IRRADIATION EFFECTS ON MATERIAL PROPERTIES OF STEELS USED
IN NUCLEAR REACTORS: A LITERATURE REVIEW
Nilgün GERÇEKER, ilksen Hilâl DÂRA
Turkish Atomic Energy Authority, Technology Department
ABSTRACT
The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials.
1. INTRODUCTION
The primary components of a nuclear power reactor can be classified in seven main categories as: nuclear fuel, structure, moderator-blanket-reflector, control element, coolant, shields and safety systems [1, 9], The main materials used in each category are listed in Table 1. The main concern of this study is the structure category in which fuel cladding (can, jacket or tube), pressure vessel, fuel coolant channels, core support plates (or grids), coolant piping system, control element mechanism can be listed. Among the metals used for structural mate rials beryllium, magnesium, aluminum and zirconium are suitable for use in thermal (research and power) reactors due to their low thermal neutron absorption cross section and austenitic stainless steels (or mild carbon steels) and nickel alloys are suitable for use in fast (research, power and breeder) reactors due to their low fast neutron absorption cross section but high thermal neutron absorption cross section. In thermal reactors however, stainless steel and carbon steels may also be used in pressure vessel and leak-tight piping materials where the neutron absorption cross-section is not important. Stainless steels have excellent mechanical properties, corrosion and oxidation resistance at elevated temperatures, which render them as candidate for use at high temperatures in
Table 1. Main materials of primary components of nuclear reactor [1]
containers and other nuclear applications. Another advantage with the utilization of stainless steel is its reasonable cost and availability.
Primary Component Main Materials
Nuclear Fuel Uranium
Structure Zirconium alloys, mild steel, stainless steel, nickel alloys
Moderator and reflector
(thermal reactor) Graphite, heavy and light water, beryllium
Blanket and reflector
(fast reactor) Depleted uranium, thorium, beryllium, graphite
Control Element Boron carbide, cadmium, hafnium, boric acid, burnable
absorbers
Coolant He, CO2, light water, heavy water, liquid metals (NaK, Na)
Shields Light, medium and heavy elements or compounds
Safety Systems Pressure suppression system, emergency core cooling
systems, instrument monitoring system
Among a large variety of AISI types, nominal compositions of the most widely used austenitic stainless steels together with the possible application areas are given in Table 2. In spite of their excellent material properties, there are still very important problems faced for austenitic stainless steel such as: (1) stress corrosion cracking at the inner surface of fuel cladding and pressure vessel (2) depending on the coolant, susceptibility to induce intergranular corrosion after coating process (3) particularly in high ferrite weldments, the probable formation of a brittle sigma phase (4) the attack of fission products and chemical compounds to the inner cladding surface to initiate corrosion cracking (5) the possible removal of sensitized protection films because of the rapidly flowing corrosive liquids and a consequent accelerated local attack (6) thermal stress fatigue cracking due to reactor kinetics.
Table 2. Composition and applications of stainless steels used in reactors AISI
Type
Alloying Element
(wt%) Application
C Cr Ni Mo Nb Fe Base and weld materials for LMFBR pressure vessels 304 0.08 18-20 8-11 - -The rest 304L 0.03 18-20 8-11 - -309 S Nb 0.08 22-26 12-15 - 0.64 (min)
Reactor vessel and core structural material in
LMFBR
316 0.1 16-18 10-14 2-3 - Cladding and welding material for LMFBR fuel elements 316L 0.03 16-18 10-14 2-3 -347 0.08 17-19 9-12 0.81 (min)
Reactor vessel and core structural material in
LMFBR
Irradiation effects on material properties -between all other inherent effects due to the reactor environment- is one of the major problems to be considered as far as the safe operation and appropriate maintenance of a power reactor is concerned. The irradiation effect (or damage) produced on a nuclear reactor material, fuel or structural material in particular, depends mainly on the neutron flux, neutron energy or energy spectrum and irradiation time and temperature. Beyond a threshold-integrated neutron flux, an irradiation effect on nuclear, physical, thermal, mechanical and some other material properties are inevitable [1,8]. The principal irradiation effects on the structural materials are considered as irradiation swelling, irradiation creep, and irradiation embrittlement however, irradiation effects on physical, thermal and mechanical properties of stainless steels will all be reviewed shortly as well as swelling, creep and corrosion problems of stainless steels associated with irradiation effects in the following sections of this paper.
2. IRRADIATION EFFECTS
2.1. Neutron Radiation Damage Mechanisms in Steels
When a material is exposed to irradiation any property of the material may change including physical dimensions as well as mechanical, thermal, optical, electrical and some other properties. The fundamental phenomenon responsible for the change is the existing crystal and defect structure, which is probably deconstructed and reconstructed atom-by-atom basis during
may be placed from its lattice sites tens or hundreds of times and the dose of radiation is measured with displacement per atom (dpa) .The ultimate property changes occur due to the combined effect of temperature, dose rate, dose and local alloy composition [2].
In Figure 1 the creation of point defects by displacements and the various paths for their disposition are illustrated to provide the reader a framework developed to understand the radiation effects in structural materials. The left path comprises the defects, which may contribute to the buildup of microstructure and to property changes and the bottom of this path (swelling, creep and embrittlement) are the effects of radiation that have occupied the major share of the studies performed up to date. The right path of the flowchart illustrates the self interstitials and vacancies that may combine at various locations producing no long term property changes.
2.2. Irradiation Effect on Mechanical Properties
Structural materials constitute the most important components of a nuclear reactor, therefore mechanical properties associated with these structures can highly influence the reactor design, operation, performance and safety. The irradiation effect on change in strength, ductility, hardness, ductile-to-brittle transition temperature, creep and fatigue will be discussed in separate subsections below.
Displac ements
t
In-Cascade Recombination and Clustering
t
Thermal Diffusion and Cluster Process
t
t
Swelling Creep Embrittlement
No Change in Properties
2.2.1. Mechanical Strength
The mechanical strength is generally expressed as the stress-strain characteristics of a material. Irradiation improves the ultimate tensile strength of the structural steel whereas decreasing the toughness and ductility. However, it is not possible to draw a general conclusion out of stress- strain diagram because the irradiation temperature, is the determining factor in assessing the irradiation effect on the mechanical strength. Depending on the neutron flux and the irradiation temperature recovery may occur simultaneously which will finally relieve some of the residual stresses and may have an annealing effect on the material concerned. The increased temperature and neutron fluence helps to recover some of the mechanical strength and toughness lost due to irradiation at lower temperatures [1]. Extravagant variants may also occur, finally developing a yield point in materials that previously lack yielding and changing sensitivity to factors such as strain rate [7].
2.2.2. Ductility, Hardness and Embrittlement
Ductility is a mechanical property responsible for a structural material to undergo permanent deformation before failure therefore is important in design and safety considerations. Ductility is expressed as the percentage elongation or percent reduction in area.
Irradiation hardening and embrittlement lead to the decrease in ductility. The embrittlement of steels exposed to radiation is straight or indirectly caused by the displacements of atoms from their original positions due to the collisions by energetic particles. The mechanism may be simulated by the interaction of atoms similar in a billiard fashion. After the collision, a displacement cascade occurs consisting of several vacancies in the middle surrounded by a cloud of interstitial atoms. Most of the defects recombine and take part in the radiation induced damage process together with the small amount of remaining few freely migrating point defects. Therefore it is evident that the radiation damage of a reactor pressure vessel for example is proportional to the number of neutrons hitting the material with high enough energy to induce atom displacement(s). However, another important point to be emphasized is that the smaller displacement cascades provide a weaker opportunity for recombination thus it can be stated that moderated neutrons can be relatively more damaging [1, 2, 8].
The combined effect of embrittlement and decreased ductility may finally lead to the fracture failure of reactor primary components. Mechanical and strength and ductility constitutes the toughness of a structural material and a change in the toughness of fuel cladding and pressure vessel particularly causes various problems on the operation, performance, economics and safety of the reactor.
2.2.3. Ductile-Brittle Transition Temperature (DBTT)
One of the major problems faced with steels used in reactor applications, is the irradiation induced embrittlement. Hardening of steel due to irradiation reduces the ductility and increases the ductile-brittle transition temperature in effect. After irradiation, DBTT increases above the
room temperature and brittle fracture can occur instead of ductile fracture. Shift of DBTT is a critical parameter for safety of particularly pressure vessels and can be attributed to vacancies, interstitials, dislocations which are formed by displacement of atoms from lattice positions to interstitial positions and impurity atoms [4,8].
2.2.4. Creep
Irradiation creep is slow, plastic and continuous deformation of a solid material under constant load generally at elevated temperatures in a radiation environment. Mechanisms of irradiation creep are directly the net climb of particularly oriented dislocations, or indirectly from any climb that triggers glide in response to the applied stress [3]. These are generally treated as steady-state processes. Recent studies are concentrated on the transient processes of cascade- induced creep, startup-induced creep by transient interstitial absorption and glide-induced creep by transient vacancy absorption. Such processes may dominate irradiation creep for low temperatures, low doses, and small dislocation pinning obstacles under appropriate circumstances. Recently, irradiation creep was studied in a wide temperature range extending to near room temperature, where measurable deformation might have not been expected. This phenomenon was investigated further in terms of mechanisms of defect migration and absorption. In the temperature regime near room temperature and below, vacancies created by irradiation are effectively frozen in place while interstitials are free to migrate because a very small fraction (but unbalanced by vacancy absorption) of the interstitials is absorbed at dislocations rather than recombining with the population of immobile vacancies, therefore significant creep rates result. At high temperatures both vacancies and interstitials are absorbed but are partitioned differently to dislocations of different orientation, or are partitioned differently to dislocations, cavities, and other types of sinks.
In general, it is difficult to determine creep rate of the components of the nuclear reactor. Therefore, some sample inspection and surveillance programs are carried out in nuclear power plants for the prediction of creep rate of the components.
2.2.5. Fatigue
Thermal and induced-load cycling is an inherent character of nuclear reactor kinetics. The repeated cycling within a certain strain or stress range may develop fatigue cracking and failure in the cyclic life of a reactor material. Irradiation shortens the fatigue lives of the irradiated steels. Temperature associated with the irradiation process is a determining factor on the final effect on fatigue life of a structural steel and as the temperature is increased the effect of temperature is slightly decreased. The mechanism of the decrease in fatigue life may be attributed to the slip bands, which penetrate the irradiated regions through the cleared channels to serve as fatigue crack initiation sites [5].
2.3. Irradiation Swelling
Irradiation swelling is defined as a volumetric instability caused by 4He in structural materials
materials may also involve an irradiation effect on the combined physical (density change), thermal (conductivity change) and mechanical properties. The increase in volume is accompanied by a decrease in density of the structural material. This can create a serious problem especially for the moving parts in the reactor.
2.4. Irradiation Effect on Corrosion
Physical, mechanical, and thermal properties of nuclear reactor materials determine the corrosion rate and behavior. Stainless steels used in reactor applications are passive in natural fresh water, mine waters, boiler condensate and steam at elevated temperatures. Experiments showed that corrosion resistance of these alloys is mainly due to the presence of thin oxide or hydrate film stabilized by chromium. Passivation is continuos under certain environments and conditions [6,8].
Mechanisms of corrosion in aqueous systems and liquid metal systems are quite different from each other and with this feature irradiation effects on corrosion rate are not similar to one another for different cases. Generally, in aqueous systems (LWR, HWR, etc.) corrosion rate increases due to irradiation. In liquid metal systems (LMFBR, CTR) only slight enhancement is observed in the corrosion rate.
In an aqueous media, irradiation effect is accelerated by chemical reactions and this brings about an increase in corrosive activity. There are three mechanisms to be concerned in the enhancement of corrosion activity:
1. Hydrogen, oxygen, hydroxyl ions, and hydrogen peroxide are formed as a result of radyolytic decomposition of water and this results in an increase of corrosive activity. 1. 2.Neutrons disturb the thin protective layer on the surface of the steel, resulting in
pitting of the metal surface,
2. 3.Change in mechanical, physical, thermal properties of the meterial brings about changes in the corrosion rate.
In liquid metal systems the basic mechanisms are:
1. Transfer of atoms, particularly impurity atoms, of the solid into liquid under activity gradient arising when metallic or nonmetallic constituents exhibit different activity in two solid alloys contacting the liquid to form a solid-liquid couple,
2. Solubility of solid in liquid changes with temperature resulting in dissolution corrosion, 3. Penetration of liquid through grain boundaries of solid.
In these mechanisms, many investigators believe that embrittlement of stainless steels are mainly due to intergranular attack. Stainless steel exhibit good corrosion resistance to liquid sodium at temperatures below 649°C provided the oxygen content exceeds 0.02 %. t. When the oxygen content exceeds this value, corrosion to some extent may be observed.
Especially, 300 series stainless steels have excellent corrosion resistance at high temperatures therefore they are widely used in nuclear reactor applications. Sensitized 300 series steels are not susceptible to intergranular attack in water at 500 °C. However, steels case hardened by nitriding or malconizing, do not have good corrosion resistance under the same conditions. Corrosion rate in supercritical water depends strongly on the temperature. Corrosion rate increases with temperature at 538 °C and 732 °C higher corrosion rates are observed.
2.5. Irradiation Effect on Thermal Properties
Thermal conductivity, thermal expansion coefficients and specific heat are the thermal properties of the materials that have to be taken into account in the reactor design. Change in these properties in metals, is the result of presence of lattice defects. Thermal conductivity and thermal expansion coefficient are quite insensitive to lattice distortions. Very low specific heat is likely to be changed in small amount by irradiation [6].
2.6. Irradiation Effect on Physical Properties
Density, electric resistivity, magnetic susceptibility and Hall effect are the basic physical properties of the nuclear fission and fusion materials. Change in electric resistivity, magnetic susceptibility and Hall effect due to irradiation are to be considered for materials in the presence of electromagnetic fields. In general, electrical resistivity increases with the integrated flux while magnetic susceptivity and Hall effect decreases with irradiation.
Radiation induced resistivity increases are primarily the result of presence interstitials and vacancies. These lattice defects produce lattice distortions which cause conduction electrons to be scattered more frequently than undistorted crystal. Increase in resistivity is independent from the temperature and strongly dependent on the number of defects.
Many studies are carried out for determination of change in electrical resistivity. If a proposed material has a resistivity increase when irradiated at temperatures common in practice, then it may be evaluated as the as the indication of changes in other properties of the material therefore further investigations should be carried out to prevent detrimental effects of irradiation in structural materials.
4.CONCLUSION
Irradiation effects on material properties of structural components constitute an extensive research area, which is of critical importance for the safe operation of nuclear power plants as well as appropriate maintenance. Steels are widely used materials in structural components of a nuclear reactor therefore a review of the irradiation effects on material properties of steels is given in the present contribution to outline a framework about the origins and mechanisms of radiation effects in materials.
The irradiation effect produced on a nuclear reactor material (fuel or structural component) depends mainly on the neutron flux, neutron energy or energy spectrum and irradiation time and temperature as well as the existing crystal and defect structure, which will be deconstructed and reconstructed in the atomic scale during irradiation process. Irradiation swelling, irradiation
creep and irradiation embrittlement are the three dominating effects to be considered, however irradiation effects on physical, thermal and mechanical properties should in any case be monitored with special care for the safe operation and maintenance of nuclear reactors.
Materials research related to nuclear technology seems to be underestimated in Turkey therefore a thorough investigation study on the component of nuclear reactors including -but not limited to- steels and the irradiation effects on the material properties would be a valuable study to demonstrate the deficiencies of our infrastructure in nuclear metallurgy. It should be emphasized that for nuclear technology, nuclear metallurgy is a very important subject to be considered as the safe operation of a nuclear power plant is only possible with properly functioning structural components.
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4. R. L. Klueh, D.J. Alexander, “Neutron Irradiation Effects on Ductile to Brittle Transition of Ferritic/Martensitic Steels”, Report, Department of Energy, 1997
5. R. Tulluri and D.J. Morrison, “ Effects of Ion Irradiation on Fatigue of Fe-12Cr-20Mn Stainless Steel for Fusion Reactor Applications”, Department of Mechanical and Aeronautical Engineering
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8. Handbook of Stainless Steels, Donald Peckner, I.M. Bernstein, Mc Graw-Hill Company,1977
9. Handbook of Engineering Materials, Douglas F. Miner, John B. Seastone, New York John Wiley and Sons Inc.,1955
