Stress Corrosion Cracking of Neutron Irradiated Cast Stainless Steels in High Temperature Water

Participants: S. Teysseyre, PI and G. S. Was, Co-PI
Sponsor: U.S. Department of Energy/UT-Batelle, LLC

  The U.S. contribution to ITER(International Tokamak Experimental Reactor) program includes about 20% of the first wall and shield, consisting of 93 modules each weighing about 3.5T and 375 FW panels. There is a potential for significant cost savings by utilizing casting technology rather than welding/HIPing wrought plate material and employing extensive machining to fabricate the shield module. It is the responsibility of the US-program to demonstrate that the utilization of cast material will not impair the mechanical performance and corrosion behavior of the shield module.

  The objective of this project is to investigate the stress corrosion cracking susceptibility of cast stainless steel in both unirradiated and neutron-irradiated condition in order to determine whether cast stainless steel can function in its intended role in ITER. The program includes the development of the facility for testing neutron-irradiated stainless steels in controlled water chemistry at temperatures below 300°C. Experiments performed in a controlled water environment will be conducted to determine the baseline stress corrosion cracking behavior of the unirradiated cast alloy and the behavior of neutron-irradiated cast alloy.

Accelerator-Based Study of Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for Very High Temperature Reactors (VHTR)

Participants: L. M. Wang, PI and G. S. Was, Co-PI; R. S. Zhou, Post-doctoral Scholar and A. Davis, Graduate Student
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI)

  Pyrolytic carbon (PyC) is one of the structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors. When the TRISO particles are under irradiation, creep of the pyrocarbon layers can cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  The primary objective of this project is to characterize the creep behavior of PyC through a systematic program of accelerator-based proton irradiation and in-situ measurements under stress at various temperatures between 400°C and 1,200°C. Test data will be analyzed to determine creep coefficients, which will then be correlated to existing coefficients measured under neutron irradiation. In addition, initial experiments on the transport of select fission products (e.g., Ag and Sr) in PyC under irradiation and stress will be conducted by implanting ions into the sample surface. The PyC microstructure will be studied with advanced analytical transmission electron microscopy (TEM).

Consortium on Cladding and Structural Materials for Advanced Nuclear Energy Systems

Participants: G. S. Was, PI and L. Wang, Co PI; with U. Wisconsin, U. C. Berkley, U. C. Santa Barbara, Penn State Univ. and Alabama A&M
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI) and Electric Power Research Institute

  The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that are yet to be resolved, or that are beyond the existing experience (dose/burnup) base, in order to 1) provide for a sound fundamental and engineering basis for operation in the intended application, 2) bring together key university, national laboratory and industry capability and support in order to provide the most comprehensive approach possible, and 3) create a long term, evolutionary program that seeks to address these and future nuclear materials issues in a progressive manner. This consortium will serve as a nucleation site, about which materials research activities will be catalyzed and grown among the leading individuals and institutions from academia, the national laboratories and industry. It represents an unprecedented opportunity to combine expertise and facilities in an effort to attack the challenge of materials behavior under irradiation on a scale that is not feasible for a single individual or institution.

  The objectives of the initial three-year phase of the consortium are to:
• Develop an understanding of the high dose radiation stability of candidate sodium fast reactor (SFR) cladding and duct alloys under a range of temperatures and doses expected in the SFR, using a closely integrated program combining targeted charged particle and neutron irradiations, in-situ irradiation and computer simulation of defect microstructure
• Determine the stability of oxide dispersion strengthened (ODS) steel and ultrafine, precipitation strengthened (HT-UPS) austenitic steel
• Characterize and understand the mechanisms of irradiation creep in SiC in TRISO fuel, ferritic-martensitic (F-M) alloys and ODS and UT-UPS steels
• Develop barrier layers for protection of F-M alloys from fuel-clad chemical interaction, and of alloy 617 from attack by coolant impurities in the VHTR intermediate heat exchanger
• Develop modeling tools to explain the behavior of F-M steels under irradiation, and predictive tools to extend the reach of our understanding beyond the experimental database

  The objectives will be accomplished in a research program consisting of three major thrusts: 1) high dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature and 3) innovative cladding concepts embodying functionally-graded barrier materials. While the initial three-year program will emphasize ion irradiation and irradiated microstructures, we expect that, if successful, the second three-year program will increasingly emphasize reactor irradiations and will include mechanical property determination through national user facilities. Industry partners (EPRI and GE) will utilize the core program as leverage to guide or support additional activities that are of special interest to them, and that fall within the scope of the core program. National laboratory partners (ANL, INL, LANL, ORNL and PNNL) will provide additional capability and direction to various aspects of the core program that are of interest to them. Our technical society partner, ASME, will introduce the data generated by the consortium into the ASME Codes & Standards (C&S) process. Beyond scientific achievements, this consortium will provide substantial additional outcomes that are expected to provide long term benefits to the advanced rector program, including the education of around eight graduate students and several post-docs, inclusion of minority students into the radiation effects and reactor materials fields through the participation of Alabama A&M University (a HBCU institution), creation of new working relationships between universities, laboratories and industry in an unprecedented manner and to an unprecedented degree, and establishment of a pathway to begin to incorporate data generated by the research thrusts into the ASME codes and standards that will be crucial for success of the advanced reactor programs.

Acquisition of a Research Grade Ion Implanter for Research and Education in Ion Beam Modification of Materials

Participants: G. S. Was, PI and L. M. Wang, Co-PI (with K. Najafi of EECE and R. Goldman of MSE)
Sponsor: National Science Foundation

  A new and highly versatile ion implanter will provide greatly expanded capabilities to the University’s research programs, attract new research projects and foster the training of graduate and undergraduate students in ion-solid interactions. The 400 kV ion implanter made by National Electrostatics Corporation consists of an ion source and lens system, a gas supply system, a 90° analyzing magnet, a mass defining slit, beam position monitor, accelerator tube, and electrostatic quadrupole triplet lens, a beamline with a Faraday cup, neutral beam trap and raster-scanner, and a target station capable of 6 inch (150 mm) wafer handling, a four-position faraday cup arrangement for dose measurement and target temperature control from LN2 temperature to 800°C, and an ion source (Danfysik model 921A) for the production of high current and high brightness ion beams. Its versatility is due to its ability to ionize materials that have a low vapor pressure by using an oven to heat the charge materials to several hundred degrees, giving it the capability of making ions from a large fraction of the periodic chart. The implanter will be utilized immediately in research programs encompassing a wide range of scientific disciplines and focusing on nanoparticle formation in metals and ceramics, semiconductor nanostructures and heterostructures, atomic and molecular structure modification, and biomedical device materials. Examples of some of the novel uses of this facility are the formation of 3-D arrays of nanostructures to enhance physical and mechanical properties of materials, semiconductor nanopatterning by seeding the formation of nanometer-sized arrays of semiconductor structures, synthesis of bipolar quantum dot thermoelectric devices, femtosecond laser-assisted molecular beam epitaxy, refractive index patterning and the improvement of photoactive devices via ion implantation, and improved adherence of polymer coatings used in next-generation embolization coils for treating neurovascular defects, such as aneurysms and brain tumors. It will also play the lead role in providing surface modification capability to users of the NSF National Nanotechnology Infrastructure Network (NNIN) at the Michigan node. Overall, this implanter will provided a critical resource to 14 active research programs encompassing the work of 28 faculty in 9 departments at Michigan and representing over $22M of active or pending research programs, and will provide a unique resource to surrounding and partner schools. A significant role of the implanter will be to promote the teaching, training and education of graduate, undergraduate students and post-docs in surface modification and materials at the nanoscale, through research projects and formal classes, and to provide special programs for undergraduate students and K-12 outreach.

Alloys for 1000°C Service in the Next Generation Nuclear Plant

Participants: G. S. Was, PI (with J. W. Jones and T. Pollock), D. Kumar and J. Kim, Graduate Students
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI) , Idaho National Laboratory

  The objective of the proposed research is to define strategies for the improvement of alloys for structural components, such as the intermediate heat exchanger and primary-to-secondary piping, for service at 1000°C in the He environment of the NGNP. Specifically, we will investigate the oxidation/carburization behavior and microstructure stability and how these processes affect creep. While generating this data, the project will also develop a fundamental understanding of how impurities in the He environment affect these degradation processes and how this understanding can be used to develop more useful life prediction methodologies. Our initial studies will focus on the mechanisms controlling the high temperature degradation of nickel-base alloy 617.

  Understanding the degradation mechanisms will allow us to predict long-term behavior (to extrapolate lab results to long-term service performance) and to identify an effective approach to modify existing alloys for improved performance. To achieve the latter, we will also investigate two material modification strategies; alloy modifications that provide additional solid solution strengthening and reduce interdiffusion (and therefore creep), and grain boundary engineering to reduce creep. The alloy selection and materials requirements will be based on the Next Generation Nuclear Plant Materials Selection and Qualification Program Plan (INEEL/EXT- 03-01128) and the research plan will be closely integrated with, and designed to complement ongoing and planned studies on alloy 617 at INEEL and ORNL. The research will also provide a platform for educating students in the area of nuclear reactor materials and related issues.

BWRVIP Highly Irradiated Stainless Steel Crack Growth

Participants: G. S. Was, PI, S. Teysseyre, Assistant Research Scientist
Sponsor: General Electric

  This program focuses on post-test fracture surface examination of CGR samples in a scanning electron microscope (SEM) in a hot cell, in support of a larger program being conducted by General Electric for the Electric Power Research Institute. The microscope we will be using is a Philips Quanta-HiVac SEM. This instrument is ideal for hot cell applications since the vacuum and column system can be separated from the computer control. Therefore, the instrument can be moved into the hot cell when needed, while the computer control is located outside. Further, this instrument has a large specimen chamber and sample mounting system, both easing SEM use within a hot cell. The Quanta SEM uses operating voltages between 1 and 30 kV, allowing for analysis on a wide range of materials and excitation of the x-rays from all elements of interest. Energy dispersive x-ray spectrometry and a back-scatter detector will provide compositional analysis of irradiated specimens. 

  Each sample fracture surface will be examined to verify the straightness of the crack front and also to verify that the crack mode was indeed intergranular. Fracture surfaces will also be used to calibrate the DCPD results. The fracture mode during crack growth will be characterized in terms of the degree of intergranularity and to characterize secondary cracking. Both halves of the CT sample will be examined.

Candidate Materials Evaluation for the Supercritical Water-Cooled Reactor

Participants: G. S. Was, PI, R. Zhou, Post-doctoral scholar; E. West, Graduate Student
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI)

  The supercritical-water-cooled reactor (SCWR) system is being evaluated as a Generation IV concept because it and builds on currently proven light water technology to provide for high thermal efficiency and plant simplification. Development, testing, and selection of suitable materials for cladding and internal components are central to the development of a SCWR. Supercritical water presents unique challenges to the long-term performance of engineering materials. Corrosion and stress corrosion cracking (SCC) in particular have been identified as critical problems because the temperature and the oxidative nature of supercritical water may accelerate the corrosion kinetics and induce stress corrosion cracking. In addition, the presence of radiation can influence corrosion and SCC both by altering the material microstructure and by accelerating corrosion and SCC due to the generation of oxygen and other free radicals via radiolysis. The existing database on the corrosion and stress corrosion cracking of materials in supercritical water is very sparse. Data on the behavior of irradiated alloys is non-existent.

  The objective of the proposed research is to investigate degradation of materials in the supercritical water environment. First, representative alloys from the important classes of candidate materials will be studied for their corrosion and stress-corrosion cracking resistance in supercritical water. These will include ferritic-martensitic steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests will be conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies will be applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and stress-corrosion cracking of alloys in the asreceived and modified/engineered conditions will be examined by irradiating samples using high-energy protons and then exposing them to supercritical water. All these tests will be performed in close coordination with, and as a complement to, the Generation IV testing programs on radiolysis corrosion/SCC of neutron irradiated materials in supercritical water. The research program will be performed by the University of Wisconsin and the University of Michigan. Both these institutions have a proven infrastructure for successfully implementing all aspects of the proposed research. The research will have a strong educational component with several graduate and undergraduate students participating.

Constant Extension Rate Testing of Alloy 690 in Supercritical Water

Participants: G. S. Was, PI; S. Teysseyre, Assistant Research Scientist
Sponsor: Electric Power Research Institute (EPRI)

  Stress corrosion cracking susceptibility of Alloy 690 has been assessed by crack growth rate tests at temperatures just below and above the critical limit of water in an effort to obtain accelerated test data and to assess the likelihood of SCC occurring in primary water conditions. Results have produced IGSCC in alloy 690 and slow but stable crack growth rates (recent unpublished data by Jacko and Andresen). Constant extension rate experiments on a different heat of alloy 690 in 400°C pure SCW containing less than 10 ppb O2 produced IGSCC, indicating a susceptibility to crack initiation under these conditions [Was, 12th Env. Deg.]. A set of CERT experiments are proposed to 1) determine the SCW conditions under which alloy 690 is susceptible to IGSCC in SCW, 2) whether the cracking depends on water density, 3) whether it is the same as that in subcritical water, and possibly 4) whether hydrogen additions can affect cracking.

High Temperature Materials for the Gas-Cooled Fast Reactor

Participants: G. S. Was, PI; G. Gupta, Graduate Student
Sponsor: Idaho National Laboratory

  Both France and the United States have a shared interest in the development of advanced reactor systems that employ inert gas as a coolant. Currently, insufficient physical property data exist to qualify candidate materials for gas-cooled fast reactor (GFR) designs. The overall goal of the GFR materials qualification program is to establish candidate metallic and ceramic materials for GFR designs and to evaluate the mechanical properties, dimensional stability, and corrosion resistance.

  As part of the GFR evaluation of metallic components, a study is underway to determine if grain boundary engineering techniques can improve the high temperature creep strength of candidate metals by optimizing grain boundary structural orientations. As part of this study, the focus of our work is in the following areas: 1) grain boundary engineering of T91 and HT-9, 2) tests to understand the thermal stability of treatments developed to optimize the grain boundary structure of T91, 3) creep testing of alloy T91 in both the as-received and optimized conditions, 4) characterization of the microstructure in the as-received, aged, crept and optimized alloy T91, and 5) grain boundary engineering of nickel-base alloy 617.

Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

Participants: G. S. Was, PI (with J. T. Busby, ORNL – collaborator), G. Jiao, Postdoctoral Scholar
Sponsor: U.S. Department of Energy, Nuclear Engineering Education Research Program (NEER)

  The purpose of this project is to establish that localized deformation in irradiated LWR core internals is a primary factor in irradiation assisted stress corrosion cracking (IASCC). This mode of degradation is a continuing problem in existing LWRs and is expected to be a more serious problem in advanced LWRs and water-cooled Generation IV concepts such as the supercritical water reactor. Progress in understanding the mechanism driving IASCC has been slow due to the difficulty in unfolding the various contributions to the irradiated microstructure that may contribute to IG cracking. However, data from both unirradiated and irradiated austenitic alloys point toward slip localization in the form of intense, dislocation channels as a common factor in the cause of IG cracking in these alloys. The plan of work seeks to establish the role of localized deformation using a series of carefully chosen alloys and a systematic set of experiments designed to quantify the degree of slip localization as a function of alloy stacking fault energy (SFE) and dislocation channeling following irradiation. Experiments in BWR normal water chemistry will provide the link between slip localization and IASCC susceptibility. A primary outcome of the project is to provide guidance for the development of mitigation measures for IASCC.

  Z. Jiao and G. S. Was, “Localized Deformation and IASCC Initiation in Austenitic Stainless Steels,” J. Nucl. Mater. (In press) Z. Jiao and G. S. Was, “The Role of Localized Deformation on IASCC of Proton-Irradiated Austenitic Stainless Steel,” 13th International Conference on Degradation of Materials in Nuclear Power Systems – Water Reactors, T. R. Allen, J. Busby and P. J. King, eds., Canadian Nuclear Society Society. (In press) Z. Jiao, N. Ham and G. S. Was, “Microstructure of He-Implanted and Proton-Irradiated T91 Ferritic-Martensitic Steel,” submitted to J. Nucl. Mater., 367-370, 440-445 (2007).

A Mechanistic Basis for Irradiation Assisted Stress Corrosion Cracking

Participants: G. S. Was, PI, Z. Jiao, Post-doctoral scholar
Sponsor: Electric Power Research Institute (EPRI)

  Irradiation assisted stress corrosion cracking (IASCC) refers to intergranular stress corrosion cracking that is accelerated under the action of irradiation in light water reactor core components. It is referred to as “assisted” because irradiation enhances, or accelerates the IGSCC process over the unirradiated state. IASCC has been a problem in the nuclear industry for the last 30 years and continues to occur due to a lack of understanding of its underlying mechanism. It is the single most important problem in core component cracking in boiling water reactors (BWR) [1] and is of growing importance in pressurized water reactors (PWR). Understanding the mechanism of IASCC is required in order to provide guidance for the development of mitigation strategies.
The IASCC problem has taken on new urgency with the proposal of more advanced water reactor concepts under the Generation IV program [2], such as the supercritical water reactor (SCWR). The SCWR represents a more demanding environment than LWRs in temperature, irradiation dose and the corrosiveness of the media itself. As such, there is an even more pressing need to develop a solution to the IASCC problem. However, in order to do so, the underlying mechanism must first be understood. This proposal aims to establish such an understanding, which will lead directly to mitigation strategies for current and future reactors. The objective is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light water reactor core components.

  Z. Jiao, J. T. Busby and G. S. Was, “Deformation Microstructure of Proton-Irradiated Stainless Steels,” J. Nucl. Mater., 361, No. 2-3, 218-227 (2007). G. S. Was, Z. Jiao and J. T. Busby, “Contribution of Localized Deformation to IGSCC and IASCC,” European Conference on Fracture – 16, Alexandroupolis, Greece, July 2006.

Radiation-Induced Segregation and Phase Stability in Candidate Alloys for the Advanced Burner Reactor

Participants: G. S. Was, J. Peniston, Graduate student
Sponsor: U.S. Department of Energy, Nuclear Energy Research Initiative (NERI)

  The primary objective of this project is to investigate the effect of irradiation on the segregation and phase stability in candidate alloys proposed for application as structural materials for transmutation in the advanced burner reactor. The project will focus on two ferritic-martensitic alloys, and will also include an experimental ODS alloy and an advanced austenitic alloy in a coordinated experimental and modeling effort to investigate the complex electronic–magnetic–elastic interactions between Cr and radiation induced defects controlling radiation induced segregation in F-M alloys. This project will provide a mechanistic understanding of segregation and phase stability that can be used to develop predictive irradiation performance models. It will also provide data against which forthcoming inreactor irradiations can be interpreted and understood, as well as guidance and direction for those experiments.

  This proposal is centered on the two F-M alloys; T91 and HT-9 as both are viable candidates for the ABTR and form the basis for more advanced alloys for the ABR, and will focus on Cr RIS and phase stability in these alloys under irradiation, as these are potentially limiting processes for their application. However, the full, irradiated microstructure needs to be considered as the radiation effects processes are interrelated. Also included in the workscope is a ferritic ODS alloy because of its superior irradiated microstructure stability and strength. In addition, an advanced austenitic candidate alloy, D9, is included because it is the leading austenitic alloy, and yet it potentially can suffer from RIS (of Si) and the formation of deleterious phases (silicides) that could affect performance. Experiments will be conduced by proton and heavy ion irradiation over the dose range 3-100 dpa and the temperature range 350-550°C with the inclusion of He at the highest doses. Analysis of RIS, phase microstructure, dislocation microstructure and hardening will be conducted on all conditions to provide a systematic set of data.

  The modeling tasks will involve ab-initio electronic structure calculations to investigate the configuration-dependent binding and migration energies of Cr with vacancy and interstitial defects, including small clusters. These values will enable development of atomisticbased kinetic Monte Carlo models similar to those employed previously to evaluate He diffusion in Fe and specifically designed to investigate the Cr diffusivity by interstitial and vacancy mechanisms. The RIS tendencies of Cr in F-M alloys will be predicted as a function of temperature and dose, based on migration mechanisms and energies obtained from ab initio calculations. The outcomes of this modeling task will be mechanistic interpretation of the complex Cr RIS behavior, and key diffusional parameters for both continuum level rate theory models and the development predictive RIS models of Cr and Si in F-M alloys. The combined experimental-modeling program is designed to provide a set of data on the behavior of RIS, phase microstructure, dislocation microstructure and hardening as a function of dose and temperature in the range 350-550°C and 3-100 dpa. This data will be used to benchmark RIS and dislocation microstructure models developed from ab initio electronic structure calculations and extended to kinetic Monte Carlo and continuum rate theory (MIK) models.

Stress Corrosion Cracking and Corrosion of Candidate Alloys for the Supercritical Water Reactor Concept

Participants: G. S. Was, PI, S. Teysseyre, Assistant Research Scientist
Sponsor: U.S. Department of Energy, International Nuclear Energy Research Initiative (INERI)

  Supercritical water presents unique challenges to the long-term operation of engineering materials. The generation of oxygen and hydrogen gas by radiolysis and the high solubility of these gases in supercritical water may result in higher corrosion and stress corrosion cracking rates than experienced with other reactor designs. In addition, radiation may accelerate or assist the stress corrosion cracking in the reactor region, and stress corrosion cracking and accelerated corrosion may occur in the preheat and cool-down sections of the circuit. The existing data base on the corrosion and stress corrosion cracking of austenitic stainless steel and nickel based alloys in supercritical water is very sparse. Data on the behavior of irradiated alloys is non-existent. Therefore, the focus of this work will be stress-corrosion-cracking behavior of candidate fuel cladding and structural materials in the unirradiated and irradiated conditions. Two high-temperature autoclave systems have been built to test the SCC and corrosion behavior of unirradiated and proton-irradiated materials. Proton irradiation is used as a surrogate for neutron irradiated material in order to get a first look at the response of candidate alloys to irradiation, and also to cover alloys for which there are currently no neutron irradiated samples for testing. A third high-temperature autoclave coupled to a loading system, and capable of straining up to four tensile samples in constant extension rate mode or one compact tension sample in crack growth rate mode is being built and operated at the University of Michigan (U-M). This system is being constructed for conducting experiments on neutron-irradiated materials. The resulting data will be used to further narrow the list of promising materials and develop appropriate stress-corrosion-cracking correlations. The capability to conduct both crack growth rate and constant extension rate tensile experiments on neutron-irradiated samples will constitute the first facility capable of assessing SCC of neutron irradiated alloys in supercritical water.

  The work plan for this three year (FY05-FY07) program consists of four principal tasks; 1) the completion of a facility to conduct crack growth rate and constant extension rate tensile tests on highly radioactive, neutron irradiated samples in supercritical water, 2) constant extension rate tests and crack growth rate tests of candidate alloys in supercritical water, 3) proton irradiation and constant extension rate tests of proton-irradiated samples in supercritical water and 4) constant extension rate tests and crack growth rate tests of candidate neutron-irradiated alloys in supercritical water.