Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage
(Sponsor: Department of Energy - Nuclear Energy University Program, IRP) (2015-2018)
Collaborators: Colorado School of Mines (leading university), NCSU, ANL, LANL, SNL, CB&I
Lifetime extension of dry storage canisters requires the ability to accurately predict and monitor material degradation so that corrective maintenance actions can be taken. Monitoring and inspection of dry storage facilities in combination with material property prediction capabilities are necessary. Chloride-initiated stress corrosion cracking (CISCC) of a spent fuel canister (primarily in welds or heat affected zones) is one of the safety concerns during the dry storage of used nuclear fuel at an Independent Spent Fuel Storage Installations (ISFSIs). Deterioration by CISCC can lead to canister penetration, potentially releasing helium and radioactive gases, and permitting air ingress which could pose a threat to fuel rod integrity. This study will result in enhanced understanding of conditions which could be conducive to CISCC initiation (such as pitting) or CISCC propagation rate, and will develop methods that could be used to identify the occurrence of CISCC in its early stages in the field. The model and methodology developed in the proposed project with quantified uncertainty can be used to inform recommendations for periodic NDE examinations to monitor the extent of any cracking.
Laboratory CISCC studies envisioned in the work
frame of this IRP include: Testing to quantify the effect of environmental and
metallurgical factors that have an impact on SCC initiation and growth rate,
such as salt concentrations, temperature, most susceptible heat affected zone
microstructure, metal stresses, pH, and relative humidity; Experimental testing
to determine the most susceptible zone within the weld heat affected zones;
Controlled and instrumented pitting initiation and crack propagation rate
studies with specimens representing the most susceptible microstructure, varying
environmental parameters and specimen tensile stress conditions to cover the
range expected on the canister surfaces.
Mechanistic and Validated Creep/Fatigue Predictions for Alloy 709 from Accelerated Experiments and Simulations
(Sponsor: Department of Energy - Nuclear Energy University Program) (2015-2018)
Collaborators: NCSU, ANL
As a promising candidate for fast reactor program, Alloy 709 possesses excellent high temperature thermomechanical properties. To support its qualification in the ASME code for Class 1 Components in Elevated Temperature Service (Section 3, Division 1, Subsection NH), we propose mechanistic methods for predicting creep and creep-fatigue deformation rates based on accelerated in-situ and ex-situ tests, and mesoscale dislocation dynamics (DD) simulations. The research work performed in this project will aim at obtaining: (i) creep and creep-fatigue data (ii) microstructure evaluation from (a) in-situ/ex-situ TEM, (b) in-situ XRD using synchrotron radiation at APS/ANL and (c) mesoscale dislocation dynamics simulations informing creep damage mechanics (CDM) model; (iii) a rational framework (CDM) of generalized viscoplastic constitutive equations to reliably predict and extrapolate the results of accelerated tests to reactor operating conditions; (iv) validations of CDM performed through predictions that can be crosschecked and benchmarked against experimental data; and (v) extrapolated creep and creep-fatigue data delivered for use in ASME code development.
"High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation"(Sponsor: Department of Energy - Nuclear Energy University Program) (2014-2017)
Participants: U. of Michigan, (lead university), U. Tennessee; Penn State; U. of Wisconsin; U of South Carolina; U. C. Berkeley; U. C. Santa Barbara; ORNL; LANL; LLNL; ANL; INL; TerraPower LLC; EPRI; U. Manchester; U. Oxford; Areva; U. Queens; CEA.
The objective of this collaborative effort is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations.
The promise for developing new, advanced nuclear reactor concepts that significantly improve on the safety, economics, waste generation and non-proliferation security of commercial nuclear power reactors, and the extension of life of existing light water nuclear reactors rests heavily on understanding how radiation degrades materials that serve as the structural components in reactor cores. In high dose fission reactor concepts such as the sodium fast reactor (SFR), lead fast reactor (LFR), molten salt reactor (MSR) and the traveling wave reactor (TWR), structural materials must survive up to 200 dpa of damage at temperatures in excess of 400°C. At such high damage levels, the major degradation modes are likely to be driven by void swelling and phase stability. Void swelling occurs at homologous temperatures of 0.35–0.55TM, which for steels (325–650°C) overlaps the temperature range of the reactor core for these four high-temperature, high-dose concepts. Exacerbating the problem is phase instability at high doses due to radiation-induced or -enhanced solute segregation and ballistic dissolution of precipitates by energetic displacement cascades. Irradiation can nucleate or dissolve phases, changing the solute composition of the matrix and enhancing void growth. Further, dissolution of particles added to increase the strength of the alloy results in softening and compromises high-temperature strength and creep. For example, γ’ matrix precipitates that provide strength to nickel-base alloys used in high-temperature applications are unstable under irradiation. Radiation can induce the formation of brittle phases along grain boundaries and other defect sinks, reducing ductility and degrading fracture toughness. Light water reactor core components have been plagued by irradiation assisted stress corrosion cracking, which is one of the most life-threatening degradation modes in LWRs, and will be exacerbated in the supercritical water reactor. Thus, virtually all existing LWRs and advanced reactor concepts require a solution to the radiation damage problem. As materials degradation due to irradiation is both a life-limiting and a concept-validating phenomenon, it is truly the grand challenge for the growth and vitality of nuclear energy world-wide.
Traditionally, research to understand radiation-induced changes in materials is conducted via radiation effects experiments in test reactors (both fast and thermal), followed by a comprehensive post-irradiation characterization plan. This is a very time consuming process because of the low damage rates that even the highest flux reactors exhibit. This fact prevents radiation damage research from “getting ahead” of problems discovered during operation. In addition the dearth of available test reactors worldwide, and the high cost of research on irradiated materials in the face of shrinking budgets put additional constraints on this approach. All of these constraints have compromised our ability to advance the understanding of neutron radiation effect at high doses. A promising solution to the problem is to use ion irradiation to irradiate materials to very high doses. The advantages of ion irradiation are many. Dose rates (typically 10-3 to 10-4 dpa/s) are much higher than under neutron irradiation (10-7 to 10-8 dpa/s), which means that 100s of dpa can be reached in days or weeks instead of years. Because there is little activation the samples are not radioactive. Control of ion irradiation experiments is much better than experiments in reactor. Measurement of temperature, damage rate and damage level is difficult in reactor, resulting in reliance on calculations to determine the total dose, and estimate irradiation temperature. By contrast, ion irradiations have been developed to the point where temperature is extremely well controlled and monitored, and damage rate and total damage are also measured continuously throughout the irradiation and with great accuracy.
Challenges to the implementation of ion irradiation as a surrogate for neutron irradiation include rate effects, small irradiation volumes, accounting for transmutation and the lack of data to establish the equivalence. Addressing these challenges constitutes the main focus of this program. This project will demonstrate the capability to evaluate the behavior of reactor materials at high irradiation doses. This effort will include a benchmarking of the microstructures formed under ion irradiation and neutron irradiation and the resulting mechanical properties by a combined experimental and analytical approach. This will be a multiphysics effort from atomistic defect studies and production, to microstructure development to the effect of these microstructure changes on mechanical properties. The final product will provide a path and a methodology for qualifying materials for service at very high doses using ion irradiation.The outcome of this program will be the establishment of the conditions by which ion irradiation can be used as a surrogate for neutron irradiation in reactor. It will provide a path for materials qualification at high doses with a significant reduction in both the time and cost to evaluate core materials response to irradiation, as well as the implementation of materials and processes to improve the safety and economics of the existing light water reactor fleet and advanced reactor designs.
Participants: University of California, Berkeley; University of South Carolina; Los Alamos National Laboratory.
For new alloying ideas, it is highly desirable to only perform small experimental heats using smaller and smaller materials testing in order to avoid the costs of manufacturing large quantities. Accelerated materials testing, is important in order to achieve high doses quickly to enable new materials concepts under radiation and lead the way towards their qualification. Most accelerated materials testing approaches involve ion beam irradiation or high dose neutron irradiation. Ion beam accelerators only have a limited penetration depth into a material (allowing only μm of irradiated materials on a given sample). On the other hand, neutron irradiated materials are difficult to deal with due to activation concerns and there is often only a limited amount of material available. Regardless, both approaches call for the development of small-scale materials testing techniques and the need to link these techniques to bulk properties. Therefore, the development of novel small-scale mechanical testing in combination with microstructural investigation and modeling is of great interest to the nuclear materials community for both materials development as well as monitoring applications. Therefore in this work the combination of modeling and experiments on multiple length scales will be used to evaluate and improve existing small scale mechanical testing techniques in order to help make them relevant to macroscopic properties and useful nuclear engineers, inspectors and designers.
It is the declared goal of this project to develop new small-scale mechanical testing techniques (e.g. hot/cold hardness/compression, tension, bending, ductile to brittle transition temperature) to allow for the estimation or direct measurement of bulk properties. The outcome of our combined experiments and modeling will significantly enhance the statistics and information that can be obtained on small radioactive archived samples as well as new ion beam irradiated specimens.This work will be conducted by close collaboration between experiments and modeling. In particular, we will focus on in situ experimental efforts that will allow us to understand mechanisms of materials deformation. Significant attention is given to nuclear engineering students with a focus in material science and material science students interested in nuclear engineering. This work will engage students in novel materials characterization techniques as well as require “out of the box” thinking to obtain the maximum amount of mechanical property information from irradiated materials. The fact that a lot of the experiments can be conducted in-situ enhances the visual “seeing is believing” output for the students and leads to a stronger engagement of the student community.
main objective is to understand deformation and fracture mechanisms in
high-temperature structural materials to ensure structural integrity and
lifetime prediction of the components (advanced steels, Ni-based alloys, SiC).
The lack of understanding of the deformation mechanisms in such marerials has
limited the development of predictive capabilities.
- Custom-made high-temperature mechanical testing machine:
- Frame capacity: 150kN
- Different load cells((30kN, 50kN,100kN) for improved precision at lowand high load ranges
- Furnace for hightemperature testing (up to 1000 C)
The goal is to elucidate the mechanisms of deformation and creep resistance in these high temperature advanced materials using a mechanistic approach which will allow for predictive mechanistic models to be derived. The study includes in-situ straining experiments to evidence the mechanisms of dislocation dynamics.Related publications: - Kaoumi D., K. Hrutkay, “Tensile deformation behavior and microstructure evolution of Ni-based superalloy 617”, Journal of Nuclear Materials, 454, 2014, p 265-273. - Hrutkay K., D. Kaoumi, “Tensile deformation behavior of a nickel based superalloy (Haynes 230) at different temperatures”, Materials Science & Engineering A, 599, 2014, p. 196–203
"Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Simulation" (Sponsor: Department of Energy - Nuclear Energy University Program) (Oct 2010 - Oct 2013)
Participants: University of Tennessee, University of South Carolina, University of Wisconsin, Penn State.
The in-service degradation of reactor core materials is related to underlying
changes in the irradiated microstructure. During reactor operation, structural
components and cladding experience displacement of atoms by collisions with
neutrons at temperatures at which the radiation-induced defects are mobile,
leading to microstructure evolution under irradiation that can degrade material
properties. At the doses and temperatures relevant to fast reactor operation,
the microstructure evolves by dislocation loop formation and growth,
microchemistry changes due to radiation-induced segregation, radiation-induced
precipitation, destabilization of the existing precipitate structure, and in
some cases, void formation and growth. These processes do not occur
independently; rather, their evolution is highly interlinked. Radiation-induced
segregation of Cr and existing chromium carbide coverage in irradiated alloy T91
track each other closely. The radiation-induced precipitation of Ni-Si
precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related.
Neither the evolution of these processes nor their coupling is understood under
the conditions required for materials performance in fast reactors (temperature
range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not
yet possible as models for microstructure evolution must be developed along with
experiments to characterize these key processes and provide tools for
extrapolation. To extend the range of operation of nuclear fuel cladding and
structural materials in advanced nuclear energy and transmutation systems to
that required for the fast reactor, the irradiation-induced evolution of the
microstructure, microchemistry, and the associated mechanical properties at
relevant temperatures and doses must be understood. Predictive modeling relies
on an understanding of the physical processes and also on the development of
microstructure and microchemical models to describe their evolution under
Denuded zones developing at grain boundaries (D. Kaoumi, J. Adamson, M. Kirk, Journal of Nuclear Materials, 445 (1–3): p. 12–19, 2014).
This project focuses on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime.
Development of Oxide Dispersion Strengthened (ODS) steels by alternative processing methods
Nanostructured oxide-dispersion strengthened (ODS) steels are candidates as a first-wall structural material for the fusion system, and as fuel cladding for fast-breeder reactors. Alternative ways of fabricating these alloys are sought which for cost effectiveness. The applicability of a novel processing technique using powders as the starting material is investigated at the University of South Carolina. This is an emerging metal working technique that causes intense plastic deformation, material mixing and thermal exposure, resulting in significant microstructural refinement, densification and homogeneity of the processed zone. Compared to other processing routes, it is a short-route, one-step, solid-state processing technique, and the microstructure of the processed zone can be controlled by optimizing the processing parameters.
Union Process Attritor
01HD Research Model