The fundamental elastic properties of metals are determined by the behavior of dislocations - which are systematic distortions introduced into the crystalline lattice. When the material is under stress, these dislocations can move. The ease with which they move, or more generally, by whether and if they can move, determines the elasticity (or plasticity) of the metal. The dislocations can be impeded in their movement by irregularities in the crystal - in the form of grain boundaries, or in the form of impurities - which can act as 'pinning centers' or 'sinks' for the dislocations.
The size and orientation of the grains within the metal relative to the dislocations, determines whether, and how easily, the dislocations can be pinned at the grain boundaries. As the grain size decreases, the dislocations are more easily pinned, and this increases the yield strength of the metal (that is, it delays the onset of plasticity). However, below a grain size of about 100 nm, a reversal of this rule has been observed - with the decline in yield strength flattening out, or the yield strength even declining with decreasing grain size.
The idea that impurities within a crystal can act as pinning centers for dislocations has been intuitively grasped for centuries, as well as the idea that deliberate introduction of misaligned dislocations and defects in a metal through heat treatment or cold rolling can lead to an overall strengthening of the metal. For example, the strength of steel itself arises from the incorporation of carbon atoms in interstices within the iron lattice, which act as pinning centers for dislocations. A similar idea underlies the concept of alloying for increased strength.
Advances in metallurgy led to heat treatments of alloys, followed by differential quenching. Deeper investigations revealed that such treatments led to the formation of a precipitate of the alloyed species, with both the alloyed species and the matrix existing in a metastable supersaturated solution. When the quenching is followed by moderate heating, the precipitate finely diffuses into the matrix, forming a dense network of pinning centers that provides very high tensile strength for the alloy. This technique became known as precipitation hardening.
The idea of deliberately introducing as fine a powder of the pinning species as possible into the steel matrix was then taken up, and eventually led to the development of steels with nanoscale powders of transition metal oxides (such as those of Yttrium or Chromium) dispersed through the matrix. These are known as Oxide Dispersion Strengthened Steels (ODSS) which, in addition to displaying high yield strength, have also displayed low irradiation and thermal creep, as well as high corrosion resistance. This is why they are being intensively investigated as possible structural and cladding materials for next generation nuclear reactors.
In my presentation Multi-scale Modeling of Radiation-Induced-Dislocation Pinning in Oxide-Dispersion Strengthened Steels (ODSS)- Part 2 , I discuss these desirable properties in detail, as well as the scientific questions that arise given the existence of these properties. For example, the precise effect that the specific alloyed species has on the properties is not known. Similarly, the role played by the metal oxide dispersion in strengthening corrosion resistance is not known, nor the details of how the dislocation pinning actually occurs, or whether there are any other novel effects, such as spin interactions, which might conceivably impact the properties of these steels. Such understanding is not only scientifically motivated, but is also necessary to develop a predictive understanding of the properties of the steels that may then be rolled into probabilistic safety analyses of the reactors themselves.
Multiscale Materials Modeling, which I have discussed in another post, naturally lends itself to addressing these issues, and is the focus of my current work. At present, the behaviour of reactor materials is modeled by phenomenological methods involving approximate relationships and observed correlations, which may not extrapolate into the operating conditions (and desired design lifetimes) of the new generation (Gen IV) reactors. The research I am undertaking is intended to fill an important and critical knowledge gap in the path toward these reactors.
It is important to realize the compelling criticality of Materials Issues in the path toward the Nuclear Renaissance - no matter which path is taken - whether a new coolant is introduced or whether new fuel cycles are introduced - the stress on the structural and cladding materials for the new generation of reactors changes both qualitatively and quantitatively. New materials will be needed that can reliably withstand this stress, and their behaviour needs to be understood and modeled, before such materials can be qualified for use in these reactors.
The Materials Research Society Fall Meeting in Boston next month is devoting an entire session solely to a Discussion of Materials Science Solutions to Impediments to the Nuclear Renaissance, in addition to a Symposium on Materials for Future Fission and Fusion Reactors, which includes several sessions both on Multiscale Modeling and on Oxide Dispersion Strengthened Steels.
Tuesday, November 11, 2008
Multiscale Modeling for Nuclear Reactor Materials
For a variety of phenomena, it is true that physical interactions between constituents at a variety of length and time scales, starting from the smallest and cascading upward to the largest relevant scale - contribute to the macroscopic behaviour. While a strictly reductionist viewpoint would be misleading, it is nevertheless true that these interactions exist in a hierarchy of length scales, with the lower cutoff at the atomic scale and the upper cutoff at the 'continuum limit' in the millimeter length scale. This is true in particular of reactor materials that are subject to neutron irradiation (they may also be subjected to alpha or gamma radiation).
A neutron flux, for example, will impact an atom, and attempt to dislodge it from its lattice site, in what is called a 'primary knock-on' event. If the dislodgement is successful, the atom becomes an interstitial, and a vacancy is created at the original location. Such a primary knock-on event, which creates that point defect, is estimated to take as little as one-hundredth of a nanosecond (10^-11 seconds), while the atom is displaced over a length scale of the order of a nanometer(the lattice parameter of most metals is just under 1 nm).
The vacancy and interstitial atom function as a bound pair, known as the Frenkel pair. Many of the plasticity properties of the metal are determined by the behaviour of the population of Frenkel pairs that are created during the course of the irradiation. Point defects and Frenkel pairs can form clusters, and clusters grow by aggregation, forming voids where vacancies within them have accumulated. The formation of voids leads to swelling, which is an isotropic volume expansion of the metal that can be, for significant radiation doses, upto several tens of percent in magnitude. This can cause a considerable change in the linear dimensions as well (this scales as the cube root of the swelling percent).
Another phenomenon that can occur, depending on the mechanics of defect-vacancy formation, is growth, where volume remains constant but there is a change in shape of the metallic object. Usually, both swelling and growth occur, at different locations within the irradiated material. In addition, a phenomenon known as irradiation creep can occur, which is the slow deformation of the metal under constant mechanical and radiative stress.
It is interactions like these, and phenomena that occur when they aggregate over successively longer length (and time) scales within the reactor material that eventually have observable consequences on components that have dimensions of the order of meters, i.e., on the macroscopic scale. These consequences include onset of plasticity, fracture, or embrittlement.
Phenomena such as defect clustering and diffusion, formation of dislocations (a dislocation being formed whenever a regular pattern of defects significantly distorts the crystalline order), formation of separate phases and radiation-induced segregation also occur in the irradiated metal, each of which, depending on conditions, can contribute to the onset of plasticity or the start of fracture in a metal.
Zipping up a description starting from the lowest relevant length (and time) scale, up to the longest relevant length (and time) scale, using the most appropriate physical description at each scale - under conditions of high thermomechanical and radiative stress - is the challenge of Multiscale Materials Modeling under Irradiation (MMM-I). A certain amount of coarse graining is inevitable when this is attempted, and both sequential and parallel approaches are possible, as well as hybrid approaches that utilize elements of both methods.
My presentation Multi-scale Modeling of Radiation-Induced-Dislocation Pinning in Oxide-Dispersion Strengthened Steels (ODSS)- Part 1 and Multi-scale Modeling of Radiation-Induced-Dislocation Pinning in Oxide-Dispersion Strengthened Steels (ODSS)- Part 2 describe these ideas and my work in progress, in more detail, in the context of oxide dispersion strengthened steels, which have been suggested as possible reactor materials for new generation reactors in view of their superior creep strength under both radiative and thermomechanical stress, as well as corrosion resistance. This is the focus of my current work.
The Fall 2008 Meeting of the Materials Research Society in Boston has Symposium W on Computational Materials Design by Multiscale Modeling.
A neutron flux, for example, will impact an atom, and attempt to dislodge it from its lattice site, in what is called a 'primary knock-on' event. If the dislodgement is successful, the atom becomes an interstitial, and a vacancy is created at the original location. Such a primary knock-on event, which creates that point defect, is estimated to take as little as one-hundredth of a nanosecond (10^-11 seconds), while the atom is displaced over a length scale of the order of a nanometer(the lattice parameter of most metals is just under 1 nm).
The vacancy and interstitial atom function as a bound pair, known as the Frenkel pair. Many of the plasticity properties of the metal are determined by the behaviour of the population of Frenkel pairs that are created during the course of the irradiation. Point defects and Frenkel pairs can form clusters, and clusters grow by aggregation, forming voids where vacancies within them have accumulated. The formation of voids leads to swelling, which is an isotropic volume expansion of the metal that can be, for significant radiation doses, upto several tens of percent in magnitude. This can cause a considerable change in the linear dimensions as well (this scales as the cube root of the swelling percent).
Another phenomenon that can occur, depending on the mechanics of defect-vacancy formation, is growth, where volume remains constant but there is a change in shape of the metallic object. Usually, both swelling and growth occur, at different locations within the irradiated material. In addition, a phenomenon known as irradiation creep can occur, which is the slow deformation of the metal under constant mechanical and radiative stress.
It is interactions like these, and phenomena that occur when they aggregate over successively longer length (and time) scales within the reactor material that eventually have observable consequences on components that have dimensions of the order of meters, i.e., on the macroscopic scale. These consequences include onset of plasticity, fracture, or embrittlement.
Phenomena such as defect clustering and diffusion, formation of dislocations (a dislocation being formed whenever a regular pattern of defects significantly distorts the crystalline order), formation of separate phases and radiation-induced segregation also occur in the irradiated metal, each of which, depending on conditions, can contribute to the onset of plasticity or the start of fracture in a metal.
Zipping up a description starting from the lowest relevant length (and time) scale, up to the longest relevant length (and time) scale, using the most appropriate physical description at each scale - under conditions of high thermomechanical and radiative stress - is the challenge of Multiscale Materials Modeling under Irradiation (MMM-I). A certain amount of coarse graining is inevitable when this is attempted, and both sequential and parallel approaches are possible, as well as hybrid approaches that utilize elements of both methods.
My presentation Multi-scale Modeling of Radiation-Induced-Dislocation Pinning in Oxide-Dispersion Strengthened Steels (ODSS)- Part 1 and Multi-scale Modeling of Radiation-Induced-Dislocation Pinning in Oxide-Dispersion Strengthened Steels (ODSS)- Part 2 describe these ideas and my work in progress, in more detail, in the context of oxide dispersion strengthened steels, which have been suggested as possible reactor materials for new generation reactors in view of their superior creep strength under both radiative and thermomechanical stress, as well as corrosion resistance. This is the focus of my current work.
The Fall 2008 Meeting of the Materials Research Society in Boston has Symposium W on Computational Materials Design by Multiscale Modeling.
Monday, November 10, 2008
Safety Issues in Nuclear Hydrogen Production with the (Gas-Cooled) Very High Temperature Reactor (VHTR)
Nuclear Hydrogen Production (NHP) is the idea that the heat and/or electricity from a nuclear reactor can be used to electrolyse, thermolyse, or thermochemically analyse water - separating hydrogen and oxygen. The resulting hydrogen is useful for a variety of markets including hydrocarbon upgrading & refining, ammonia production, and transportation. The merit of the nuclear hydrogen production scheme is that it does not produce greenhouse gases. Alternative hydrogen production schemes currently in use invariably generate carbon dioxide at some stage. The most common technique is methane steam reforming - here carbon dioxide first arises from the burning of methane to produce heat to generate steam, then again in the reforming step that separates the carbon from the methane by successive oxidation to generate hydrogen. Even the alternative electrolysis method that is also common generates carbon dioxide when fossil fuels are burned in the generation of the electricity - whether from gas-fired or coal-fired plants.
While nuclear hydrogen production (NHP) schemes thus are carbon-dioxide emission free, they do involve the use of corrosive fluids (such as sulphuric acid) at high temperatures; generation of noxious & poisonous gases (such as sulphur dioxide and hydrochloric acid); and issues from the possible generation of radioactive products such as tritium. As well, heat transfer fluids carrying heat from the reactor to the thermochemical plant may interact with the chemicals in the plant. Or, in the current conceptual version of the Very High Temperature Reactor, the coolant might be helium, as might also be the heat transfer fluid. Some contamination of tritium in helium is possible and more might arise from transmutation of helium isotopes. In addition, the very idea of having a thermochemical plant located close to a nuclear reactor is unprecedented. Both the hydrogen and the oxygen produced in the thermochemical scheme will need to be safely stored, to prevent the possibility of leakage, deflagration or detonation.
Such considerations impose a number of requirements on the layout of nuclear plants and thermochemical plants, they impose requirements on the amount and quantity hydrogen and oxygen that may be created and stored on site, and on the location of the control room in a nuclear power plant. They also have implications for probabilistic safety analyses (PSA) for combined nuclear-and-thermochemical plants.
The idea of nuclear hydrogen production is especially attractive also in the context of extraction and processing of crude oil from the Alberta Tar Sands. Several reactor-hydrogen production configuration schemes are currently being examined and considered for this project at the present time. Nuclear steam methane reforming (which generates steam from nuclear heat, and uses nuclear heat in the reaction), for example, is a possible transitional technology that decreases the overall carbon footprint of the process (it does not completely eliminate the carbon dioxide production, since it retains the reforming step). For all of these considerations, it is possible that the Tar Sands may emerge as the locus of the first commercial nuclear hydrogen production project anywhere.
I discussed issues surrounding this in my paper Safety Issues in Nuclear Hydrogen Production with the Very High Temperature Reactor (VHTR) presented at the Canadian Nuclear Society Annual Conference 2008, Toronto.
Materials Challenges for the Supercritical Water-cooled Reactor (SCWR)
The Supercritical Water-cooled Reactor (SCWR) is the most promising evolution of the water-cooled reactor technology that currently dominates the commercial market for nuclear reactors. Essentially the idea is to increase the thermodynamic efficiency of the reactor type by going to higher temperatures and pressures, and thus also going to the supercritical state. Supercritical water also has a much higher specific heat, enabling a higher heat transfer per unit mass. By going to higher pressures and temperatures, it is also possible to avoid phase changes within the coolant loop altogether. This substantially reduces the requirements for pumps and compressors within the coolant loop, simplifying it considerably and introducing even greater economy in the whole power plant.
The merit of the supercritical coolant idea has been tested and the concept has actually been deployed for fossil fuel power plants (coal-fired) already, where thermodynamic efficiencies have considerably risen as a result. Some jurisdictions have also mandated that all future coal-fired power plants be of the supercritical type. The supercritical water coolant imposes more stringent requirements for plant materials used in fossil fuel-fired power plants as well, and this experience is relevant to the Supercritical Water-cooled Reactor (SCWR) concept.
However, what is different about nuclear reactors, of course, is the radiation dose that reactor materials will experience, in addition to the thermochemical stress that the supercritical water environment might impose. This is even more the case when new fast neutron spectrum fuel cycles are introduced, with radiation doses upto a hundred times higher than the thermal spectrum neutrons that current generation reactors use.
Superior irradiation creep strength, and superior thermomechanical behavior in general then becomes a a very desirable critical property to have for cladding materials. I discuss the Materials Challenges for the Supercritical Water-Cooled Reactor in my paper published in the Canadian Nuclear Society Bulletin, Vol. 29. No. 1 pp. 32-38 March 2008.
The supercritical water environment is also chemically different in that it dissolves organic species but not inorganic ones, the reverse of ordinary water. Thus the supercritical water environment is expected to create novel corrosion challenges as well, and these also must be studied and understood.
Materials such as Oxide Dispersion Steels (ODS Steels) have been proposed for use as structural and cladding materials in the Supercritical Water-cooled Reactor SCWR (as well as other Generation IV Reactor concepts). These materials appear to display very desirable irradiation and thermal creep properties, as well as desirable electro-chemical properties (corrosion resistance). Understanding the physical origin of these properties by developing suitable multiscale materials models (MMM) is the focus of my present research.
The merit of the supercritical coolant idea has been tested and the concept has actually been deployed for fossil fuel power plants (coal-fired) already, where thermodynamic efficiencies have considerably risen as a result. Some jurisdictions have also mandated that all future coal-fired power plants be of the supercritical type. The supercritical water coolant imposes more stringent requirements for plant materials used in fossil fuel-fired power plants as well, and this experience is relevant to the Supercritical Water-cooled Reactor (SCWR) concept.
However, what is different about nuclear reactors, of course, is the radiation dose that reactor materials will experience, in addition to the thermochemical stress that the supercritical water environment might impose. This is even more the case when new fast neutron spectrum fuel cycles are introduced, with radiation doses upto a hundred times higher than the thermal spectrum neutrons that current generation reactors use.
Superior irradiation creep strength, and superior thermomechanical behavior in general then becomes a a very desirable critical property to have for cladding materials. I discuss the Materials Challenges for the Supercritical Water-Cooled Reactor in my paper published in the Canadian Nuclear Society Bulletin, Vol. 29. No. 1 pp. 32-38 March 2008.
The supercritical water environment is also chemically different in that it dissolves organic species but not inorganic ones, the reverse of ordinary water. Thus the supercritical water environment is expected to create novel corrosion challenges as well, and these also must be studied and understood.
Materials such as Oxide Dispersion Steels (ODS Steels) have been proposed for use as structural and cladding materials in the Supercritical Water-cooled Reactor SCWR (as well as other Generation IV Reactor concepts). These materials appear to display very desirable irradiation and thermal creep properties, as well as desirable electro-chemical properties (corrosion resistance). Understanding the physical origin of these properties by developing suitable multiscale materials models (MMM) is the focus of my present research.
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