An Assessment of Radiation Damage Models and Methods

The current state of development of the primary models used for investigating and simulating irradiation effects in structural alloys of interest to the U.S. DOE's Generation-IV reactor program are discussed. The underlying theory that supports model development is also described where appropriate. First, the key processes that underlie radiation-induced changes in material properties are summarized, and the types of radiation effects that subsequently arise are described. Future development work needed in order for theory, modeling, and computational materials science to support and add value to the Gen IV reactor materials program are then outlined. The expected specific outcomes and overall benefits of the required effort are: the knowledge to extrapolate material behavior to conditions for which there are no experimental data; systematic understanding of mechanisms and processes to enable confident interpolation between point-by-point experimental observations; acceleration of the development, selection, and qualification of materials for reactor service; and prediction of material response to real-world operating load histories which often involve a complicated superposition of time, temperature, radiation dose rate, and mechanical loading conditions. Opportunities for international collaboration to accelerate progress in all of the required research areas are briefly discussed, particularly in the context of two well coordinated, broad-based research projects on modeling and simulation of radiation effects on materials that are currently funded in Europe. In addition to providing the opportunity for substantial leveraging of the DOE-funded activities in this area, these projects may serve as models for future development within the Gen-IV program. The larger of these two projects, which involves 12 European research laboratories and 16 universities, is called PERFECT and is funded by the European Union. A smaller effort focusing on developing predictive models for fusion reactor materials is funded within the United Kingdom. Increased formal collaboration with these projects by DOE-funded materials scientists would be of substantial benefit to the Gen-IV program

Radiation Damage Theory

This chapter presents an overview of basic radiation damage theory, including older and more recent models, to provide framework, within which radiation effects, such as void swelling, can be rationalized. A complete review of the literature is not attempted, but sufficient references are given to provide a decent introduction to a quite large number of publications in the field. Many derivations are different from and, in our view, more elegant than in the original publications. The work is directed to both theoreticians and experimentalists, and, especially, to those passionate individuals who are going to take the radiation damage theory (RDT) to the future

Theoretical Investigation of Microstructure Evolution and Deformation of Zirconium Under Cascade Damage Conditions

This work is based on our reaction-diffusion model of radiation growth of Zr-based materials proposed recently in [1]. In [1], the equations for the strain rates in unloaded pure crystal under cascade damage conditions of, e.g., neutron or heavy-ion irradiation were derived as functions of dislocation densities, which include contributions from dislocation loops, and spatial distribution of their Burgers vectors. The model takes into account the intra-cascade clustering of self-interstitial atoms and their one-dimensional diffusion; explains the growth stages, including the break-away growth of pre-annealed samples; and accounts for some striking observations, such as of negative strain in prismatic direction, and co-existence of vacancy- and interstitial-type prismatic loops. In this report, the change of dislocation densities due to accumulation of sessile dislocation loops is taken into account explicitly to investigate the dose dependence of radiation growth. The dose dependence of climb rates of dislocations is calculated, which is important for the climb-induced glide model of radiation creep. The results of fitting the model to available experimental data and some numerical calculations of the strain behavior of Zr for different initial dislocation structures are presented and discussed. The computer code RIMD-ZR.V1 (Radiation Induced Microstructure and Deformation of Zr) developed is described and attached to this report

THEORETICAL INVESTIGATION OF MICROSTRUCTURE EVOLUTION AND DEFORMATION OF ZIRCONIUM UNDER CASCADE DAMAGE CONDITIONS

This work is based on our reaction-diffusion model of radiation growth of Zr-based materials proposed recently in [1]. In [1], the equations for the strain rates in unloaded pure crystal under cascade damage conditions of, e.g., neutron or heavy-ion irradiation were derived as functions of dislocation densities, which include contributions from dislocation loops, and spatial distribution of their Burgers vectors. The model takes into account the intra-cascade clustering of self-interstitial atoms and their one-dimensional diffusion; explains the growth stages, including the break-away growth of pre-annealed samples; and accounts for some striking observations, such as of negative strain in prismatic direction, and co-existence of vacancy- and interstitial-type prismatic loops. In this report, the change of dislocation densities due to accumulation of sessile dislocation loops is taken into account explicitly to investigate the dose dependence of radiation growth. The dose dependence of climb rates of dislocations is calculated, which is important for the climb-induced glide model of radiation creep. The results of fitting the model to available experimental data and some numerical calculations of the strain behavior of Zr for different initial dislocation structures are presented and discussed. The computer code RIMD-ZR.V1 (Radiation Induced Microstructure and Deformation of Zr) developed is described and attached to this report

Materials Degradation in Light Water Reactors: Life After 60,???

Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor

Improving atomic displacement and replacement calculations with physically realistic damage models

Atomic collision processes are fundamental to numerous advanced materials technologies such as electron microscopy, semiconductor processing and nuclear power generation. Extensive experimental and computer simulation studies over the past several decades provide the physical basis for understanding the atomic-scale processes occurring during primary displacement events. The current international standard for quantifying this energetic particle damage, the Norgett-Robinson-Torrens displacements per atom (NRT-dpa) model, has nowadays several well-known limitations. In particular, the number of radiation defects produced in energetic cascades in metals is only similar to 1/3 the NRT-dpa prediction, while the number of atoms involved in atomic mixing is about a factor of 30 larger than the dpa value. Here we propose two new complementary displacement production estimators (athermal recombination corrected dpa, arc-dpa) and atomic mixing (replacements per atom, rpa) functions that extend the NRT-dpa by providing more physically realistic descriptions of primary defect creation in materials and may become additional standard measures for radiation damage quantification.Peer reviewe

Primary radiation damage : A review of current understanding and models

Scientific understanding of any kind of radiation effects starts from the primary damage, i.e. the defects that are produced right after an initial atomic displacement event initiated by a high-energy particle. In this Review, we consider the extensive experimental and computer simulation studies that have been performed over the past several decades on what the nature of the primary damage is. We review both the production of crystallographic or topological defects in materials as well as radiation mixing, i.e. the process where atoms in perfect crystallographic positions exchange positions with other ones in non-defective positions. All classes of materials except biological materials are considered. We also consider the recent effort to provide alternatives to the current international standard for quantifying this energetic particle damage, the Norgett-Robinson-Torrens displacements per atom (NRT-dpa) model for metals. We present in detail new complementary displacement production estimators ("athermal recombination corrected dpa", arc-dpa) and atomic mixing ("replacements per atom", rpa) functions that extend the NRT-dpa, and discuss their advantages and limitations. (C) 2018 The Authors. Published by Elsevier B.V.Peer reviewe

Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development