TerraPower has revitalized the manufacturing of HT9, and optimized its heat treatment to be used as fuel claddings in the Traveling Wave Reactor. TerraPower initiated a comprehensive test program to compare the mechanical and thermal creep properties of the optimized TerraPower HT9 to the historical HT9. The uniaxial tensile tests show TerraPower HT9 has improved yield strength compared to historical HT9 across all temperatures. Charpy impact toughness tests on TerraPower HT9 show lower upper shelf energy compared to historical HT9, but the ductile to brittle transition temperature (DBTT) between the steels remain similar. Three point bend tests between room temperature and 400 °C show comparable fracture toughness to those of historical HT9. Thermal creep test data suggest TerraPower HT9 have improved creep strength compared to historical HT9.

The void swelling and microstructure evolution of ferritic-martensitic alloys HT9, T91 and T92 were characterized following irradiation with Fe++ ions at 460 degrees C to damage levels of 75-650 displacements per atom with 10 atom parts per million pre-implanted helium. Steady state swelling rate of 0.033%/ dpa was determined for HT9, the least swelling resistant alloy, and 0.007%/ dpa in T91. In T91, resistance was due to suppression of void nucleation. Swelling resistance was greatest in T92, with a low density (similar to 1 x 10(20) m(-3)) of small voids that had not grown appreciably, indicating suppression of nucleation and growth. Additional heats of T91 indicated that alloy composition was not the determining factor of swelling resistance. Carbon and chromium-rich M2X precipitates formed at 250 dpa and were correlated with decreased nucleation in T91 and T92, but did not affect void growth in HT9. Dislocation and G-phase microstructure evolution was analyzed up to 650 dpa in HT9. Published by Elsevier B.V.

G.S. Was, ⇑ Z. Jiao, E. Getto, K. Sun, A.M. Monterrosa, S.A. Maloy, O. Anderoglu, B.H. Sencer and M. Hackett University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48109, United States Los Alamos National Laboratory, MST-8, Ms-H816 LANL, Los Alamos, NM 87545, United States Idaho National Laboratory, P.O. Box 1625, MS 6188, Idaho Falls, ID 83415, United States TerraPower LLC, 330 120th Avenue NE Suite 100, Bellevue, WA 98005, United States University of Michigan, 413B Space Res Bldg., Ann Arbor, MI 48109, United States

Ferritic/martensitic steels are candidates for fast reactors because of their sodium compatibility, superior resistance to corrosion and radiation damage, including swelling, and excellent thermal conductivity and thermal expansion coefficient. One significant limitation of any cladding material is its susceptibility to swelling at high doses. While HT9 has neutron irradiation performance data up to ∼200 dpa, dose requirements for the Traveling Wave Reactor (TWR) may be much higher. Obtaining higher-dose data will take many years, but in the interim, heavy ion irradiation could provide a useful tool toward predicting the swelling trends beyond 200 dpa. In this study, HT9 was irradiated from 440–480°C using 5 MeV Fe++ ions. The samples are compared to a portion of HT9 fuel assembly duct from FFTF, which was characterized after neutron irradiation at 440°C with an accumulated dose of 155 dpa. Comparisons are made of the void size and density using transmission electron microscopy (TEM). The increase in dose from 280 dpa to 375 dpa increased void size, number density and swelling at 440°C, while swelling was generally lower at 480°C for the same helium pre-implantation conditions. Helium generally enhanced the nucleation of voids, as measured by the void density.

The TerraPower Traveling Wave Reactor (TWR) is a sodium-cooled fast reactor design that utilizes a high-burnup metallic uranium fuel cycle. The fuel system depends on a cladding material with demonstrated swelling resistance to high doses as well as adequate thermal creep strength. HT9 steel is a leading cladding candidate for the first TWR, having demonstrated excellent swelling and strain performance to doses > 200 dpa. A strain model was developed as a design tool to predict fuel pin deformation as a function of irradiation dose, stress, and temperature. The sources of strain deformation will be described along with the uncertainties in utilizing existing data to build a mechanistic model. The strain model is then incorporated into a fuel performance code to provide new insight in deformation behavior of HT9 fuel pins.

Progress in understanding radiation damage in structural materials is hampered by the lack of test reactors, long irradiations and high cost. Here we show that through strict control of experimental parameters and accounting for He production and damage-rate differences, the microstructure of ion-irradiated ferritic-martensitic steel closely resembles that created in-reactor across the full range of microstructure features. The level of agreement establishes for the first time the capability to tailor ion irradiation to emulate in-reactor radiation damage. (C) 2014 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Nuclear power systems are constructed from a wide range of metallic alloys, subjected to taxing environmental conditions, and required to resist cracking and degradation of their principal mechanical and physical properties for decades. Fusion welding is, in general, the joining method of choice due to its hermeticity, high joint efficiency, and economic advantages relative to mechanical or brazed joints. However, it is often fusion welds, or their heat affected zones that prematurely degrade or fail due to the complex interplay of physical defects, compositional gradients, metallurgical changes, and residual stresses. This article presents the current mechanistic understanding of welding defects, reviews recent developments in assessing residual stresses & plastic strains and relates these factors to the in-service performance of welds. Lastly, the weldability of common structural alloy systems is reviewed.

Rate theory modeling was used to simulate the effects of oversized solute additions on radiation-induced segregation in austenitic stainless steels. The purpose was to understand the effects of a solute-vacancy trapping mechanism on radiation-induced segregation and to define key parameters that most affect segregation behavior. Sensitivity analysis of the model showed the solute-vacancy binding energy to be the most important model parameter. Binding energies from ab initio first principles were calculated for oversized solutes of Pt, Ti, Hf and Zr, with energies of 0.31, 0.39, 0.71 and 1.08eV, respectively. Differences in binding energies, despite similar sizes of the atoms, suggests that the short-range electronic interactions play an important role in determining binding energy. The model results show oversized solutes to be most effective at reducing grain boundary Cr depletion at temperatures of 450–500°C for a dose rate applicable to proton irradiations. The reduction increases with increasing oversized solute concentration, where it saturates at approximately 0.1at.%.

Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3dpa at 400°C and 1dpa at 500°C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3dpa at 400°C and above 1dpa at 500°C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS.

The addition of oversized solutes has the potential to reduce the effects of radiation-induced segregation (RIS) in austenitic alloys. This RIS has been implicated as one of several factors in enhancing stress corrosion cracking (SCC) under irradiation, so oversized solute additions could promote SCC resistance. Either Zr or Hf was added to 316-type stainless steel, at levels between 0.05 and 0.37 at. pct. Samples were irradiated with 3 MeV protons to 3 dpa at 400 °C and analyzed using high-resolution–scanning transmission electron microscopy (HR-STEM) with energy-dispersive X-ray spectroscopy (EDS), to measure the grain-boundary (GB) composition. The Zr additions substantially reduced the amount of RIS, while the Hf was much less effective. Despite similar sizes, first-principles calculations using the Vienna Ab Initio Simulation Package (VASP) demonstrate that solute-vacancy binding for Zr is 1.05 eV vs 0.69 eV for Hf. This difference results in the greater effectiveness of Zr in reducing RIS, as determined by kinetic rate theory calculations, in agreement with experimentally-measured results.

AbstractDeleterious effects of radiation in nuclear reactor systems cause material degradation and the potential for component failure. Radiation damage is fundamentally due to freely migrating point defects produced in collision cascades. A reduction in the freely migrating point defect population should, then, reduce radiation damage and increase component lifetime. The addition of oversized solute atoms such as Zr or Hf to 316SS, a common structural material in reactors, is expected to reduce point defect population through a trapping mechanism that enhances recombination. The mechanism, however, requires a strong binding energy between the oversized solute atom and vacancies in order for the mechanism to significantly reduce the defect population. Experimental measurements of this binding energy are unavailable, but can be determined with atomistic calculations. Ab initio methods are used here to determine binding energies and atomic volumes of either Hf or Zr oversized solutes with vacancies in a face-centered cubic Fe matrix. The binding energies are then used to parameterize a kinetic rate-theory model, which is used here to calculate radiation-induced segregation (RIS). The calculated values of RIS are then compared to experimental measurements to benchmark the calculations and offer insight into the proposed point defect trapping mechanism.

Oversize solute additions to stainless steels hold promise for reducing radiation-induced segregation (RIS), possibly delaying the onset and severity of irradiation-assisted stress-corrosion cracking (IASCC). The Modified Inverse Kirkendall (MIK) model for RIS in austenitic stainless steels was adapted to include the effects of defect trapping, which is expected to be caused by oversized solutes, on RIS. The model accounts for the sensitivity of RIS to both dose rate and temperature. Model results show that the primary contribution to the reduction in RIS occurs through vacancy trapping and recombination with migrating interstitials, requiring strong solute binding energies on the order of 1.0 eV. The maximum reduction in RIS due to defect trapping increases with dose rate and temperature. The general trends of the model are consistent with experimental data from proton and neutron irradiations.