Thermal Kinetics of Helium Irradiation Hardening in Selected Alloys for the Canadian Gen. IV Nuclear Reactor Concept

In this thesis, we present nano-indentation measurements performed to quantify the increase in hardness as a result of He+ and Fe4+ implantation in both Inconel 800H and AISI 310 alloys. After annealing, the softening rate of He+ and Fe4+ implanted samples were compared, and it is found that Ni can slow the helium diffusion. Thermal activation energy Q characterizing this process was similar to the computed thermal activation energy QHe for interstitial helium diffusion within pure nickel. Indentation hardness tests were also performed at various indentation strain rates, to further study the effect of implanted helium as an obstacle to plastic deformation. It was observed, for both AISI 310 and Inconel 800H, the strain rate sensitivity m decreases, and the activation volume V* increases significantly after annealing. This suggests that helium defects (voids or bubbles) within the metal become more stable with annealing, as they tend to form bigger bubbles in the grain boundaries

Micro mechanical testing of candidate structural alloys for Gen-IV nuclear reactors

Ion irradiation is often used to simulate the effects of neutron irradiation due to reduced activation of materials and vastly increased dose rates. However, the low penetration depth of ions requires the development of smallscale mechanical testing techniques, such as nanoindentation and microcompression, in order to measure mechanical properties of the irradiated material. In this study, several candidate structural alloys for Gen-IV reactors (800H, T91, nanocrystalline T91 and 14YWT) were irradiated with 70 MeV Fe9+ ions at 452 °C to an average damage of 20.68 dpa. Both the nanoindentation and microcompression techniques revealed significant irradiation hardening and an increase in yield stress after irradiation in austenitic 800H and ferritic-martensitic T91 alloys. Ion irradiation was observed to have minimal effect on the mechanical properties of nanocrystalline T91 and oxide dispersion strengthened 14YWT. These observations are further supported by line broadening analysis of X-ray diffraction measurements, which show a significantly smaller increase in dislocation density in the 14YWT and nanocrystalline T91 alloys after irradiation. In addition, good agreement was observed between cross-sectional nanoindentation and the damage profile from SRIM calculations

Kinetic and Thermodynamic Modeling of Long Term Phase Stability in Alloy 800H Subjected to LWR Core Conditions

An in depth literature review of Incoloy Alloy 800H was conducted and presented to summarize the current understanding of microstructural evolution under irradiation and secondary phase precipitate stability. Due to a lack of radiation induced segregation (RIS) data for Alloy 800H, Isopleth sections varying Cr, Ni, Ti, and Si were generated from a computational thermodynamics approach using ThermoCalc and analyzed to compensate for knowledge related to radiation induced precipitates (RIP’s). These isopleths were analyzed for a composition range based off previous knowledge of RIS tendencies in austenitic stainless steels. Analysis of four major binary phase diagrams and complex phase diagrams calculated through ThermoCalc and MatCalc were used to simulate the precipitation kinetics during the lifetime of Incoloy Alloy 800H used in a light water reactor (LWR) core setting. These aging simulations were then conducted using the MatCalc heat treatments tool with M23C6 [Chromium Carbide], Sigma, and Ni3Ti [Gamma Prime] set as the precipitates of interest. A discrepancy was found relating to the presence of sigma phase at low temperatures between ThermoCalc and MatCalc complex phase diagram calculations. Several minor phases were noted from the complex phase diagrams and isopleths for further research. Isopleth sections revealed that no major RIP’s should form given the current assumption of RIS behavior. Simulations of precipitation kinetics predict a precipitate coarsening somewhere between 6-7 years of operation for M23C­6 [Chromium Carbide] precipitates. This results in a decline in number density and an increase in precipitate size. Anticipated radiation induced segregation has very little effect on M23C6­ [Chromium Carbide] precipitate size, however increasing RIS results in the formation of fewer M23C6 [Chromium Carbide] precipitates. Sigma phase is found to increase in amount and decrease in size as segregation increases until the number of precipitates reaches a maximum between 20.42 and 14.42 wt% Cr. At doses greater than this the density of sigma precipitates is expected to decrease while the size of precipitates is expected to remain consistent

Thermal kinetics of ion irradiation hardening in selected alloys for the Canadian Gen. IV nuclear reactor concept

Canada is designing supercritical water fission reactors (SCWR) to increase the thermal efficiency of nuclear power generation from ~34% to ~48%. The temperature and pressure of a supercritical water reactor core is very high compared to a CANDU reactor. This thesis examines irradiation hardening and thermal recovery of two candidate alloys, AISI 310 and Inconel 800H, for the Canadian SCWR. Samples of both alloys are mechanically ground and polished, then irradiated using 8.0 MeV Fe ions. The use of ion irradiation safely and quickly simulates neutron damage. The change in the hardness of the samples is then studied during a series of thermal anneals at temperatures ranging from 400° to 600° C. This study found virtually all irradiation-induced hardening had recovered within 100 minutes of exposure to these temperatures

Scientific Assessment in support of the Materials Roadmap enabling Low Carbon Energy Technologies: Technology Nuclear Energy

This scientific assessment serves as the basis for a materials research roadmap for the nuclear fission technology, itself an integral element of an overall "Materials Roadmap Enabling Low Carbon Technologies", a Commission Staff Working Document published in December 2011. The Materials Roadmap aims at contributing to strategic decisions on materials research funding at European and Member State levels and is aligned with the priorities of the Strategic Energy Technology Plan (SET-Plan). It is intended to serve as a guide for developing specific research and development activities in the field of materials for energy applications over the next 10 years. This report provides an in-depth analysis of the state-of-the-art and future challenges for energy technology-related materials and the needs for research activities to support the development of nuclear fission technology both for the 2020 and the 2050 market horizons. It has been produced by independent and renowned European materials scientists and energy technology experts, drawn from academia, research institutes and industry, under the coordination the SET-Plan Information System (SETIS), which is managed by the Joint Research Centre (JRC) of the European Commission. The contents were presented and discussed at a dedicated hearing in which a wide pool of stakeholders participated, including representatives of the relevant technology platforms, industry associations and the Joint Programmes of the European Energy Research Associations.JRC.F.4-Safety of future nuclear reactor

AUSTENITIC STAINLESS STEELS FOR FUTURE NUCLEAR FUEL CLADDINGS

Nuclear power systems have been under continuous development since the first nuclear power plant started operation in 1954. They are categorized into different generations, with each new generation having significant technological advances over the previous one. The worldwide effort to develop the next generation of nuclear reactors was defined at the Generation IV International Forum (GIF) in 2000. Six types of design were proposed, including supercritical water cooled reactor (SCWR). Materials in this reactor will be exposed to more severe environments than the current generation of reactors to assure higher efficiency in energy production and the current materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, the behavior of two existing nuclear materials, stainless steels 310S and 316L was investigated, under conditions approximating the nuclear reactor environment. An environment with dynamic loop of supercritical water (SCW) was used to test the performance of the alloys and the oxides formed were analyzed. Oxidation of the alloys in air was also performed for comparison. It was found that although both alloys showed good oxidation resistance in air at 600ºC, stainless steel 310S has better resistance in SCW environment compared to stainless steel 316L. A thin protective oxide layer of Mn2CrO4 spinel delays oxidation in alloy 310S. In order to improve the oxidation resistance of 310S and 316L stainless steels, thermo-mechanical processing (TMP) was applied to modify their microstructures. The deformation and annealing texture of the as-received and processed samples were investigated by means of X-ray diffraction (XRD) and orientation imaging microscopy (OIM). Different rolling paths and different deformation levels before annealing were used to produce samples of different grain size with similar texture and samples of similar grain size with different textures. Subsequently, the oxidation resistance of thermo-mechanically processed 316L and 310S samples in SCW was studied. It was found that the oxidation resistance of stainless steels 316L and 310S can be improved up to four and five times, respectively, by decreasing the grain size below a critical value of 3 µm. It was demonstrated that samples with smaller grain size provided higher fraction of grain boundaries for fast diffusion of chromium to reach the surface and compensate losses due to dissolution of chromium in the oxidation media. External oxide layers formed on as-received and thermo-mechanically processed stainless steel 316L samples was characterized to establish possible correlation between orientation of the substrate and oxide grains. Micro and macro textures of the substrate and the oxide layers were examined and the results showed that the texture of substrate did not affect the texture of magnetite (Fe3O4) in the upper oxide layer. In addition, the texture of magnetite did not affect the texture of hematite (Fe2O3) on samples where hematite was an additional oxide phase. The strong texture of both oxides was explained with surface free energy minimization and strain energy minimization theory. This means that the texture of both oxides is dictated by a competition between their surface and strain energies

AUSTENITIC STAINLESS STEELS FOR FUTURE NUCLEAR FUEL CLADDINGS

Nuclear power systems have been under continuous development since the first nuclear power plant started operation in 1954. They are categorized into different generations, with each new generation having significant technological advances over the previous one. The worldwide effort to develop the next generation of nuclear reactors was defined at the Generation IV International Forum (GIF) in 2000. Six types of design were proposed, including supercritical water cooled reactor (SCWR). Materials in this reactor will be exposed to more severe environments than the current generation of reactors to assure higher efficiency in energy production and the current materials used for fuel cladding need to be improved or new materials should be developed. In this thesis, the behavior of two existing nuclear materials, stainless steels 310S and 316L was investigated, under conditions approximating the nuclear reactor environment. An environment with dynamic loop of supercritical water (SCW) was used to test the performance of the alloys and the oxides formed were analyzed. Oxidation of the alloys in air was also performed for comparison. It was found that although both alloys showed good oxidation resistance in air at 600ºC, stainless steel 310S has better resistance in SCW environment compared to stainless steel 316L. A thin protective oxide layer of Mn2CrO4 spinel delays oxidation in alloy 310S. In order to improve the oxidation resistance of 310S and 316L stainless steels, thermo-mechanical processing (TMP) was applied to modify their microstructures. The deformation and annealing texture of the as-received and processed samples were investigated by means of X-ray diffraction (XRD) and orientation imaging microscopy (OIM). Different rolling paths and different deformation levels before annealing were used to produce samples of different grain size with similar texture and samples of similar grain size with different textures. Subsequently, the oxidation resistance of thermo-mechanically processed 316L and 310S samples in SCW was studied. It was found that the oxidation resistance of stainless steels 316L and 310S can be improved up to four and five times, respectively, by decreasing the grain size below a critical value of 3 µm. It was demonstrated that samples with smaller grain size provided higher fraction of grain boundaries for fast diffusion of chromium to reach the surface and compensate losses due to dissolution of chromium in the oxidation media. External oxide layers formed on as-received and thermo-mechanically processed stainless steel 316L samples was characterized to establish possible correlation between orientation of the substrate and oxide grains. Micro and macro textures of the substrate and the oxide layers were examined and the results showed that the texture of substrate did not affect the texture of magnetite (Fe3O4) in the upper oxide layer. In addition, the texture of magnetite did not affect the texture of hematite (Fe2O3) on samples where hematite was an additional oxide phase. The strong texture of both oxides was explained with surface free energy minimization and strain energy minimization theory. This means that the texture of both oxides is dictated by a competition between their surface and strain energies