Arvind S. Kumar

The influence of specimen size on charpy impact testing of unirradiated HT-9

Abstract The effect of specimen size on the impact properties of HT-9 has been studied. The data from precracked and notched-only specimens of HT-9 were used along with previously published data on other steels to develop correlations that predict from subsize specimen data the ductile brittle transition temperature (DBTT) and the upper shelf energy (USE) anticipated for full size specimens. Each correlation is based on physical insight and can be applied equally well to both precracked and notched-only specimens. The correlation for USE works best for fracture energy in a range characteristic of irradiated materials, while a previously published correlation works better for more ductile materials. The DBTT correlation, however, works well on all materials for which sufficient data have been published.

The influence of silicon on void nucleation in irradiated alloys

Abstract The addition of silicon to pure nickel. Ni-Cr alloys and Fe-Ni-Cr alloys raises the diffusivity of each of the alloy components. The resultant increase in the effective vacancy diffusion coefficient causes large reductions in the nucleation rate of voids during irradiation. This extends the transient regime of swelling, which is controlled not only by the amount of silicon in solution but also by the precipitation kinetics of precipitates rich in nickel and silicon.

Preliminary FRAPCON-3Th steady-state fuel analysis of ThO2 and UO2 fuel mixtures

Abstract Recent investigations into the performance and economics of mixed thoria-urania (ThO2/UO2) fuel cycles in light water reactors indicate that there may be advantages to using these fuels at high burnups. The Idaho National Engineering and Environmental Laboratory (INEEL) modified FRAPCON-3, a U.S. Nuclear Regulatory Commission-sponsored software package developed by Pacific Northwest National Laboratory for use on mixed thoria-urania fuels. The modifications constituted the first stage of fuel performance evaluations supported by the Nuclear Energy Research Initiative (NERI) project titled Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors. The goal of this NERI project is to develop mixed ThO2/UO2 fuels that can be operated to a relatively high burnup level in current and future commercial power reactors. This paper describes in detail the INEEL’s modifications to the FRAPCON-3 thermal conductivity subroutine FTHCON and the techniques used to validate the modifications. The paper presents the general fuel design criteria used to model mixed thoria-urania fuel and a steady-state analysis of a mock thoria-urania fuel using the FRAPCON-3Th code. The paper also presents the data analyses for the mock thoria-urania fuel and offers suggestions for future upgrades and improvements to the code.

Effect of specimen size on the impact properties of neutron irradiated A533B steel

Abstract A new methodology is proposed to correlate the upper shelf energy (USE) of full-size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material. The methodology appears to be more satisfactory than those methodlogies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with speciment size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full-, half-, and third-size specimens. The ductile-to-brittle transition temperture (DBTT) increased due to irradiation at 150°C to a nominal fluence of 1.0 × 1019 n/cm2 (E >MeV) by 78, 83 and 70°C for full-, half-, and third-size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

Dimensional change correlations for 20% cold-worked AISI 316 stainless steel for fusion applications☆

Abstract Correlations developed for void swelling and irradiation creep often arise from two separate data bases and sometimes are not well-matched, reflecting apparently minor but actually significant differences in environmental history. Previously published correlations are very empirical in nature and reflect prevailing misconceptions about the parametric sensitivity of swelling and irradiation creep. In this study a well-defined data base for swelling of 20% cold-worked AISI 316 is used in conjunction with the latest insights on swelling and creep to develop two matched dimensional change correlations for temperatures in the range of 300 to 650 °C. The new creep model, in particular, is much simpler than the complicated empirical correlation published earlier for this steel.

Transmutation-induced deposition profiles in halos surrounding spherical precipitates

Abstract The deposition profiles that form around precipitates undergoing energetic transmutation reactions can be used to study the effects of helium, hydrogen and lithium on microstructural development during neutron irradiation. A completely general derivation has been performed to determine the deposition profile concentrations for any transmutation reaction and precipitate/matrix combination. Calculations of the deposition profiles were performed for the 10B(n, α)7Li reaction and selected precipitate/matrix stopping power ratios. With increasing ratio the deposition profiles become more uniform but the reaction products are deposited over larger volumes leading to lower concentrations in the matrix.

Kinetics of Radiation-Induced Segregation in Ternary Alloys

Model calculations of radiation-induced segregation in ternary alloys have been performed, using a simple theory. The theoretical model describes the coupling between the fluxes of radiation-induced defects and alloying elements in an alloy A-B-C by partitioning the defect fluxes into those occurring via A-, B-, and C-atoms, and the atom fluxes into those taking place via vacancies and interstitials. The defect and atom fluxes can be expressed in terms of concentrations and concentration gradients of all the species present. With reasonable simplifications, the radiation-induced segregation problem can be cast into a system of four coupled partial-differential equations, which can be solved numerically for appropriate initial and boundary conditions. Model calculations have been performed for ternary solid solutions intended to be representative of Fe-Cr-Ni and Ni-Al-Si alloys under various irradiation conditions. The dependence of segregation on both the alloy properties and the irradiation variables, e.g., temperature and displacement rate, was calculated. The sample calculations are in good qualitative agreement with the general trends of radiation-induced segregation observed experimentally.

Saturation of proton-induced swelling in AISI 316

Abstract There is some evidence which suggests that void swelling may not increase continuously with increasing irradiation but may saturate at a level dependent on irradiation conditions, helium/dpa level or artifacts of the simulation such as the presence of injected interstitials. An experiment was therefore performed to determine the saturation level of swelling of annealed AISI 316 at 625°C in the absence of helium and injected interstitials. Using 140 keV protons and step-height measurements it was found that saturation did not occur until 260% swelling was achieved at ~ 600 dpa. Prior to saturation the swelling curve exhibited the anticipated bilinear form with a steady state swelling rate of 0.64%/dpa based on a 20 eV threshold energy.

Layer formation on metal surfaces in lead–bismuth at high temperatures in presence of zirconium

Abstract If the operating temperature lead–bismuth cooled fission reactor could be extended to 800 °C, they could produce hydrogen directly from water. A key issue for the deployment of this technology at these temperatures is the corrosion of the fuel cladding and structural materials by the lead–bismuth. Corrosion studies of several metals were performed to correlate the interaction layer formation rate as a function of time, temperature, and alloy compositions. The interaction layer is defined as the narrow band between the alloy substrate and the solidified lead–bismuth eutectic on the surface. Coupons of HT-9, 410, 316L, and F22 were tested at 550 and 650 °C for 1000 h inside a zirconium corrosion cell. The oxygen potential ranged from approximately 10−22 to 10−19 Pa. Analyses were performed on the coupons to determine the depth of the interaction layer and the composition, at each time step (100, 300, and 1000 h). The thickness of the interaction layer on F22 at 550 °C was 25.3 μm, the highest of all the alloys tested, whereas at 650 °C, the layer thickness was only 5.6 μm, the lowest of all the alloys tested. The growth of the interaction layer on F22 at 650 °C was suppressed, owing to the presence of Zr (at 1500 wppm) in the LBE. In the case of 316L, the interaction layers of 4.9 and 10.6 μm were formed at 550 and 650 °C, respectively.

Irradiation and Thermal Creep of a Titanium-Modified Austenitic Stainless Steel and Its Dependence on Cold Work Level

Abstract A titanium-modified austenitic stainless steel similar to the fusion PCA alloy was creep tested at three cold work levels under both thermal aging and neutron irradiation conditions. The creep behavior exhibits a complex nonmonotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550°C. The 20% cold work level is preferable to the 10% level, in terms of its initial strength, in-reactor creep and swelling response.