel cladding materials for generation IV reactors. Ó 2011 Elsevier Ltd. All rights reserved.

1. Introduction The growing understanding of the link between carbon emissions and global warming and the rapid economic growth of some developing countries (such as India and China, whose electricity generation are still strongly based of fossil fuels) produced the current global energy scenario, which is expecting a rapid growth on the share of ‘‘green’’ power generation in the global electricity energy matrix. According to recent statistics, electricity generation from fossil fuels still accounts for approximately 33% of the carbon entering the atmosphere of the planet every year, while the share of renewable energy in the global electricity generation is around 18% and the share of nuclear energy accounts for approximately 15%. Under the Kyoto Protocol, most of the industrialised countries committed themselves to a reduction of 5.2% (from 1990 level) on the emission of greenhouse gases by the year 2012. As a consequence, some of these nations are now facing the dilemma between the growing energy demands and the need to reduce the greenhouse emissions, which is also leading to a discussion on their nuclear energy policies (see Fig. 1) [1–6]. There are important issues concerning the safety and management of nuclear power generation. For instance, high-level nuclear waste must be isolated within safety containers, which are stored for periods of approximately 50 years to allow its radioactivity to decay to one thousandth of its initial value. In addition, the final disposal of high-level nuclear waste requires its full isolation from the environment for much longer periods (1000 years). As a matter of fact, current open nuclear fuel cycle results in low energy efficiency, converting only 1% of the U energy content into electricity and producing large volumes of nuclear waste. In France, Japan, and UK, the used nuclear fuel can be recycled to recover U and Pu, which can be reprocessed into new nuclear fuel, increasing the energy efficiency and minimising the amount of long-lived radioactive elements. In this sense, more advanced and better performing nuclear reactors based on the concept of closed nuclear fuel cycle ⇑ Tel.: +55 11 3091 5619; fax: +55 11 3081 5243. E-mail address: [email protected] 1350-6307/$ - see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.engfailanal.2011.06.010

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C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962

Fig. 1. Emissions of fossil CO2 (g/kW h) electricity delivered to household customer, showing the low level of emissions for hydropower, wind power, solar cell, biofuel and nuclear power generation. CHP is combined heat and power plant and SOFC is solid oxide fuel cell [3].

are expected to remove most of the long-term radiotoxicity, simplifying the management and safety of nuclear waste repositories [3,5–8]. The history of the development of the nuclear power generation throughout the 20th century can be roughly divided into three distinct parts [5,6,8–13]:  Before the second world war (WWII), basic research was based on the understanding of atomic fission and control of nuclear chain reaction, see Table 1.  During the WWII, R&D activities were focused on the production of nuclear weapons (construction of nuclear breeder reactors to produce Pu) and, just after the WWII, on the construction of the nuclear power reactors for submarines, see Table 2.  After the 1950s, R&D activities shifted towards the civilian nuclear power generation, see Table 3. As a consequence, nuclear power generation showed a substantial growth in the number of reactors during the 1960s and 1970s, followed by a significant decline in this growth rate during 1980s and 1990s, mainly because of safety and cost issues, keeping the

Table 1 Understanding the nuclear fission reaction [5,6,8–13]. Year

Information

1896

Becquerel discovered invisible emanations (b radiation and a particles) from U, which were latter defined as radioactivity by Pierre and Marie Curie Rutherford and Soddy showed that elements, such as U and Th, became different elements through the process of radioactive decay Rutherford fired a particles from a Ra source into N gas and found that nuclear rearrangement was occurring, with the formation of O gas by particle absorption. This theory was supported by further experimental work and new radioactive substances were discovered Chadwick discovered the neutron in 1932. Cockcroft and Walton produced nuclear transformations by bombarding atoms with accelerated protons, creating artificial radionuclides Nuclear fission was first experimentally achieved by Fermi in Rome, when his team bombarded U with neutrons, producing a variety of artificial radionuclides Nuclear fission was achieved by Hahn and Strassman, who bombarded the nucleus of a U atom with neutrons, causing it to split into new lighter elements, releasing energy. It was soon recognised that a self-sustaining nuclear chain reaction releasing enormous amount of energy would be feasible. Bohr proposed that fission was more likely to occur in the U235 isotope than in the U238 and that fission would occur more efficiently with the use of slow neutrons. Slizard and Fermi proposed the use of a moderator material to slow down the emitted neutrons from the fission reaction to sustain a chain reaction. Perrin demonstrated that a chain-reaction could be sustained in a U-H2O mixture provided that external neutrons were injected into the system Bretscher and Feather showed that the neutron absorption of slow neutrons in U235 and U238 led to the formation of a more fissionable element with atomic number equal to 94 and named plutonium (Pu)

1902 1919 1932 1934 1938

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C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962

1945

Table 2 Nuclear fission: the military years [5,6,8–13]. Year

Information

1939

World War II began. USA and Germany raced to build the first atomic bomb. A massive ‘‘research and product’’ programme (Manhattan Project) was launched by the USA to produce uranium and plutonium for developing nuclear weapons Creation of the first nuclear reactor, known as Chicago Pile-1 in the USA. Large reactors were built at the Hanford Site to breed Pu for use in nuclear weapons. The result was the manufacture of atomic bombs dropped on Japan in 1945. By the end of WWII researchers predicted and described the applications of nuclear energy to produce electricity The Atomic Energy Act of 1946 (McMahon Act) determined how the US federal government would control and manage the nuclear technology it had jointly developed with its wartime allies (Britain and Canada). It also ruled that nuclear weapon development and nuclear power management would be under civilian, rather than military, control, and established the US Atomic Energy Commission for this purpose The US Congress authorised the construction of a nuclear-powered submarine (Nautilus) for the US Navy, which was powered by a pressurised water reactor (PWR), using Zr alloy as fuel cladding material, light water (H2O) as coolant and moderator material and enriched uranium as fuel material. PWR became the most widely used power reactor

1942 1945 1946

1951

Table 3 Nuclear fission: civilian energy generation [5,6,8–13]. Year

Information

1951

The experimental breeder reactor (EBR-I, 200 kWe) started operation in USA, becoming the world’s first electricity-generating nuclear power plant, producing electricity from Pu fuel The Soviet Union developed the RBMK reactor (AM-1, 5 MWe), which used graphite as moderating material, H2O as coolant material, Zr alloy as cladding material and natural U as fuel. It became the world’s first nuclear power station designed for commercial use The UK developed the Magnox Reactor (50 MWe), which used unenriched U metal fuel, Mg alloy fuel cladding, graphite moderator and CO2 coolant. In 1956, fours reactors at Calder Hall Power station were connected to the national grid The USA began the operation of a large scale nuclear power plant (Shippingport, 60 MWe), with PWR reactors using Zr alloy as cladding material, H2O as coolant and moderator material, and enriched U as fuel France developed its own nuclear, whose design was similar to the UK’s Magnox Development of CANDU reactors, characterised by the use of natural U fuel, heavy water (D2O) as coolant/moderator material and Zr alloy as fuel cladding material UK developed the advanced gas-cooled reactor (AGR), which used CO2 as coolant material, stainless steel as cladding material and enriched U as fuel Between 1973 and early 1990, nuclear energy’s share of US electricity increased from 4 per cent to 20 per cent, while oil’s share dropped from 17 per cent to 4 per cent. In France nuclear contribution to electricity production has risen from 8% in 1974 to 78% today Economical and safety factors stopped the growth of the nuclear power industry. A serious accident occurred at the power plant at Three Miles Island (PWR) in the USA and after this event no US nuclear plants were ordered. The downturn affected Europe and many voters rejected nuclear power Explosion of a nuclear reactor (RBMK) at Chernobyl in 1986. The Soviet Union’s nuclear programme was held up Fukushima nuclear plant accident (BWR). Review of the German nuclear program. German chancellor Angela Merkel declared that her government plans to close down all its nuclear power plants

1954 1956 1957 1959 1962 1964 1973 1979

1986 2011

global share of nuclear power for electricity generation at a constant level around 15% [5–14]. At the end of 2008, there were 438 nuclear power reactors in operation in the world, which most of them being classified as generation II (GII) reactors, see Table 4. Current GIII reactors include just few improvements in the design of GII reactors to achieve longer operational life, higher thermal efficiency and better safety systems. Recent expansion of the nuclear industry remains centred in Asia, but there are political changes in Europe and USA, which might increase the share of nuclear energy in the global electrical energy matrix, despite the Fukushima accident and the current position of the German Government to close down all its nuclear power plants. The UK government published in 2008 a document stating the importance of nuclear energy on its energy matrix to meet carbon reduction targets and ensure secure energy supplies. There are currently 19 nuclear reactors in operation in the UK, but a significant amount of them will be decommissioned by 2023, urging new investments on either the construction of new nuclear plants or gross investments on the use of renewable energy sources. For instance, in 2006 the UK nuclear plants generated 19% of UK electricity (against 36% from gas and 38% from coal), but in 2007 this share dropped to 15% and in 2008 to 13.5%, mainly due to technical problems with these old reactors. By contrast, the nuclear energy share in China’s and India’s electrical energy matrixes is currently insignificant, but these countries are presenting a fast economical growth, with increasing demands for electricity generation, which is still largely dependent on fossil fuels for both countries (80% for China and 70% for India). The Chinese government plans to increase its nuclear capacity to 200 GWe by 2030 (20% of the total electricity), while India expects to supply 25% of its electricity from nuclear power by 2050. In 2004 there were already six European countries generating over half their electricity from nuclear energy (France, 78.1%; Lithuania, 72.1%; Slovakia, 55.2%; Belgium, 55.1%; Sweden, 51.8%; and Ukraine, 51.1%) [4,5,14–19]. GIV reactors are a set of theoretical designs currently under investigation in order to use advanced fuels and operate at higher burn-up rates, minimising nuclear waste and costs and increasing energy efficiency, sustainability and safety. Most of these GIV designs, however, are not expected to be available for commercial construction before 2030, presenting, therefore,

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Table 4 Common design of nuclear fission reactors (GI and GII), showing the variety of operating conditions and materials used for fuel, moderator, coolant and fuel cladding [1,34,68].



Name of the reactor

Fuel material

Moderator material

Primary coolant material

Fuel cladding material

Operation parameters

GCR, Gas-cooled reactor (GI)

Natural UO2

Graphite

CO2

Magnesium alloys

PWR, Pressurised water reactors (GII)

Enriched UO2 MOX

H2O

H2O

Zirconium alloys

BWR, Boiling water reactor (GII)

Enriched UO2 MOX

H2O

H2O

Zirconium alloys

PHWR, Pressurised heavy water reactor (GII)

Natural UO2

D2O

D2O

Zirconium alloys

AGR, Advanced gas-cooled reactor (GII)

Enriched UO2

Graphite

CO2

Austenitic stainless steel

RBMK, High power channel-type reactor (GII)

Enriched UO2

Graphite

H2O

Zirconium alloys

P: 2.7 MPa Tinlet: 250 °C Toutlet: 400 °C P: 15 MPa Tinlet: 290 °C Toutlet: 325 °C P: 7 MPa Tinlet: 280 °C Toutlet: 330 °C P: 10 MPa Tinlet: 270 °C Toutlet: 310 P: 4 MPa Tinlet: 340 °C Toutlet: 635 °C P: 8 MPa Tinlet: 260 °C Toutlet: 290 °C

MOX: mixed oxide fuel.

interesting challenges for the development and selection of materials for in-core and out-core components [4,5,20–24]. The present paper reviews the selection of nuclear fuel cladding materials since the early reactors and illustrates some of the main in-service failure modes, briefly discussing the various challenges facing the development of fuel cladding materials for generation IV reactors.

2. Basics of the core of nuclear fission reactors Nuclear fission reactors are based on the control of a nuclear reaction, where a large fissile atomic nucleus (for instance, U or 239Pu) absorbs a neutron and undergoes a nuclear fission reaction, with its nucleus splitting into lighter nuclei and releasing a large amount of energy. For instance, a kilogram of uranium (235U) contains approximately three million times more energy than a kilogram of coal burned conventionally. The heat produced by several self-sustained nuclear fission reactions can be converted into electricity by the use of heat exchangers and steam turbines. The total binding energy per nucleon (see Fig. 2) is the energy required to break apart a nucleus into its constituent nucleons and it is used as a measurement of nuclear stability of the elements. High values of binding energy represent greater element stability, with iron being the most stable element [34]. When one nucleus is converted into another or others of higher binding energy there is a release of energy. For instance, two nuclei of low binding energy can combine by nuclear fusion to form a more stable nucleus (see extreme left region of Fig. 2), while nuclei of high binding energy can move into a more stable configuration by atomic fission, releasing a comparatively smaller quantity of energy than nuclear fusion (see extreme right region of Fig. 2) [1,25–29]. During atomic fission, the parent nucleus absorbs a neutron (through neutron bombardment), becoming unstable and decaying into fission fragments of unequal mass combined with kinetic energy, releasing during this process c rays, a and b particles and several fast neutrons (13,800 km/s), with a kinetic energy of about 2 MeV each. These fast neutrons could induce further fission events in other fissile nuclei, but in practice these neutrons must be slowed down into thermal neutrons (kinetic energy of about 0.025 eV and velocity around 2.2 km/s at room temperature) by a neutron moderator material in order to produce further fission reactions. Thermal neutrons have a larger effective neutron absorption cross-section than fast neutrons and can be frequently absorbed by fissile nuclei, creating unstable fissile isotopes, which lead to further nuclear fission reactions, producing a sustainable chain reaction [25–29]. Additionally, neutron moderator materials should present a low value of neutron absorption cross section to maximise the number of thermal neutrons available for fission. Efficient neutron moderator materials must, therefore, present high values of neutron scattering cross-section, rs, and low values of neutron absorption cross-section, ra (see Fig. 3). Hydrogen, for instance, presents a good neutron scattering, but it features a high value of neutron absorption cross section, so reactors using H2O as a moderator material are forced to employ enriched UO2 to compensate the neutron losses imposed by the absorption [1]. By contrast, heavy water (DO2) features a low value of neutron absorption cross section. The effectiveness of a neutron moderator material, however, also depends on the mass of the nucleus of the moderator material so a moderating ratio factor, M (see Eq. (1)) can be defined and used for the selection of efficient neutron moderator materials [1,25–29]. For instance, a high moderating ratio associated with a lower atomic number (see upper left region of 235

C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962

1947

Fig. 2. Binding energy per nucleon for pure elements, showing fuels for nuclear fusion and nuclear fission [1,28,34].

Fig. 4) indicates efficient neutron moderator materials, such as D2O, which allows the use of unenriched UO2. The selection of moderator materials for early European nuclear programs was very limited in terms of neutron economy, as there was no enriched UO2 fuel available, leaving the materials selection for moderator materials between DO2 and graphite. In the USA, however, H2O was soon adopted as a moderator material associated with the use of enriched UO2 fuel [1,5–8,25–29].

M ¼ nrs=ra

ð1Þ

where n is the fraction of neutron energy lost per scattering event; n = 6/(3A + 1), where A is the atomic mass number. The nuclear chain reaction is controlled not only by the velocity of the emitted neutrons, but also by the number of emitted neutrons available in the system. This can be done by the use of neutron absorption elements in the core of the reactor in order to gain further control on the nuclear chain reaction. These control elements absorb a fraction of the emitted neutrons, controlling the rate of fission of the fuel by quenching the chain reaction that generates them. The best materials for neutron control rods should therefore present high absorption cross-sections (such as Boron, seen on the extreme right of Fig. 3), without transmuting into fissionable material [1]. During the design of the nuclear reactor’s core, there is a trade-off between multiple design constraints, such as maximising the neutron economy, the service temperature and the thermal efficiency of the reactor. As seen in Table 4, there are currently many designs of GII nuclear reactors (based on thermal neutrons), applying different materials for fuel, moderator, control rods, coolant and fuel cladding. Pressurised water reactors (PWR) constitute the vast majority of all western nuclear power plants (61%) and use zirconium alloys as fuel cladding material [1,4,14,25–29]. 3. Materials selection for nuclear fuel cladding The fission of U atoms produces radioactive materials, which emit neutrons, c rays along with a and b particles. These particles can be lethal to humans, so the use of fuel cladding keeps the radioactive materials isolated from the coolant/moderator, which surrounds the cladding and maintains the fuel rods cooled. The material for fuel cladding is selected after other design aspects of the reactor’s core have already been decided, such as the nuclear fuel and the moderator and coolant materials, defining, therefore, the multiple design constraints for the selection of the fuel cladding material. These constraints includes the neutron absorption cross section, the maximum service temperature, the creep resistance, the mechanical strength, the toughness, the neutron radiation resistance, the thermal expansion, the thermal conductivity and the chemical compatibility with fissile products and coolant, moderator and fuel materials [1,30–34]. For instance, the cladding material should be transparent to neutrons, meaning that the material should present a low value for the

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Fig. 3. Neutron scattering cross-section versus neutron absorption cross-section for pure elements, indicating the best materials for moderators (upper left) and control rods (lower right) [1,28,34].

Fig. 4. Moderating ratio versus atomic weight for pure elements, indicating the most effective neutron moderator materials (see upper left) [1,28,34].

neutron absorption cross-section to minimise neutron losses (such as Mg, Be, Si, Al, and Zr, see Fig. 5). Additionally, the cladding material should present an acceptable service temperature to increase the thermal efficiency of the reactor, leading to a trade-off between service temperature and neutron transparency. Table 5 shows the calculated values of the neutron absorption cross section, the yield strength and the relative neutron absorption for a given stress (barns/MPa) for pure elements in

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relation to Zr (the most used fuel cladding material at the moment). The results indicate that many multi-component systems combining Be, C, Mg, Zr, Si, and O (including metallic and ceramic systems) might be considered as potential new materials for nuclear fuel cladding in terms of neutron economy. As a matter of fact, Al, Mg, and Zr alloys have already been used as fuel cladding materials; with great advantages for Zr alloys at the moment [30,31]. The fuel cladding material should also be radiation resistant and there are several ways in which fission-generated neutrons can interact with the crystal lattice of the cladding, producing point, line and volume defects, such as dislocation loops and void and self-interstitial clusters. As a result, there will be a change in the mechanical properties of the material, as long as the service temperature is below 40% of the homologous temperature of the material [35]. Furthermore, radiation can also alter the redistribution of alloying elements, the stability of phases and the kinetics of phase transformations, with consequences not only to the mechanical properties, but also to corrosion and creep resistance of the cladding material. The cladding material should be able to withstand the service temperatures (in normal or accident conditions), making thermal creep resistance and microstructural stability key issues for the selection of materials as there is a trend to increase the service temperatures to achieve higher thermal efficiency for the reactor. Finally, the cladding material should be corrosion/oxidation resistant to the environment (coolant/moderator/fuel/fission products) and should present high thermal conductivity to increase the energy efficiency of the reactor; and low thermal expansion coefficient to minimise the thermal stresses in the cladding/pellet interface [8,29–33,35]. It is worthwhile to observe the historic evolution of the design constraints trade-off in order to understand the trends for new materials for GIV reactors. According to Howe [9], during the early 1940s the materials challenges for the nuclear technology grew rapidly for the construction of Pu-producing reactors. Beryllium (hexagonal close packed) was considered as a candidate nuclear fuel cladding material since early 1950s due to its unique nuclear properties (see Fig. 5 and Table 5). This material, however, found very limited use as structural material in nuclear applications due to its low ductility even in the absence of irradiation, as Be has only two operating slip modes (basal slip on {0 0 0 1} planes and prismatic slip on {1 0 1 0} planes) at low temperatures. Additionally, Be undergoes significant embrittlement at low doses of irradiation due to the irradiation growth of anisotropic crystals and the formation of He bubbles as a result of (n, a) reactions. Additionally, Be also presented catastrophic corrosion behaviour above 500 °C. Work on Be–Ca alloys was developed to improve the corrosion resistance and optimise the mechanical properties and the radiation-induced embrittlement. Cost, brittleness (especially after irradiation) and toxicity ruled out the commercial application of the metal as cladding material. Nowadays, the main research for the use of Be is in area of fusion reactors, but safety aspects, such as its reactivity with steam and water under accident conditions, and its radiation resistance are unique challenges for its successful application [36–40].

Fig. 5. Temperature resistance (given as melting temperature) versus neutron absorption cross-section for pure elements, showing a trend for new cladding materials with a trade-off between higher service temperature and lower neutron transparency (see arrow) [1,28,34].

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C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962 Table 5 Calculated effective neutron absorption cross section (neutron absorption cross section per unit of yield strength) for pure elements in comparison to Zr [31,34].

*

Elements

Neutron absorption cross section (Barns)

Yield strength (MPa)*

Relative effective neutron absorption cross section in relation to Zr

Be C Mg Si Zr Al Mo Cr Nb Fe Ni V Sn

0.009 0.004 0.063 0.160 0.185 0.231 2.480 3.050 1.150 2.550 4.430 5.040 0.630

200–350 24–28 65–100 165–180 135–310 30–40 170–350 185–280 75–95 110–165 80–280 125–180 7–15

0.04 0.2 1 1 1 8 10 15 15 20 30 40 70

Values for pure commercial alloys in the annealed condition.

The investigation of aluminium (face-centred cubic) cladding of U bars for research reactors began in 1942 and the activities of the Metallurgical Laboratory of the University of Chicago focused on the design of a full-scale water-cooled reactor, using U slugs sealed in Al cans inside Al tubes, creating the fuel cladding concept. The materials selection was probably based on simple design constraints such as maximum service temperature, aqueous corrosion resistance and neutron economy. The researchers, however, had to develop the manufacturing process for leak tight canning of fuel slugs and to investigate the working conditions of the water (coolant/moderator) in order to control the corrosion rate of Al. Satisfactory canning of U slugs in Al was finally completed and the B-Reactor (105-B) at Hanford was the first large-scale Pu production reactor in the world (1944), operating at 250 MGW and using graphite as neutron moderator and H2O as a coolant (see Fig. 6). Since then, Al alloys have been used not only as fuel cladding material, but also as fuel matrix and structural material in nuclear, especially in research reactors. Al is indeed a corrosion resistant metal with a relatively low neutron absorption cross section (see Table 5 and Fig. 5), but its maximum service temperature (around 200 °C, limited by oxidation and mechanical strength) was considered unsatisfactory for use in early nuclear power reactors, being replaced in the US by the use of austenitic stainless steel, despite its effective neutron cross section absorption value being 15 times bigger than that for Zr alloys [9,39,41–45]. In the late 1950s, the designers of the gas cooled reactor (GCR, see Table 4) selected Mg alloys as cladding material to withstand working temperatures up to 400 °C and to minimise the neutron absorption (see Table 5 and Fig. 5). The GCR uses graphite as neutron moderator material and CO2 as coolant material, being able to use natural uranium as fuel. There were two main types of GCR reactors: the Magnox reactor, developed by the UK; and the UNGG (Uranium Naturel Graphite Gaz) reactor, developed by France. The first Magnox reactors at Calder Hall were designed principally to produce Pu for nuclear weapons, but it was soon realised that the large quantities of heat generated during the production of Pu could be used to generate power as a useful by-product. The fuel cladding was used to prevent the oxidation of the fuel by the coolant, and the escape of fission products into the circulating gas stream. Additionally, the cladding provided a heat transfer surface to improve the efficiency of heat removal from fission reaction. In this sense, the design constraints for the selection of cladding material were that the material had to be compatible with fuel, fission products and coolant; to present a low neutron absorption cross-section; to offer a feasible processing route; and to feature good creep properties. Only two materials were used as primary fuel cladding materials in GCR reactors: Magnox A12 alloy (Mg with 0.7–0.9%Al) and ZA alloy (Mg with 0.45–0.65%Zr). Designers, however, soon discovered two major drawbacks: Mg alloy restricted the maximum coolant temperature to 500 °C due to creep and oxidation resistance, limiting the thermal efficiency of the reactor; and Mg alloy presented high corrosion rate in water, preventing long-term storage in spent fuel pools. The latter restriction made fuel reprocessing an essential part of the nuclear fuel cycle in both France and UK. In the late 1960s, the Magnox reactor was further developed to advanced gas-cooled reactor (AGR), while in France, the UNGG design was replaced by the pressurised water reactor (PWR) [6,39,46–48]. During the 1960s, the designers of the advanced gas-cooled reactor (AGR, see Table 4) selected austenitic stainless steel (20Cr/25Ni steel) as a fuel cladding material to increase the service temperature to 600 °C, despite of the high value for the effective neutron absorption cross section of this steel, which is approximately 15 times higher than Mg alloys (see Fe, Cr and Ni in Fig. 5 and Table 5). The original concept of the AGR, however, was based on the selection of Be as a cladding material, but when this material was shown to be unsuitable [36–40], the enrichment level of the fuel had to be raised to compensate the neutron losses caused by the use of a stainless steel fuel cladding. The use of austenitic stainless steel along with enriched fuel in AGR reactor allowed a higher thermal efficiency than GCR reactors (40 against 30%) [48]. However, microstructural characterisation of post-irradiated cladding samples soon revealed the chemical interaction between the fission products and the cladding material, causing, for instance, the intergranular attack at the inner surface of cladding by volatile fission

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Fig. 6. Schematic plutonium reactor in Hanford (1944, Manhattan project, first large-scale plutonium production reactor in the world) using Al fuel cladding [41].

products, such as iodine [49–51]. Additionally, post-irradiated cladding samples also featured physical bonding with the fuel, which might lead to in-service failures due to pellet-cladding mechanical interaction (PCI), in the case associated with the deleterious effect of neutron irradiation on the ductility of stainless steel. One of the main challenges facing the further use of austenitic stainless steel as cladding material is indeed its low radiation resistance. Even at low irradiation temperatures, fast neutrons causes hardening and loss of ductility of austenitic stainless steels (see Fig. 7a), while at elevated temperatures He produced by thermal neutron (n, a) reactions induces low ductility due to gas-induced intergranular cracking [52–54]. In 1947 the US Bureau of Mines developed Zr sponge using the Kroll process, allowing the availability of good quality Zr at reasonable prices. The unique properties of Zr (low effective neutron absorption cross section, see Fig. 5 and Table 5; good mechanical properties, high service temperature and corrosion and radiation resistances, especially when compared with austenitic stainless steel, see Figs. 7a and b) made it an ideal cladding material for the US Navy nuclear propulsion program in the 1950s (see PWR reactor in Table 4). By mid 1960s Zr alloys were already the main fuel cladding material in both light and heavy water reactors, despite of its highly anisotropic properties [13,30–32,55–59]. Zr alloys, however, still present major challenges for improving the reactor’s performance because of their complex corrosion mechanisms. For instance, at high temperatures (340 °C) aqueous corrosion of Zr alloys controls the life of PWR fuel cladding. Corrosion is normally accompanied by the ingress of hydrogen into the cladding, which might promote delayed hydride cracking (DHC), degrading its mechanical properties. The DHC mechanism involves: the precipitation of hydrides in the matrix; the dissolution of matrix hydrides from a region surrounding the plastic zone at the tip of a pre-existing crack; the reprecipitation of hydrides at the tip of the crack, with plate-like precipitates oriented perpendicular to the direction of the main stress (reprecipitation and reorientation); and the step-like crack growth along the hydride precipitate by a shear mechanism. The cyclic hydride-induced crack growth is repeated until the crack reaches a critical size, leading to the failure of the cladding [55–58]. Increases in coolant temperature, power up-rates, extended burn-ups, and longer residence times have led to the development of new Zr alloys (see Table 6) by thermo-mechanical processing and microstructural and alloy design, but its maximum service temperature is still around 400 °C, creating immense challenges for further advances on new Zr-alloys [6,13,30–32,55–65]. There is in fact a clear trend to increase the efficiency of nuclear reactor by demanding higher service temperatures, longer in-service life, more corrosive environments and higher neutron doses, challenging the selection of materials for fuel cladding [1,5–8,29–33].

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Fig. 7. Change in tensile stress with neutron irradiation in (a) austenitic stainless steel, AISI 316L; and (b) Zircaloy4. These graphs demonstrate that Zircalloy4 is much more resistant to higher neutron dose than austenitic stainless steel [34].

4. New trends for fuel cladding material The economics of current nuclear power plants is improved through increasing fuel burn-ups, which results in less nuclear waste. There has been a continuous historical increase in fuel burn up: from 20 GWd/tU in GI reactors to 60 GWd/ tU in GII and GIII reactors; and to 200 GWd/tU in GIV reactors. In GIV reactors the chain reaction is generally sustained by fast neutrons associated with the use of highly enriched U or Pu and lack of neutron moderator material. Some important

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C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962 Table 6 Zirconium alloys used for fuel cladding [55]. Elements

Zircaloy 2

Zircaloy 4

Zr–Nb

Zirlo

M5

MDA

E 365

HANA 4

Sn Fe Cr Ni Nb O

1.2–1.7 0.07–0.2 0.05–0.15 0.03–0.08 – 1400 ppm

1.2–1.7 0.18–0.24 0.07–0.13 – – 1400 ppm

– – – – 2.4–2.8 0.13

0.96 0.1 – – 0.99 1430 ppm

– – – – 1.0 1250 ppm

0.8 0.2 0.1 – 0.5 –

1.2 0.35 – – 1.0 –

0.4 0.2 0.1 – 1.5 –

design constraints for the selection of GIV cladding materials are: low effective neutron cross section absorption; superior resistance to fast-neutron radiation damage; outstanding dimensional stability against thermal (500–1000 °C) and irradiation creep (doses up to 200 dpa); corrosion resistance to the working environment, especially to coolants and fissile products; and feasible manufacturing route and assembling. Because of these strict constraints, most materials employed for the fuel cladding and other structural components in current commercial reactors might not be suitable for the use in GIV reactors. Several materials, such as ferritic–martensitic stainless steel (F/M steel), oxide dispersion strengthened (ODS) alloy, nickel-based super alloys (Inconel and Incoloy), refractory metals and ceramic materials (SiC and ZrC), have already been suggested as candidate fuel cladding materials for GIV reactors, see Table 7 [29,66–77]. In fact, it is getting more important to develop and select temperature and radiation resistant materials for GIV reactors in comparison with current GII reactors (see Fig. 8 and Table 7). Radiation-induced changes in material properties are the result of microstructural defects caused by neutron-atom elastic collisions, with the formation of displacement cascades and radiation-induced defects. Usually, the fuel cladding material present sufficient thermally activated diffusion to enable recombination of most of these defects (90–99% of these displaced atoms eventually recombine with vacancies). The remaining defects can be identified as regions where there is either a deficiency of lattice atoms (such as vacancies clusters) or an excess of lattice atoms (such as self interstitial atom clusters and interstitial-type dislocation loops). These defects might produce changes in the mechanical, corrosion and physical properties of the cladding material [29,35,55,66,75]. At low homologous temperatures (0.4) and low radiation doses (0.001–0.1 dpa), clusters of radiation-induced defects act as obstacles to dislocation motion, promoting radiation hardening and reducing the ductility, especially in body centred cubic (BCC) crystals, defining, therefore, a constraint for the lower service temperature. At intermediate homologous temperatures (0.4–0.6) and higher radiation doses (1–10 dpa), radiation-induced segregation and radiation-induced precipitation can lead to localised corrosion or mechanical property degradation, such as grain boundary embrittlement [35,66]. As a matter of fact, irradiation-driven diffusion affects the redistribution of alloy elements, which might change the shape and distribution of precipitates, even promoting the homogenisation of a multi-phased material [29]. Additionally, under the above service

Table 7 Main characteristics of GIV nuclear fission reactor systems [66–75].



Reactor type

Fuel

Coolant

Moderator

Neutron spectrum

Core outlet temperature (°C)

Dose (dpa)

Candidate cladding material

Super critical watercooled reactor (SCWR)

UO2 (thermal) MOX (fast)

Water

Water

Thermal or Fast

 550

10–40

Sodium-cooled fast reactor (SFR) Lead-cooled reactor (LFR)

UPuC/SiC U–Pu–Zr MOX

Liquid Na

–

Fast

 550

90–160

Zr alloys Austenitic stainless steel F/M steels Ni–based superalloys ODS alloys F/M steels ODS alloys

Nitrides, MOX

Liquid Pb alloys

–

Fast

550–800

50–130

Gas-cooled fast reactor (GFR)

(U, Pu)O2 Carbide fuel (U, Pu)

He

–

Fast

 850

50–90

Molten salt reactor (MSR) Very high temperature reactor (VHTR)

Salt

Molten salt

Graphite

Thermal

700–800

100–180

Austenitic stainless steel F/M steels ODS alloys SiC Refractory alloys ODS alloys Refractory alloys SiC No cladding

TRISO

He

Graphite (thermal)

Thermal or Fast

1000

7–30

ZrC coating

UOC

SiC coating

TRISO: Tristructural isotropic fuel, a type of micro fuel particle consisting of a fuel kernel composed of uranium oxide (sometimes uranium carbide) in the centre, coated with four layers of three isotropic materials.

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Fig. 8. GII and GIV design constraints: operating temperature as a function of neutron dose (dpa) for different nuclear reactor systems, indicating a clear trend to higher service temperature and higher neutron dose in GIV reactors. On the right, few candidate materials (SiC, ODS and F/M steels) are listed as a function of the service temperature [29].

conditions, void swelling from vacancy accumulation can create undesirable volumetric expansion, while radiation induced creep and/or anisotropic growth can produce dimensional expansion along directions of high stress and/or specific crystallographic directions [66]. Radiation also enhances creep mechanisms in metals by counteracting the need for high temperatures, so irradiation-produced vacancies provide the means of dislocation climb around radiation-stable precipitates [59,69]. Fig. 9, for instance, brings a summary of various radiation damage mechanisms as a function of the service temperature. Thermal creep usually defines the constraint for maximum service temperature, while radiation embrittlement defines the constraint for minimum service temperature (see Fig. 10) [66,75]. A major focus of research in developing radiation resistant materials is to stabilise the displaced atoms, vacancies, and lattice distortions by designing self-healing micro and nanostructure in order to capture and hold migrating radiationinduced defects. This includes the introduction of special grain boundaries, the control of grain size and texture, and the introduction of radiation-defect sinks, such as trace elements, deformation and dispersed strengthening precipitates [29,55,66,75]. In this sense, the use of computational materials science is expected to provide significant advances for the development of radiation-resistant materials for GIV reactors. Modelling has provided some understanding of the radiation-damage processes, such as primary cascade formation, formation of radiation-induced defects and the dynamics of gliding dislocations in the presence of radiation-induced obstacles. The methodology commonly applied is the multiscale modelling, which is based on the simultaneous modelling of the crystal lattice and microstructure under irradiation, altering dimension (from the atomic to the macroscopic scale) and time (from sub picoseconds to few decades) [73–75]. However, the atomistic processes responsible for the property changes of irradiated materials, see Fig. 11, are not fully understood, requiring a synergy between modelling and experimental work in order to understand the effect of neutron radiation on material performance, allowing the design of radiation-resistant microstructures [75]. As a matter of fact, the combination of high temperature, high fast neutron radiation dose, aggressive environment and longer in-service life is major obstacle for the selection of materials and the viability of some of the GIV reactors, see Table 7 [66–75]. Microstructural changes caused by fast neutron irradiation will affect both the creep and corrosion properties of the cladding material during service, adding more obstacles to the fully assessment of in-service properties of new materials. Comprehensive validation of the performance of these advanced materials in prototypic operating environments will be a

Fig. 9. Diagram showing various radiation damage mechanisms as a function of the service temperature (represented as homologous temperature) [76].

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Fig. 10. Estimated operating temperature window (see black bar) for structural materials in nuclear energy systems (neutron dose between 10 and 50 dpa) [66].

Fig. 11. Schematic diagram of various radiation damage processes and their effect on material performance [75].

key step to obtain their commercial acceptance [66]. The main challenges facing the candidate cladding materials for GIV reactors will be briefly discussed below. 4.1. Zirconium alloys There are significant advantages associated with using a more neutron-transparent cladding material than the other alloys currently envisaged for the SCWR reactors (see Tables 7 and 8 and Fig. 8), but the design constraints for SCWR also focus on the high temperature (creep and corrosion) properties (up to 600 °C) and irradiation resistance [32,68]. According to

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C.R.F. Azevedo / Engineering Failure Analysis 18 (2011) 1943–1962 Table 8 Calculated effective neutron absorption cross section (neutron absorption cross section per unit of yield strength compared to Zr alloys) and maximum service temperature for selected materials [34,66].

*

Material

Maximum service Temperature*(°C)

Effective neutron absorption cross section in relation to Zr alloys

Zirconium alloys FM stainless steels Inconel Austenitic stainless steels ODS alloys Nb–1Zr alloy ZrC SiC Tantalum alloys Molybdenum alloys Tungsten alloys

400 500 500 600 700 800 900 900 1000 1100 1200

1 15 15 15 15 20 0.20 0.10 50 10 35

Based on neutron irradiation between 10 and 50 dpa.

Motta et al. [32], Zr alloys initially did not receive much consideration to SCWR reactors because it was believed that their mechanical strength and corrosion resistance would not be satisfactory at higher service temperature and radiation exposure. The corrosion of Zr alloys usually produces a multi-layered oxide structure, due to a repetitive process of oxide growth followed by a transition to the next layer and the full understanding and control of its passivation is of major importance. The protective behaviour of the oxide layer has been associated with the texture of the oxide microstructure and the presence of interfacial tetragonal Zr3O along with monoclinic Zr2O, but the use of Zr alloys as cladding material still limits the service temperature to less than 400 °C [32,57,58,68]. Efforts, however, are being made to improve the high temperature corrosion resistance of Zr alloys, although the understanding of the corrosion mechanism and the role of alloying elements and microstructure are still uncertain [32,58,61–64]. The development of new Zr alloys for cladding nuclear will have to overcome not only the improvement of the corrosion resistance at higher temperature, but also the requirements of higher creep and radiation resistances [30–32,55,67,68,70–72,76–78].

4.2. Stainless steels Austenitic stainless steels present good creep resistance to higher temperatures coupled with reasonable corrosion/oxidation resistance, being a candidate material for fuel cladding of SWRC and LFR reactors for temperatures up to 600 °C (see Tables 7 and 8 and Figs. 8 and 10). However, radiation damage, such as large amount of void swelling, radiation-induced segregation, He embrittlement, irradiation creep and microstructural instability, remains a major performance-limiting factor even at moderate neutron irradiation doses [52,54,68]. For example, radiation-assisted depletion of Cr from the grain boundaries may induce susceptibility to corrosion in water cooled systems. Additionally, austenitic steels (FCC) are much more prone to radiation void swelling than ferritic (BCC) and martensitic (body-centred tetragonal, BCT) steels. In recent past, one of the main challenges was to increase the radiation resistance of austenitic alloys by designing self-healing microstructures (doping with trace elements, cold deformation and precipitation of dispersed phases) to prevent void swelling [66–69,75,79–82]. So far, however, the challenges for austenitic stainless steel remain unfulfilled and other families of stainless steel seem more likely to meet the strict design constraints for the selection of fuel cladding materials [66–75]. Ferritic/martensitic (F/M) stainless steels, presenting 9–12%Cr, are potential candidates for the cladding material in a number of GIV reactors, such as SFR, SCWR and LFR reactors in temperatures up to 500 °C, see Tables 7 and 8 and Figs. 8 and 10 [66–71,73,83–88]. When compared to austenitic stainless steels, F/M steels feature better thermal properties (higher thermal conductivity and lower thermal expansion); better compatibility to heavy liquid metals, such as Pb used as coolant for LFR; better void swelling and He-embrittlement resistance, even at displacement doses above 150 displacements per atom (dpa) due to its BCC/BCT crystal structure and its tempered martensite microstructure, with high density of irradiation-defect sinks [68,71,73,83]. Additionally, the radiation-induced hardening and embrittlement are negligible when irradiation occurs in the temperatures above 450 °C. However, microstructural instability of F/M steels under irradiation and high temperatures is a major concern because of its deleterious effect on both mechanical and corrosion properties. For instance, radiation-induced segregation is promoted with an increase of neutron irradiation fluence, leading to intergranular segregation and intergranular fracture. Additionally, thermal and irradiation-induced microstructural instabilities, such as the precipitation of brittle phases; the growth of the lath-packet substructures and the M23C6 and (V or Nb)C precipitates; and the formation of grain boundary cavities degrade the mechanical properties of F/M steels, limiting its the maximum service temperature [68,71,83–87]. As a matter of fact, in the extremely demanding design conditions for commercial fast reactors, both austenitic and ferritic–martensitic steels seem to have severe limitations, the former due to swelling and the latter due to insufficient creep strength. One approach to improve high-temperature strength in F/M steels is by introducing a high number density of stable, fine-scale, precipitates that pin the motion of dislocations. Considerable progress has been made on the development of oxide dispersion strengthened (ODS) F/M steels and ferritic alloys [66–71,73,86].

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4.3. Ni-base alloys The austenitic materials that are most resistant to swelling are high nickel precipitation-hardening alloys. Ni alloys have good creep rupture properties and high temperature strength, being considered as a cladding material for SCWR, especially Inconel 690, 625 and 718 (see Tables 7 and 8 and Figs. 8 and 10). The main problem with Ni-based alloys is their low radiation resistance, due to its FCC matrix, which promotes radiation embrittlement, swelling and phase instability (radiation induced precipitation of a large amount Ni3Ti(Al) phase). Compared to stainless steels, there is less experience with radiation effects on the Ni-base alloys, but available information indicates that these alloys suffer considerable levels of swelling and large reduction in ductility during neutron irradiation around 15 dpa and temperatures between 400 and 600 °C. The loss in ductility was associated with various grain boundary phenomena, such as: radiation-enhanced precipitation of intermetallic and growth of thermally-precipitated carbide phases, such as M23C6 and M6C; radiation-induced segregation; and precipitation of He bubbles. It seems possible, however, to optimise the microstructure of Ni alloys to fight the deleterious effects of neutron irradiation, but the lower driving force for the development of ODS alloys based on a radiation-resistant BCC matrix might play an important role on the potential use of Ni alloys as a cladding material [68,70,71,73,88–90]. 4.4. Refractory alloys Refractory metals are candidate materials for fuel cladding of LFR and GFR reactors, with service temperatures up to 850 °C (see Tables 7 and 8 and Figs. 8 and 10) [66,68,70]. These alloys show melting temperatures above 1850 °C, but most of them are not candidate materials for fuel cladding due to their high values of neutron absorption cross section (see Fig. 12 and Table 8). According to Murty and Charit [68], although refractory metals possess good creep resistance and swelling resistance up to high burn-ups, they have poor oxidation resistance coupled with low temperature radiation embrittlement and major fabrication difficulties. Refractory metals, however, have been considered for use in NASA’s nuclear reactors, which operates above 800 °C [91]. For instance, niobium alloys are suitable for applications up to 800 °C and tungsten alloys up to 1200 °C [66,91]. Table 9 shows various engineering aspects of some refractory alloys on a 10-point scale (1 being the worst and 10 being the best) and although refractory metals generally possess good creep, oxidation and swelling resistances, they show average radiation resistance, high cost and fabrication problems [69]. For instance, some BCC refractory metals exhibit ductile-to-brittle transition temperatures (DBTT) above room temperature, requiring particular care in fabrication and handling (W is brittle around 330 °C). Nb alloys offer the advantages of fabrication, high ductility and melting

Fig. 12. Refractory metals: neutron absorption cross section versus melting temperature for pure elements, indicating the trade-off between neutron transparency and service temperature [34].

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Table 9 Summary of various engineering aspects of some refractory alloys [34,69]. Aspect

Nb–1Zr

Ta-10 W

Mo–0.5Ti–0.1Zr

W-Re

Fabricability Weldability Creep strength Oxidation resistance at 500 °C Alkali metal compatibility Radiation effects Cost (£/kg)

8 7 6 Limited use 8 6 119–196

7 7 8 Acceptable 9 6 326–521

4 4 8 Acceptable 9 5 16–17.5

3 3 8 Excellent 9 4 289–318

temperature, and low DBTT and neutron absorption cross-section (Figs. 10 and 12). However, the strength of Nb alloys is relatively low and their resistance to oxidation at elevated temperature is extremely poor, requiring extra care during fabrication and handling. Nevertheless, Nb–1%Zr–0.1%C alloy is the current choice for uranium nitrate fuel cladding material for space nuclear power systems and to avoid its potential attack by fission products a thin layer of rhenium on the inner surface of the cladding acts as a protective barrier [91,92]. Once again it is believed that ODS alloys might be a more feasible choice for fuel cladding material, especially for commercial reactors, such as LFR and GFR [66,68,71–74,91].

4.5. ODS alloys Priority has been given to the development and study of Fe–Cr oxide dispersion strengthened (ODS) alloys for structural in-core applications of GIV reactors within the European GETMAT (Gen IV and transmutation materials) project. ODS alloys might show good properties for high temperature (800 °C) and high burn-up applications, being a candidate material for the fuel cladding of SFR, SCWR, LCR and GCR reactors (see Tables 7 and 8 and Figs. 8 and 10) [66–74,91]. An important approach for designing the microstructure of this radiation-resistant alloys is based on the introduction of a high, uniform density of thermal and irradiation-stable nanoscale particles, using particles of titania (Ti2O3) and yttria (Y2O3), distributed in a ferritic or tempered martensitic matrix, with chemical composition typical of F/M stainless steels or Fe–Cr Incoloy [66,91]. These nano-particles are thought to act simultaneously as obstacles to dislocation motion, providing high creep strength, and as sinks for the radiation-induced point defects, providing good radiation damage resistance [66,68,71,91,93]. Manufacturing of ODS alloys involves mechanical alloying (MA) of metallic and ultra fine oxide powders, through the repeated fracturing and welding of a mixture of powder particles in a highly energetic ball mill, which results in the formation of microcomposite particles with a refined and highly deformed microstructure. The MA is followed by the consolidation of the microcomposite powder by thermo-mechanical processing, such as hot extrusion, rolling or hot isostatic pressing [91,93,95,96]. The final product features an oriented microstructure, presenting highly-directional properties, with the creep strength in the transverse or hoop direction being approximately 50% of that in the axial direction [93,95]. In this sense, manufacturing ODS alloys products with equiaxed grains seems to be a crucial challenge for its use as a fuel cladding material [93]. Additionally, the manufacturing of long fuel rods also challenges the processes of consolidation and joining [91,94–96]. According to Yvon and Carré [71] the main in-service issues for ODS alloys under fast neutron radiation are the effect of phase instability under fast neutron radiation on the mechanical properties at lower temperatures; and the effect of the stability of the oxide dispersion and the effect of intermetallic phase precipitation on the toughness at higher temperatures (creep properties). Furthermore, the use of ODS alloys for nuclear application is nearly unexplored and there is some concern on the deleterious effects of high doses of neutron irradiation (up to 200 dpa) on the macro and microstructure of ODS alloys [73]. Tensile specimens of ODS martensitic 9Cr-ODS and 12Cr-ODS ferritic steels have been irradiated in the Japanese experimental fast reactor up to 15 dpa and at a temperature range of 400–550 °C and the results showed adequate strength and ductility, with microstructural stability of the size of oxide particles and the density of dislocations [93]. Higher neutron exposure results, however, are needed to assess the radiation resistance of ODS alloys [93,97]. Finally, the selection of cladding material for LFR and SFR reactors imposes an additional challenge to the cladding material and the compatibility of the ODS alloys in these environments should be further investigated [98,99]. For instance the evaluation of Na environment effects on the corrosion and mechanical properties was carried out and the results showed excellent Na-resistance up to 700 °C in stagnant conditions. In flowing conditions, weight increase and microstructural instability (a transforming into c) was observed at temperatures over 650 °C, but the fine Y2O3 oxide particles remained stable for both experimental conditions [98]. Schroer et al. [99] showed that 9Cr–2 W ODS steel presented a promising long-term performance in oxygen-containing flowing LBE at 550 °C for exposure times up to 10,000 and 20,000 h. They stated that with increasing the concentration of dissolved oxygen in liquid Pb or LBE, the mechanism of material degradation changes from corrosion to oxidation, accompanied by the formation of an oxide scale on the material surface, which can minimise its dissolution. For liquid Pb and LBE, the upper limit for oxygen enrichment is determined by the oxidation rate of the material, which increases with increasing oxygen content of the liquid metal [99]. Corrosion resistance of ODS ferritic steels containing 3.5 wt% Al and 14–17 wt% Cr was investigated at 550 and 650 °C in a stagnant LBE containing 106 and 108 wt% oxygen. All the ODS steels containing about 3.5 wt% Al showed good corrosion resistance after a 5000 h exposure [100]. Finally, according to Allen and Crawford [101] structural stainless steels and ODS alloys developed for the fuel cladding of SFR

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reactor have showed inferior chemical compatibility with lead and lead-bismuth environments, especially at higher temperatures, while ceramic material has currently proven better corrosion and radiation resistance under higher temperatures.

4.6. Ceramic materials Ceramic materials, such as SiC and ZrC, are considered as potential cladding material for LFR, VHTR and GFR reactors, which will be exposed to temperatures above 850 °C and fast-neutron radiation [71]. The main challenge for the fuel cladding material of GFR reactor includes high resistance to fast-neutron damage and high temperature exposure (up to 1600 °C in accident situations). For the LFR reactor, there is an additional concern about the chemical compatibility of the cladding material with the Pb or Pb–Bi coolant and the mixed nitride fuel. In both reactors, SiC is a candidate material for the fuel cladding. For VHTR reactor, ZrC and SiC have been considered as a coating material for the TRISO fuel (tristructural isotropic fuel, a micro fuel particle consisting of a fuel kernel composed of the fuel coated with four layers of three isotropic materials) to provide the mechanical stability to the fuel and act as the main diffusion barrier to the release of fission products [70].

4.6.1. SiC SiC and SiC/SiC are being considered as the primary candidate materials for fuel cladding in a GFR, LFR and VHTR reactor (see Tables 7 and 8 and Figs. 8 and 10). SiC cladding has a number of potential advantages, when compared to other candidate materials, such as better high-temperature corrosion properties (leading to higher thermodynamic efficiency), best safety and waste disposal restrictions, greater high-temperature strength and lower effective neutron absorption cross section. Additionally, SiC cladding is chemically compatible with mixed carbide fuel and its possible incompatibility with nitride fuel can be prevented through the use of a cladding liner [70,102,103]. SiC composites have demonstrated good irradiation behaviour and should be capable of maintaining mechanical properties at radiation damage levels beyond 50 dpa at temperatures around 1000 °C [104]. One of the main challenges for the use of SiC as a fuel cladding material is its brittle behaviour, which can promote in-service failure due to pellet-cladding mechanical interaction. SiC/SiC composites can provide additional strength and flexibility for SiC structures. Additionally, SiC is vulnerable to oxidation under certain conditions and its susceptibility to stress corrosion cracking should be further investigated [70,102,103]. Finally, cost, joining and chemical stability with the nuclear fuel and coolant are the major issues confronting the development of SiC fuel cladding [66,70,104]. Furthermore, SiC gradually loses strength due to neutron irradiation, affecting its mechanical integrity during service. Numerous issues need further R&D activities in order to investigate the use of SiC as in-core structural material of GIV reactors [66].

4.6.2. ZrC ZrC is one of the most promising coating materials for TRISO particle fuels in VHTR (see Tables 7 and 8 and Fig. 8) because of its neutronic and thermal properties, good chemical stability and high retention of fission products. A key part in developing the use of ZrC is to ensure adequate mechanical properties and dimensional stability in response to fast-neutron radiation and feasible fabrication route for the coating process [70,105,106]. The main disadvantages of ZrC when compared to SiC are the lack of data on high burn-up long-duration irradiation exposure tests; its lower mechanical strength; its lower oxidation resistance; its higher resonance neutron absorption; its larger thermal expansion coefficient ( 30% higher than SiC); and the complex fabrication route for ZrC layer deposition [70,107].

5. Conclusions  The multiple design constraints for the selection of nuclear fuel cladding material includes the service temperature, the creep resistance, the mechanical strength, the toughness, the neutron radiation resistance (thermal or fast), the neutron absorption cross section, the thermal properties and the chemical compatibility with fissile products and coolant, moderator and fuel materials.  The values of the effective neutron absorption cross section indicated that metallic and ceramic multi-component systems, combining Be, C, Mg, Zr, O, and Si, might be considered as potential new materials for fuel cladding in terms of neutron economy.  There is a clear trend to increase the efficiency of nuclear power plants by prioritising the following design constraints: higher service temperatures; longer in-service life; and resistance to more corrosive environments and to higher levels of fast neutron radiation. New materials are under consideration for application in GIV reactors, such as oxide dispersion strengthened (ODS) steel, refractory alloys and ceramic materials (SiC and ZrC). However, further investigation of the processing routes, corrosion properties and the effects of high (thermal and fast) neutron exposure on mechanical properties are required to assess the potential of these materials as in-core components of GIV reactors.

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Acknowledgements The present work was carried by the author as an Academic Visitor at the Department of Materials, Imperial College during the year of 2010. The author would like to thank Dr.-Ing. A.H. Feller, Mr. D. Palmer, Prof. A.T. Motta (Pen State University), Prof. I.G.S. Falleiros (Universidade de São Paulo) and Prof. T. Lindley (Imperial College) for their vital support during the execution of the present work, which was sponsored by USP (Universidade de São Paulo) and CNPq (National Council for Scientific and Technological Development, Ministry of Science and Technology of Brazil, Process: 201325/2009-8). Finally, the author would like to express gratitude to his colleagues at the Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo for their valuable collaboration during his sabbatical leave and to the unconditional guidance of Prof. T. Cescon (Instituto de Pesquisas Tecnológicas do Estado de São Paulo).

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